ORNL-TM-3039 by Jon Morrow

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Contract No. W-7405-eng-26 Reactor Division

MSRE SYSTEMS AND COMPONENTS PERFORMANCE by Molten-Salt Reactor Experiment S t a f f

--Edited and Compiled by R. H. Guymon

JUNE 1973

This document t a i n s information of a preliminary nature and was prepared primarily f o r i n t e r n a l use a t t h e Oak R i d g e National Laboratory. It i s subject t o r e v i s i o n or correction and t h e r e f o r e does not represent a f i n a l report. NOTICE:

OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 operated by U. S. ATOMIC ENERGY COMMISSION NOTICE

T h b report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, m k e a any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights.

DISTRtdUTION OF THIS DOCUMENT IS UNUM -

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iii

CONTENTS Page 1 3

..,..... ..............-..... .... ................ 2. DESCRIPTION OF THE PLANT . . . . . . . . . . . . 4 3. CHRONOLOGY OF OPERATION AND MAINTENANCE . . . . . 9 4. PLANT PERFORMANCE AND STATISTICS. . . . . . . . . . 15 4.1 Cumulative Statistics. . . . . . . . . . . . . . . . 15 4-2 Availability during Various Periods. . . . . . . . 18 4.3 Interruptions of Operations. . . . . . . . . . . . . 22 4.4 Time Required for Operational Tasks. . . . . . . . 48 4.5 Changes Made in the Plant. , . . . . . . . . . . . . 59 4.6 Tabulation of Recorded Variables at Full Power . . . . 63 5. FUELSYSTEM.. . . . . . . . . . . . . . . . . . . . . . . 89 5.1 Description. . . . . . . . . . . . . . . . . 89 5.2 Purging of Moisture and Oxygen from the System . . . . 89 5.3 Fuel-Circulating-System Volume Calibration . . . . . 91 . 5.4 Drain Times. .-.. . . . . . . . . . . . . . . . . . 91 5.5 Mixing of Fuel and Flush Salts . . . . . . . . . . . 93 5.6 Primary System Leak. . . . . . . . . -.. . . . . . . 93 5.7 Operation . . . . . . . . . . . . . . . . . . . . . 94 5.8 Fuel Pump and Overflow Tank. .. .. . .. .. .. .. . . 110 94 5.9 Primary Heat Exchanger . . 5.10 Reactor Vessel and Reactor Access Nozzle . .. . 112 5.11 Fuel and Flush Salt Drain Tanks. .. 129 . . . . 137 6. COOLANTSYSTEM . . . . . . . . . . . . . . . 6.1 Description . . . . . . . . . . . . . . . . 137 6.2 Purging Moisture and Oxygen from the System. . . . . . 137 6.3 Coolant Circulating System Calibration and Drain The. 137 6.4 Operation. . . . . . . . . . . . 139 6.5 Coolant Salt circulating Pump. . . . 139 6.6 Radiator . . . . . . . . . . . . . . 140 6.7 Performance of Main Blowers, MB-1 and MB-3 . 147 6.8 Coolant Drain Tank .... . .. . 159 ABSTRACT1. INTRODUCTION

COVER-GAS SYSTEM 7.1 7.2 7.3

e................161

. . . . . . . .. .. . . . . .. . .. .

Initial Testing Normal Operation Conclusions and Recommendations.

161 162 164

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. iv

........................ 165 8.1 Description ...................... 165 8.2 Experience with Coolant Salt Off-gas System . . . . . . 167 . 8.3 Experience with the Fuel Off-Gas System . . . . . . . . 173 8.4 Difficulties with the Fuel Off-gas System . . . . . . . 177 8.5 Subsequent Operating Experience with Fuel Off-gas System ....................... 201 8.6 Discussion and Conclusions . . . . . . . . . . . . . . 212 9. FUEL ANDCOOLANTPUMPLUBEOILSYSTEMS . . . . . . . . . . . . 215 9.1 Description ...................... 215 9.2 Installation and Early Problems. . . . . . . . . . . . 215 Addition of Syphon Tanks to the Oil Catch Tanks . . . . 216 9.3 9.4 OilLeakage . . . . . . . . . . . . . . . . . . . . . . 217 9.5 Change-out of Oil Pumps . . . . . . . . . . . . . . . . 218 9.6 Test Check of One'Oil System Supplying Both the Fuel and Coolant Salt Pumps . . . . . . . . . . . . . . . 218 Oil Temperature Problems . . . . . . . . . . . . . . . 218 . 9.7 Increase in Radiation Levels the Lube Oil Packages. 219 9.8 Analysis of Oil .................... 220 9.9 Replacement of Oil . . . . . . . . . . . . . . . . . . 220 9.10 9.11 Discussion agd Recommendations . . . . . . . . . . . . 222 10. COMPONENT COOLING . . . . . . . . . . . . . . . . . . 223 10.1 Primary System .................... 223 10.2 Secondary System . . . . . . . . . . . . . . . . . . . 231 232 11. VENTILATIONSYSTEM ..................... 11.1 Description. . . . . . . . . . . . . . . . . . . . . . 232 11.2 Operating Experience . . . . . . . . . . . . . . . . . 232 11.3 Conclusions and RecommendatZons. . . . . . . . . . . . 235 12. WATER ........................ 237 12.1 Potable Water System . . . . . . . . . . . . . . . . . 237 12.2 Process Water System . . . . . . . . . . . . . . . . . 238 12.3 Cooling Tower Water System . . . . . . . . . . . . . . 239 12.4 Treated Water System . . . . . . . . . . . . . . . . . 240 12.5 Condensate System . . . . . . . . . . . . . . . . . . 245 12.6 Nuclear Instrument PenetratLon. . . . . . . . . . . . .245 12.7 General Water Systems Conclusions. . . . . . . . . . . 246 247 13 . LIQUID WASTE SYSTEM ..................... 13.1 Description. . . . . . . . . . . . . . . . . . . . . . 247 13.2 Preliminary Testing . . . . . . . . . . . . . . . . . . .247 13.3 Operating Experience . . . . . . . . . . . . . . . . . 248 14. HELIUM DETECTOR SYSTEM . . . . . . . . . . . . . . . . . 251 14.1 Description and Method of Operation. . . . . . . . . . . .251 14.2 Calibration . . . . . . . . . . . . . . . . . . . . . .251 14.3 Operating Experience . . . . . . . . . . . . . . . . . .255 14.4 Discussion and Recommendations . . . . . . . . . . . . 256 8

OFF-GAS SYSTEM

SYSTEMS

LEAK

Page

. . . . . . . . . . . . . . . . . . . . 259 15.1 Description . . . . . . . . . . . . . . . . . . . . . 259 15.2 Operating Experience . . . . . . . . . . . . . . . . . 259 15.3 Conclusions . . . . . . . . . . . . . . . . . . . . . . 260 261 16 . ELECTRICAL SYSTEM . . . . . . . . . . . . . . . . . . . . . . 16.1 Description. . . . . . . . . . . . . . . . . . . . . . 261 16.2 Alternating Current System . . . . . . . . . . . . . . 262 16.3 250-V dc System . . . . . . . . . . . . . . . . . . . . 265 16.4 Reliable Power System . . . . . . . . . . . . . . . . . 265 16.5 48-V dc System . . . . . . . . . . . . . . . . . . . . 266 16.6 Diesel Generators. . . . . . . . . . . . . . . . . . . 267 16.7 100-kVa Variable Frequency Motor-Generator Set . . . . 267 268 16.8 Conclusions . . . . . . . . . . . . . . . . . . . . . . 17 . HEATERS . . . . . . . . . . . . . . . . . . . . . . . . . . . 269 17.1 Description . . . . . . . . . . . . . . . . . . . . . . 269 17.2 Preoperational Checkout . . . . . . . . . . . . . . . . 273 17.3 System Heatup . . . . . . . . . . . . . . . . . . . . . 273 17.4 Heater Performance . . . . . . . . . . . . . . . . . . 278 17.5 Systems Cooldown Rate . . . . . . . . . . . . . . . . . 287 17.6 Discussion . . . . . . . . . . . . . . . . . . . . . . 291 18. SAMPLERS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 292 18.1 Fuel Sampler-Enricher . . . . . . . . . . . . . . . . . 292 18.2 Coolant Sampler. . . . . . . . . . . . . . . . . . . . 320 19 . CONTROLRODS . . . . . . . . . . . . . . . . . . . . . . . . . 323 . 323 -19.1 Description. . . . . . . . . . . . . . . . . . . . . . 19.2 Initial Testing . . . . . . . . . . . . . . . . . . . . 324 19.3 Periodic Testing . . . . . . . . . . . . . . . . . . . 325 19.4 . . . . . . . . . . . . . . . . . 326 19.5 Improved Rod Scram Testing . . . . . . . . . . . . . . 336 19.6 Maintenance Experience . . . . . . . . . . . . . . . . 338 19.7 Discussion and Recommendations . . . . . . . . . . . . 339 20. FREEZE VALVES . . . . . . . . . . . . . . . . . . . . . . . . 341 20.1 Introduction . . . . . . . . . . . . . . . . . . . . . 341 20.2 Description of the Design of the Freeze Valves . . . . 341 20.3 Description of the Operation of the Freeze Valves . . . 344 20.4 Operating Experience . . . . . . . . . . . . . . . . . 346 20.5 Recommendations. . . . . . . . . . . . . . . . . . . . 353 21 . FREEZE FLANGES . . . . . . . . . . . . . . . . . . . . . . . . 356 21.1 Description . . . . . . . . . . . . . . . . . . . . . . . 356 21.2 Operation . . . . . . . . . . . . . . . . . . . . . . . 356 21.3 Conclusions . . . . . . . . . . . . . . . . . . . . . . 358 15

INSTRUMENT AIR SYSTEM

Operating Experience

....

.......

......

...

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vi Page

. . . . . . . . . . . . . . . . . . . . . . . . . 360 22.1 Description and Criteria . . . . . . . . . . . . . . . 360 22.2 Methods Used to Assure Adequate Containment and Results . . . . . . . . . . . . . . . . . . . . . . . 360 22.3 Discussion of Cell Leak Rate Determination . . . . . . 370 22.4 Vapor Condensing System . . . . . . . . . . . . . . . 376 22.5 Recommendations .................... 377 23 . BIOLOGICAL SHIELDING AND RADIATION LEVELS . . . . . . . . . . 380 23.1 Radiation Surveys -Approach to Power . . . . . . . . . 380 23.2 Radiation Levels During Operation. . . . . . . . . . . 383 23.3 Conclusions . . . . . . . . . . . . . . . . . . . . . . 386 24. INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . 389 24.1 Introduction . . . . . . . . . . . . . . . . . . . . . 389 24.2 Description . . . . . . . . . . . . . . . . . . . . . . 389 24.3 Initial Checkout and Startup Tests . . . . . . . . . . 393 24.4 Periodic Testing . . . . . . . . . . . . . . . . . . .394 24.5 Performance of the Nuclear Safety Instrumentation. . . 397 24.6 Performance of the Wide-Range Counting Channels. . . . 400 24.7 Performance of the Linear Power Channels . . . . . . . 402 BF. Nuclear Instrumentation. . . . . . . . . . . . . . 403 24.8 24.9 Nuclear Instrument Penetration . . . . . . . . . . . . 403 24.+0 Performance of the Process Radiation Monitors . . . . . 403 24.11 Performance of the Personnel Radiation Monitoring and Building Evacuation System . . . . . . . . . . . . . 404 24.12 Performance of the Stack Monitoring System . . . . . . 404 24.13 Performance of Thermocouples and the Temperature Readout and Control Instrumentation. . . . . . . . . 404 24.14 Performance of Pressure Detectors . . . . . . . . . . . 408 24.15 Performance of Level Indicators . . . . . . . . . . . 408 24.16 Performance of the Drain Tank Weighing Systems . 409 24.17 Performance of the Coolant Salt Flowmeters . . . . . . 410 24.18 Performance of Relays . . . . . . . . . . . . . . . . . 411 24.19 Training Simulation . . . . . . . . . . . . . . . . . 411 24.20 Miscellaneous . . . . . . . . . . . . . . . . . . . . 412 24.21 Conclusions and Recommendations . . . . . . . . . . . . 412 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 413 22

CONTAINMENT

MSRE SYSTEMS AND COMPONENTS PERFORMANCE

ABSTRACT When t h e MSRE was shut down i n December 1969, i t had accumulated 13,172 full-power hours of operation,

S a l t had been c i r c u l a t e d i n t h e

f u e l system f o r 21,788 hours and i n t h e coolant system f o r 26,076 hours. E s s e n t i a l l y no d i f f i c u l t y w a s encountered with t h e primary system during operation.

After t h e r e a c t o r was shut down t h e r e w a s an i n d i c a t i o n

of a l e a k i n t h e drain-tank piping a t / o r near a f r e e z e valve.

Further in-

v e s t i g a t i o n w i l l be made later as t o t h e nature and cause of t h i s leak. There w a s a small continuous leakage of l u b r i c a t i n g o i l i n t o t h e f u e l pump throughout t h e operation.

This, togdther with s a l t m i s t , caused peri-

odic plugging i n t h e off-gas system which was designed f o r clean helium. F i l t e r s i n s t a l l e d i n t h e main l i n e s proved very e f f e c t i v e . I n e a r l y operation, d i f f i c u l t y w a s encountered with t h e coolant radiator.

The doors would not seal, t h e r e were too many a i r leaks, and thermal

i n s u l a t i o n w a s inade+ate.

After t h e s e were repaired, t h e system operated

f i n e except f o r a f a i l u r e of one of t h e main blowers and some trouble with t h e blower bearings. Only r e l a t i v e l y minor d i f f i c u l t i e s were encountered with t h e containment and o t h e r systems.

INTRODUCTION P. N. Haubenreich Operation of t h e MSRE constituted a major s t e p toward the objectives of t h e Molten-Salt Reactor Program.

The goal of t h i s program i s t h e de-

velopment of l a r g e , fluid-fuel reactors having good neutron econorqy and producing law-cost e l e c t r i c i t y . '

The MSRE w a s b u i l t t o demonstrate t h e

p r a c t i c a l i t y of t h e molten-salt r e a c t o r concept, with emphasis on t h e compatibility of t h e materials (fluoride salts , graphite , and container a l l o y ) , the performance of key components , and t h e r e l i a b i l i t y and'maint a i n a b i l i t y of t h e plant. I n t h e course of 5 years of t e s t i n g and operation of t h e MSRE (1964

1969) t h e operators accumulated considerable experience w i t h t h e various components and systems i n t h e r e a c t o r plant.

This experience, properly

disseminated, should be valuable i n t h e continuing development of moltensalt reactors.

Much has already been published i n t h e Molten S a l t Reactor

Program semiannual progress reports ( R e f s . 2 t o 14) but such reporting is pie'cemeal and sometimes r a t h e r condensed. much very d e t a i l e d information i n t e s t

On t h e other hand, t h e r e i s

reports and operations and mainte-

nance f i l e s , but these are r e l a t i v e l y inaccessible and s p e c i f i c informat i o n i s tedious t o e x t r a c t .

It w a s considered worthwhile, t h e r e f o r e , t o

e x t r a c t , organize, evaluate and report t h e experience with MSRE systems and components. The purpose of t h i s report is t o present a convenient, comprehensive

description of t h e MSRE experience.

The first chapters b r i e f l y describe

t h e p l a n t and o u t l i n e t h e chronology of i t s operation.

N e x t there is a

chapter on t h e o v e r a l l p l a n t performance, includihg s t a t i s t i c s r e l a t i v e t o r e l i a b i l i t y and maintainability.

The chapters which follow a r e each

devoted t o one system o r component.

Finally, t h e r e i s a chapter of dis-

cussion and conclusions.

LiJ

3 2.

DESCRIPTION OF THE PLANT

R. H, Guymon 5-

The MSRE was a single-region, c i r c u l a t i n g molten-salt fueled, thermal r e a c t o r which produced heat at t h e r a t e of about 8 Mw.

The f u e l was UF4

i n a c a r r i e r salt of LiF-BeF$ZrFt.,.’ A t t h e operating temperature of 1200°F, t h i s salt i s a l i q u i d which has very good physical properties:

viscosity

about 8 centipoise, density about 135 l b / f t 3 , and’vapor pressure l e s s than 0.1 mm Hg. The design conditions a r e shown i n t h e flow diagram (Fig. 2-11,

general arrangement of t h e plant i s shown i n Fig. 2-2. ~

The

The salt-containing

piping and equipment was made of Hastelloy-N, a nickel-molybdenum-ironchromium a l l o y with exceptional r e s i s t a n c e t o corrosion by molten f l u o r i d e s and with high s t r e n g t h at high temperature. I n t h e r e a c t o r primary system, t h e f u e l salt w a s r e c i r c u l a t e d by t h e sump-type c e n t r i f u g a l pump through t h e shell-and-tube h e a t exchanger and

t h e r e a c t o r vessel. i n Figure 2-3.

The 5-ft diam. by 8-ft high r e a c t o r v e s s e l i s shown

It w a s f i l l e d with 2-in. by 2-in.

graphite moderator s t r i n g e r s

which had grooves machined i n t h e s i d e s t o form flow channels f o r t h e f u e l

salt.

Since t h e graphite i s compatible with t h e molten salt, it was possible

t o use unclad graphite which i s d e s i r a b l e t o obtain good neutron economy. The heat generated i n t h e f u e l salt as it passed through t h e r e a c t o r w a s t r a n s f e r r e d i n t h e heat exchanger t o a molten LIF-BeF2 coolant s a l t .

The

coolant salt w a s c i r c u l a t e d by means of a second sump-type pump through t h e h e a t exchanger and through t h e r a d i a t o r . flow blowers p a s t t h e r a d i a t o

A i r was blown by two a x i a l

tubes t o remove t h e heat which w a s s e n t

up t h e coolant s t a c k where it w a s d i s s i p a t e d t o t h e atmosphere.

Drain tanks were provided f o r s t o r i n g t h e f u e l and coolant salts at high temperature when t h e r e a c t o r was not operating.

LiF-BeF2 f l u s h salt

used f o r flushing t h e f u e l system before and a f t e r maintenance was s t o r e d i n the fuel flush tank.

The salts were drained by gravity.

They were

t r a n s f e r r e d back t o t h e c i r c u l a t i n g systems by pressurizing t h e tanks with helium.

i c\

FILTEl

STrn

FbN

I-= I I

i i

Fig. 2.1

Design Flowsheet of the MSRE

OWL-DWG 634209R

REACTORCONTROL

6. COOLANT PUMP

Fig. 2.2

Layout of the MSRE

12. CONTAINMENT VESSEL 13. FREEZE VALVE

6 OR&-LR-OW

GRAPHITE SAMPLE ACCESS PORT

61097RqA

FLEXIBLE CONDUIT TO CONTROL ROD DRIVES

COOLING AIR LINES

ACCESS PORT COOLING JACKETS

FUEL OUTLET CORE ROD THIMBLE LARGE GRAPHITE SAMPLES

REACTOR ACCESS PORT SMALL GRAPHITE SAMPLES HOLD-DOWN ROD OUTLET STRAINER

CORE CENTERING GRID

FLOW DISTRIBUTOR VOLUTE GRAPHITE -MODE

CORE WALL COOLING ANNUI REACTOR CORE CAN REACTOR VESSEL

ANTI-SWIRL VANES VESSEL DRAIN LINE

Fig. 2.3

SUPPORT GRID

Details of the MSRE Core and Reactor Vessel

The f i s s i o n product gases, krypton and xenon, were removed continuously from t h e c i r c u l a t i n g f u e l salt by spraying salt at a r a t e of 50 gpm i n t o t h e cover gas above t h e l i q u i d i n t h e f u e l pump tank.

There they trans-

f e r r e d from t h e l i q u i d t o t h e gas phase and were swept out of t h e tank by a small purge of helium.

After a delay of about 1-1/2 h r i n t h e piping,

t h i s gas passed through water-cooled beds of activated charcoal.

The

krypton and xenon were delayed u n t i l all but t h e 8 5 K r decayed and then were d i l u t e d w i t h a i r and discharged t o t h e atmosphere. The f u e l and coolant systems were provided with equipment f o r taking samples of t h e molten salt while t h e reactor w a s operating at power.

The

fuel sampler was a l s o used f o r adding s m a l l amounts of f u e l t o t h e reactor

while at power t o compensate f o r burnup. The negative temperature c o e f f i c i e n t of r e a c t i v i t y of t h e f u e l and graphite moderator made nuclear control of t h e system very simple.

However,

t h r e e control rods were provided f o r adjusting temperature, compensating f o r buildup of f i s s i o n products, and f o r shutdown. The p l a n t w a s provided with a simple processing f a c i l i t y f o r t r e a t i n g

f'ull 75-ft,3 batches of h e 1 salt with hydrogen f l u o r i d e o r f l u o r i n e gases. The hydrogen f l u o r i d e treatment was f o r removing oxide contamination from

t h e salt as H20.

The fluorine treatment u t i l i z e d t h e f l u o r i d e v o l a t i l i t y

process f o r removing t h e uranium as uF6. Auxiliary systems included:

(1)a cover-gas system w i t h t r e a t i n g

s t a t i o n s f o r removing oxygen and moisture from t h e helium cover gas; (2) two closed-loop o i l systems f o r cooling t h e f u e l andd.coolant pumps

and providing l u b r i c a t i o n t o t h e bearings; ( 3 ) a closed loop component coolant system f o r cooling t h e control rods and other i n - c e l l components;

( 4 ) s e v e r a l cooling water systems including a closed-loop t r e a t e d water system f o r cooling c e r t a i n i n - c e l l equipment; (5) a v e n t i l a t i o n system f o r contamination control; and (6) an instrument a i r system. A l l of t h e primary salt system w a s located i n t h e r e a c t o r and drain

tank c e l l s .

These s e a l e d pressure vessels provided secondary containment.

For a fuller description of t h e p l a n t , see Reference 15.

CHRONOLOGY OF OPERATION AND MAINTENANCE . R. H. Guymon

Design of t h e MSRE began i n t h e summer of 1960 and by August 1964, i n s t a l l a t i o n w a s f a r enough along t o permit t h e planned non-nuclear t e s t i n g 1 6 t o begin.

Milestones t h a t were passed i n t h e years which fol-

lowed a r e l i s t e d i n Table 3-1. Table 3-1.

Milestones i n MSRE Operation

October 24, 1964

S a l t f i r s t loaded i n t o tanks S a l t first c i r c u l a t e d through core First criticality (

'U)

F i r s t operation i n megawatt range

Full power reached

Nuclear operation w i t h

' U concluded

S t r i p uranium from f u e l s a l t F i r s t c r i t i c a l with 233U Reach f u l l power with 2 3 3 U Nuclear operation concluded

1965 June 1, 1965 January 24, 1966 May 23, 1966 March 26, 1968 August 23-29, 1968 October 2 , 1968 January 28, 1969 December 12, 1969 J ~ ~ U ~ W 12,Y

The a c t i v i t i e s during t h e p e r i o d of non-nuclear t e s t i n g , zero-power

experiments, and preparation f o r power operation a r e outlined i n Fig. 3-1. During t h e prenuclear t e s t i n g , except f o r t h e usual s t a r t u p troubles due t o e a r l y f a i l u r e s and instrument malfunctions, all systems operated w e l l and t h e r e were no unanticipated problems i n handling t h e molten salt.

Some plugging of t h e offgas system valves and f i l t e r s w a s encoun-

tered. b u t t h i s did not seem serious. C r i t i c a l i t y w a s a t t a i n e d by adding UFb-LiF enriching salt t o t h e carrier salt.

More uranium w a s added t o b r i n g t h e f u e l gradually t o t h e

desired operating concentration while t h e control rods were c a l i b r a t e d and r e a c t i v i t y c o e f f i c i e n t s were measured.

kid

A t t h e conclusion of t h e zero-

power nuclear experiments t h e r e a c t o r w a s shut down t o f i n i s h construction of t h e vapor-condensing system, t o carry out some inspection, and t o

M S R E ACTlVlTl ES JULY 4964

INSPECTION AND

DECEMBER 1965

P R E N U C L E A R TESTS C O M P L E T E SYSTEM

PRELlM IN A R Y T E S T I N G

F I N A L PREPARATIONS FOR POWER OPERATION

-- -- - r

FINISH INSTALLATION OF SALT SYSTEMS

INSTALL CORE SAMPLES INSPECT FUEL PUMP HEAT-TREAT CORE VESSEL

INSTALL CONTROL RODS SAMPLER-ENRICHER

LEAKTEST, PURGE Et HEAT SALT SYSTEMS

Ezzzzzm

TEST SECONDARY

TEST AUX. SYSTEMS

TEST TRANSFER, FILL Et DRAIN OPS.

OPERATOR TRAINING

LOAD SALT INTO DRAIN TANKS COOLANT SALT

FLUSH SALT

1 I

= a -

I I

Fig. 3.1

I I

LOAD Et CIRCULATE CARRIER

CIRCULATE C R FL SALTS

CONTAINMENT ADJUST Et MODIFY RADIATOR ENCLOSURE

LOAD U - 2 3 5 INZERO-POWER NUCLEAR EXPERIMENTS

1 I

LOW- POWER (0-50 kw) EX ezzl PTS.

mezzZ2za

FINISH VAPOR-CO~D. SYSTEM MODIFY CELL PENETRATIONS REPLACE RADIATOR DOORS

OPERATOR TRAINING

I, I

I I

MSRE Activities July 1964 -December 1965

â&#x20AC;&#x2122;

make some repairs and modifications, including replacement of t h e heatwarped radiator doors.

The first t e s t of t h e secondary containment w a s

made during t h i s shutdown period. After tests i n t h e kilowatt range showed t h a t t h e dynamics of the system were as expected, t h e approach t o f u l l power w a s s t a r t e d i n January 1966. Plugging i n t h e f u e l offgas system occurred immediately

af%erincreasing t h e power t o one megawatt.

Almost three months were

spent investigating and remedying t h e offgas problem by i n s t a l l a t i o n of

a large, efficient f i l t e r .

As shown i n Fig. 3-2, t h e power ascension w a s

resumed i n April and fulLpower, which w a s limited by t h e capability of t h e heat-removal system, w a s attained i n M a y . High-power operation w a s abruptly h a l t e d i n July when t h e hub and blades of one of t h e main blowers i n t h e heat-removal system broke up. While t h e reactor w a s down, t h e array of graphite and metal specimens was removed f r o m t h e core and a new array i n s t a l l e d .

During the flushing

operations before t h e specimen removal, t h e fuel-pump bowl w a s accidentally over-filled and some flush salt froze i n t h e attached gas l i n e s .

Tempo-

rary heaters were i n s t a l l e d remotely t o c l e a r t h e l i n e s and i n November t h e offgas l i n e at t h e pump bowl w a s cleared by running a t o o l through it. Most of t h e remaining operation with 235Uw a s at power.

Restrictions

continued t o develop i n t h e offgas system but caused l i t t l e delay i n t h e program.

Shutdowns were necessary t o make repairs on t h e f u e l sampler-

enricher, t h e component coolant pumps, and in-cell air l i n e disconnects of high c e l l leak r a t e . The core whose leakage caused an specimens were replace and various experiments conducted.

nated on March 26, 1968 s o

ly, annual containment t e s t s were made, Operation with 235U f u e l w a s termie uranium could be s

ed from t h e

f i e 1 salt and replaced w i t During t h e shutdown.

and t h e f u e l piping and vesse troscopy t o determine t h e dis

lowed, t h e core specimens were replaced,

e surveyed using remote ganrma ray specion of f i s s i o n products. Necessary and

preventive maintenance w a s a l s o done i n preparation f o r operation w i t h 2 3 3 ~fuel.

A f t e r t e s t i n g and extensive modification of t h e excess fluorine d i s posal system, t h e flush salt and f u e l s a l t were processed t o s t r i p the

SALT

SALT N NEL UMI

Nn Loop UfNAMKS TESPS

866 J-

S 0

N V E m P T E GAS NFUELLDU'

J (967

_] FUEL

2 4 6 8 WWER (Mw)

INMSTIGATE COMR GAS. XENON. AND FWON PROWCT BEHAVlOA ADD PLUTONIUM R R A D I A E MCdPSUITED U

MAP EP. aposlm WlTH W M P , SPECTROMETER

MEASURE TRITIUM, SAMPLE FUEL REMOVE CORE WRAY PUTREACTORINSTANDBY

-0

Fig. 3 . 2

Outline of the Four Years of MSRE Power Operation

uranium and remove t h e corrosion products produced duri tion."

the fluorina-

Loading of 233Ui n t o t h e c a r r i e r salt, throu& s p e c i a l equipment

attached t o a drain tank began immediately thereafter.

The c e i l s were

closed and leak-tested before t h e reactor w a s made c r i t i c a l by s m a l l additions of 233U through t h e sampler-enricher.

On October 8, 1968, Glenn

Seaborg manipulated t h e controls t o r a i s e t h e power t o 100 kw, making the MSRE t h e world's first reactor t o operate on 233U. E a r l y i n t h e zero-power physics experiments, t h e amount of gas en-

t r a i n e d i n t h e c i r c u l a t i n g f u e l had increased from c0.l t o

- 0.5 uol %.

Investigation of this, as w e l l as d i f f i c u l t i e s w i t h t h e sampler-enricher and plugging i n t h e offgas system, delayed the approach t o f u l l power ~

u n t i l January 1969.

As the power w a s increased i n t o t h e megawatt range, there were observed f o r t h e first time sporadic s m a l l increases ( - 5 t o 10%) i n nuclear power f o r a few seconds, occurring w i t h a varying frequency somewhere around 10/hr.

The characteristics of t h e t r a n s i e n t s pointed t o changes

i n gas volume i n t h e f u e l loop and it appeared t h a t they were most l i k-. ely caused by occasi9nal release of some gas t h a t collected i n t h e core. This hypothesis w a s -supported when, l a t e i n February, a variable frequency generator w a s used t o operate t h e f u e l pump at reduced speed.

The gas

entrainment i n t h e f u e l circulatfng loop decreased sharply (from 0.7 t o

cO.1 vol $) snd t h e p e r t y b a t i o n s ceased e n t i r e l y . 1 8 Operatibns continued at various nuclear powers and fuel-pump speeds u n t i l June 1when t h e reactor w a s shut down'to replace t h e core specimens, investigate the d i s t r i b u t i o n of f i s s i o n products i n the primary loop by gamma scanning, remove t h e offgas r e s t r i c t i o m , and t e s t the secondary containment.

For t h e first time one of t h e control rods did not scram

d it w a s replaced during t h e shutdown.

The. remaining months of operation were spent i n various studies of '

t h e behavior of tritium, xenon, and c e r t a i n other f i s s i o n products i n the reactor,

.In November it became evident t h a t s u f f i c i e n t funds would.'not

14 t h e c e l l atmosphere, indicating a leak i n t h e primary system apparently near a freeze valve. After t h e f i n a l shutdown, t h e f a c i l i t y w a s placed i n a standby cond i t i o n t o await post-operation examinations planned f o r early i n t h e next f i s c a l year.lg

The operating crews were disbanded and over t h e next s i x

months most of t h e engineers were reassigned as they completed analyses and reporting of t h e reactor experience.

As of t h i s writing, t h e post-

operation examination has not been accomplished. In t h i s report and elsewhere MSRE run numbers a r e often used t o i d e n t i f y t h e period of operation.

Generally a new run number was assigned

at t h e end of a major shutdown o r when there were s u b s t a n t i a l changes i n t h e purpose o r type of operation.

The s t a r t i n g dates f o r each of the 20

runs are l i s t e d i n Table 3-2.

Table 3-2.

Run No.

S t a r t i n g Datea

Dates of MSRE Runs

Run No.

S t a r t i n g Datea

January 9 , 1965

January 24, 1967

M s y 11, 1965

June 8, 1967

M a y 31, 1965

September 3, 1967

4 5

December 5, 1965

September 19, 1967

February 7, 1966

August 1 4 , 1968

March 26, 1966

15 16

7 8 .9

June 11, 1966

December 10, 1968 January 10, 1969

September 14, 1966

17 18

Movember 6, 1966

July 31, 1969

December 6 , 1966

April 11, 1969 November 20, 1969

%ese are t h e dates on which samples , logs , e t c . , began t o receive new nmibers, and a r e not generally t h e dates of reactor s t a r t u p .

PLANT PERFORMANCE AND STATISTICS

R. H. Guymon

P. N. Haubenreich -"

"They've kept the darned thing running and when they shut down they can get back on the line; you can't knock that!" An unidentified "AEC spokesman", as quoted i n October 1 2 , 1967 Nucleonics Week. '

"So far the.MoZten S a l t Reactor Experiment has operated successfully and has earned a reputation for r e l i u b i t i t y . ' I USAEC &'airman Glenn T. Seaborg a t the ceremony marking the f i r s t operation of a reactor fueled w i t h 233V, October 8 , 1968. The IdSFU3 ran long and it ran w e l l .

It ran long enough with 235U I

f u e l t o a t t a i n t h e o r i g i n a l goals of t h e experiment, then continued through more than a year of added experimenis with 233U f'uel.

These

statements a r e supported by t h e s t a t i s t i c s on t h e o v e r a l l plant performance presented i n t h i s chapter.

Later chapters describe the performance

of individual components and systems which made up t h e plant. ..

4.1

Cumulative S t a t i s t i c s

. t h e MSRE ran are presented Several d i f f e r e n t indications of how .. long i n Table 4-1. The time c r i t i c a l i s a commonly used index f o r r e a c t o r ex.I

periments.

Another b a s i s of comparison with other reactors i s t h e number of equivalent full-power hours, Integrated power (Mw-hrs) is closely related.

(The figures quoted h e r e are based on a Full power of 8.0 Mw,

which i s t h e value i n d i c a t e d by heat balances.20 Analyses of i s o t o p i c ., changes14 i n d i c a t e t h a t f u l l power w a s 7.34 Mw, i n which case t h e tabul a t e d Mw-hrs should be

l i e d by 0.92.)

, w e amounts of time t h a t

salt c i r c u l a t e d i n t h e loops

of i n t e r e s t from t h e standpoint of t h e

demonstration of m a t e r i

i b i l i t y . ' Each pump operated a length of

t i m e equal t o t h e salt

n plus helium c i r c u l a t i o n i n that loop. of various kinds on d i f f e r e n t p a r t s of

The number of t h e r

t h e salt system are l i s t e d i n Table 4-2.

The numbers a r e small i n com-

parison with t h e permissible numbers of cycles except i n t h e case of t h e

Table

4-1.

Accumulated Operating Data

' U Operation

3U Operation

Total

Time C r i t i c a l ( h r s )

11,515

6,140

17,655

Integrated Power (Mw-hrs)&

72 ,441

33,296

105,737

Equivalent Nl-Power Hours

9,005

4,167

13,172

15,042 16,906

6,746 9,170

21,788 26,026

4,046 3,172

3 ,384 1,535

7,430 4,707

20,789 17,444

10,059 9,994

30,848

14 6

S a l t Circulating Time ( h r s )

Fuel Loop

Coolant Loop i

H e l i u m Circulating Time (hrs )

Fuel Loop

Coolant Loop

Time Above 900'F

(hrs)

Fuel Loop

Coolant 'Loop

27,438

F i l l and Drain Cycles Fuel Loop Coolant Loop

, !

$ased

on heat balances w h i c h i n d i c a t e d full power w a s 8.0 Mw.

51 19

(c3

c Table

4.2

MBRE Cumulative

Cycle History

Coolwlt PlmlR 12 ; 19 971 I 1 101 ! Freeze Flames 100. 101. 102' 13 ' 51 I 18 ;QTi Freeze Flames 200. 201 12 : ., 18 ; 97' Penetrations 200. 201 12 Freeze Vdilve 101 13 Freeze Valve 104 '21 Freeze Valve 105 21 23 Freeze Valve â&#x20AC;&#x2122; 106 15 Freeze Valve 107 Freeze Valve 108 16 15 Freeze Valve 109 Freeze Valve 110 8 Freeze Valve 111 6 2 ,FYeeee Valve 112 Freeze Valve 204 12 peeze Valve 206 12 1 4 hrs FD-1 Cooler 7. pp-4hr :_ * These figures are based on the original calculations, If 'they were cycle tests, the usage factors would be 23.04s and 13.37s.

: ,

156

L 1 ! 1 I !,

: 1f 1 1

* 57 . 50" 29 12 20 14 14 li 21 4 4

15 13

62 14 57 44 22 28 30

10 6 2 42 41

based on freeze

flangd

thermal

18 freeze flanges.

Here t h e f u e l flanges experienced 99% of t h e permissible

number of cycles based on early calculations o r 23% of the number permissi-

tad

b l e on t h e b a s i s of extended tests of a prototype flange. In order t o put t h e MSRF: record i n proper perspective, it i s necess a r y t o measure it against those

development.

of other reactors i n a similar stage of

A comparison on t h e b a s i s of equivalent full-power hours is

made i n Table 4-3.

It is not invidious t o s a y t h a t t h e MSRE compares w e l l

with other reactors having t h e same purpose; t h a t i s , t o demonstrate t h e p r a c t i c a l i t y of a reactor concept. 4.2

Availability during Various Periods

The b e s t index of the r e l i a b i l i t y of a plant should be t h e f r a c t i o n of t i m e (over some extended period) that it i s available f o r i t s intended service.

Perhaps inevitably, however, i n experimental plants w h e r e t h e

objectives include more than simply generating power, t h e d e f i n i t i o n of a v a i l a b i l i t y i s not always so clearcut.

Therefore we have presented in.

Table 4-4 an index which, i n p r i n c i p l e , i s ,less s i g n i f i c a n t b u t whose d e f i n i t i o n is q u i t e clear; namely, t h e time t h a t t h e r e a c t o r w a s actually

c r i t i c a l i n each 3-month period during t h e 4 years of parer operation. Percentages of elapsed t i m e are shuwn f o r s e l e c t e d i n t e r v a l s , but these must not be regarded as a measure of r e l i a b i l i t y since t h e r e a c t o r w a s s u b c r i t i c a l much of t h e t i m e because the planned program ,required it. (During shutdowns f o r core specimen removal o r t h e s u b s t i t u t i o n of 2 3 3 ~ f, o r

exmple. )

I n t h e MSRE t e s t program'6 it w a s planned t h a t t h e r e be a period of operation f o r t h e primary purpose of demonstrating plant r e l i a b i l i t y . This phase of t h e program covered t h e last 1 5 months of operation w i t h

235U. Table 4-5 gives a breakdown of t h e time during t h i s period. The reactor w a s c r i t i c a l 80% of the time and t h e a v a i l a b i l i t y of t h e p l a n t , as defined i n t h i s t a b l e , w a s 86%. This amply j u s t i f i e d D r . Seaborg's statement quoted at t h e head of t h i s chapter. Another indication of r e l i a b i l i t y is haw long a plant can be kept continuously on l i n e .

In t h e f i n a l run with 235U, t h e f u e l salt w a s i n

t h e loop continuously f o r j u s t over 6 months, before it w a s drained f o r

Table 4.3

Equivalent N l - P o w e r Hornsa

Produced by Early U. S. Reactors of Several Types

me Aqueous Homogeneous Organic-Cooled

Reactor

First Critical

m-2

12/57

OMRE Piqua

9/57

LMFBR

PWR BWR MSR

4/63 1/66

Interval (years)

EFPH

3,100 ,-

5.6

5,934 5,642

2.6

2/64

6.8

8/62

9/64

2.1

2,661

Peach Bottom

3/66

3.8

10 ,836b

8/51

SRE

4/57

3.4

Hallam

Sodium-Gr aph it e

I-FI'GR

6/63

Operation Terminated

8,140

EBR-1 EBR-2 Fermi

12/62 b b

11.3

11/63 8/63

Shippingport (Core #I)

12/57

2/64

6.2

27,781'

10.5

6.1 6.3

EBWR

12/56

6/67

VBWR

8/57

12/63

6.3

11,164 11,814

MSRE

6/65

12/69

-4.5

13,172

Calculated from thermal Mw-hrs and i n s t a l l e d capacities reported i n USAEC-DRIYT booklet "Operating History of' U. S. Nuclear Power Reactors d a t a from April 1964 Nucleonics. 1969" except Shippingport Co

EFPH quoted i s through 1969.

boperation i s not

ed.

two cores) through 1969 i s 43,400 EFPH.

Total f o r Shippingport

Table 4-4

Time C r i i c a l Each Quarter dur,.ig t h e

Four Years o f Power Operation (1966

Year

Quarter 1 2

1966

3 4

62 1070 413 1221

57%

(235u)

73.9% (1967 1

2045

735 1800

1969

C r i t i c a l Time Elapsed Time

1852 1186 1292 2144

1967

1968

Hours Critical

- 1969)

1375 1054 1176

(d i

Table 4-5

Breakdown of Time during Spstained Operation Phase-

O f MSRE Program (December

1 4 , 1966

Activity o r Condition

Available

Maintenance

Critical Changing specimens Annual tests

Time

26 d 5 d

Sampler l a t c h (8/67)

35 d 3 d 3 d

V i e w i n r e a c t o r c e l l (6/67)

1.1

0.6

6 d 0 -9

Miscellaneous

Total elapsed

Per centage

8934 h r

13 d

Component cooling pump (9/67 )

r 1

26, 1968)

Air-line disconnects (1/67)

Sampler wiring (12/67)

Other

- March

468 d

100.0

2.2

t h e planned processing.

During these 188 days the r e a c t o r w a s counted as

unavailable only 63 h (1.4% of t h e time) while r e p a i r s were being made t o t h e fuel-sampler drive.

It w a s a c t u a l l y c r i t i c a l f o r 97.8% o f t h e time.

The f i n a l phase of t h e MSRE operation, w i t h 233U f u e l , was not aimed

primarily at demonstrating r e l i a b i l i t y , 2 1 but it turned out t h a t t h e availab i l i t y of t h e p l a n t w a s s t i l l remarkably high.

This i s shown i n Table

4-6.

Here available t i m e includes ( i n addition t o t h e c r i t i c a l time) s u b c r i t i c a l i n t e r v a l s during the zero-power experiments, time spent i n charging the pump power supply during variable-speed t e s t s , changing t h e core specimens, gamma-scanning t h e drained salt systems, and experiments on behavior of gas i n f l u s h salt.

4.3

Interruptions of Operations

Figure 4-1 provides a broad v i e w of t h e MSRE operation and m a d o r i n t e r ruptions.

This figure w a s prepared f o r inclusion (together w i t h similar

graphs from many other r e a c t o r s ) i n t h e USAEC's annual presentation t o t h e JCAE and, according t o t h e rules, assigns a b r i e f , descriptive reason f o r each shutdown of 5 days o r more.

A great deal more d e t a i l as t o t h e nature

and cause of these and s h o r t e r i n t e r r u p t i o n s i n operation i s given i n t h e

4 figures and

11 t a b l e s w h i c h follow.

Figure 4-2 covers 1966, the f i r s t y e a r of power operation.

It shows

when salt w a s i n the f u e l loop, when t h e r e a c t o r w a s at power, and assigns

a nwtiber t o each i n t e r r u p t i o n i n e i t h e r .

A number i n a c i r c l e means t h e

i n t e r r u p t i o n w a s due t o some experiment.

A number i n a square means t h a t

t h e i n t e r r u p t i o n w a s necessary because of trouble w i t h some system o r com-

ponent.

Figures 4-3, 4-4, and 4-5 display t h e same kind of information

f o r 1967, 1968, and 1969 respectively. I n Tables 4-7 through 4-10, t h e i n t e r r u p t i o n s i n operation during each year are categorized as t o t h e type of i n t e r r u p t i o n , t h e cause, and other work t h a t w a s accomplished during t h e interruption.

Each interrup-

t i o n l i s t e d i n these four tables i s described more fully i n Tables 4-11 through

4-14.

Table 4-6

Time C r i t i c a l and Time Available During 233U Phase

O f MSRE Program (September 10,

Elapsed

Period

Zero-power expts (9/10/68 - 11/28/68) 1895 Interim (11/28/68 - 1/12/69) 1092 F i r s t Power R u n s (1/12/69

1968

- 6/1/69)

3355 1642

- December 1 2 ,

Hours Crit.

Avail.

1969)

Percent age C r i t . Avail.

649 86

1852

34.2

97.7

110

7.9

10.1

3175

3268 408

94.6

97.4 24.8

- 8/9/69) 0 0 Later Parer R u n s (8/10/69 - 12/12/69) 2974 2230 2971 75.0 99.9 - - Overall (9/10/68 - 12/12/69) 10959 6140 8609 56.0 78.6 S h u t d m (6/1/69)

. .

...

. ..

..,

_ _ . . .... ..

...._.____......I.

.... .. .

_..._..._.._..-.I.

.. .

i .

._... ...... ~

120

110

100

MAJOR SHUTDOWN PERIODS 1. Lowpowertosting. 2. Remove plugged capillary, check valve and filter in offgas system. 3. Identify plugging material md install larger valves and newly designed filter in offplr system. 4. Repair elactrial wiring short in fuel sample enricher. 6. Leak-ten m t w cell. 6. Removeand replace coMample array; replace ona

MSRE Initial Criticality

6/1/85

Critical on 2 3 3 ~

10/2/88

Installedcapacity Reactor Thermal Mw

Power Generation 1-

Thermal Mwh

33,283 I

blower. 7. Clear BM line plugcsursd by salt ovarfill; replace second blown. 8. Ctar salt plug from gas line; repair air nhn in

TOTAL TMwh

,L/

105,737

reactor cell.

9. Replam leaking alr-line d i m n e c n in reactor cell. 10. Remove and replace eocbumple army; annual tom of containmentand controls. 11. Replace fuel sampler drivo; rnriwe sample latch. 12. Remove and replace corr-umple array. Annual tests. Shake down q g a r i n g plant. Psalt torem0v.U. Load U. 13. Mix salt. Rsprir component cooling blown. 14. Repair fud sampler drive. Sewice control-rod drive. 15. Remornand replace cornample array. Replace control rod and d r h . Clar offgm lina. Annual

........

12 1 1 1 . 1 . .

133

taa. 18. Meaure fission product deposition.

17. Conclude nuclear operations 12/12/89.

I I

..A. I 1

2 3 ,

1965

Fig. 4 . 1

Integrated Power and Major Shutdowns of the MSRE

tl3MOd

do01 1 3 n d NI 1 l V S

F 0

' d .

2 0

c C

z D

n m

FUEL

FLUSH

SALT IN FUEL LOOP Loop

POWER (MW)

POWER

'v, 3

-I

...

4 0

3 2

0, u)

0 yc

D I A

z K

3 I

v) v)

3 LL

(MW) ki3MOd

tf3MOd

HSfllJ

do01 13nd NI I l V S

> 0 z

m w

0 rn 0

z 0 <

8-I

FUEL

...

r I

=xI

.I__

29 l b b b 4-7.

IWlWWFTIOm W OPBRLTIOR DURIRG 1W

C ~ U Mand Related Actlvlties

I I -

lo.

rite

r5-11%

2 3 4 5 6 7

1/21

1/22

Y A*

IS 66 -

B -

P Y

1/23 1/23 1/23 1/24 1/25 kyi/:

P-Y

2/16

11 l2

416 4/12 4/14 4/16 4/20 4/20 4/21 4/22 4/25 4/25 4/28 4/28 4/29 4129 51IJ

P-Y

13 14 15 l6

M 19 2O 21 22

23 24 25 26

n 28 29 9 31 32

33 34 35

37 38 39

ho bl

b2 L3

P P U

P P

A Y*

P P

A* Y

P Y

P P-H

P-Y

A* A*

P-1

U PH

51l2 5/16

5/19 5/23 5/25 5/26 5/26 518 5126 5/28 6/14 6/14 6119 6/26 6/n 7/14 7/15 7/17

P P Yff

P-H

Y P-H

A A

0/10

0112

0/16 0131

a/ P

1f.U 1/12

P P A

P-Y

1/15 1/17

i3 54 55 56

P-Y

2/23 2/24

'1 2/25

P P -

P -

*%portable Unscbedulsd Rod &m

A X

P 8 H

Autcoultic Yanrvll primary &Cotrdary

Worked on huing period Follovlng ihe Interruption; Includes Preventiw Walntenancs

P-H

-I

Utut H a p p e d

Period Contaimlent Testa

Samplers

l m e z e Valves

IControl Rods

1nstrumantati01

Air System

Instrument

Electrical and Hcaters

TVA Power outage

Contalmnt & Ventilation

Ccmponent Coolant System

oiigas System 1

Fuel & Coolant S a l t system

~ u m a nError

Check L i s t s

Unplanned, but due to lkpt.

Coolant Drain

Fuel Drain

LcaB &ran

R o d Scram

Radiator or Blowers

x x

s x s s

= x

> >

5imt % p p n d

-1

I -i -

?! 88

a: -

P P

Y Y Y

Y Y

P P

PY P

Y Y Y

Y Y

P P P

8 8

P ll

P P P

Y Y

8 8

Y II

P P

8 8 8

P P

8 8

I 1 1

I 1

, 1 _1

Table 4.11

Description of Interruptions of Operation .

No.

Date

1/54/20

Low-power t e s t s t o c a l i b r a t e nuclear instruments, check r a d i a t i o n l e v e l s , c o l l e c t data f o r CS and noise analysis, obtain r e a c t i v i t y b a l and check out t h e control c i r c u i t s .

1/21

Rod scram due t o accidental shorting of t h e 32-v dc power supply while investigating a No. 3 s a f e t y channel t r i p .

1/22

Rod and load scram t o t e s t c i r c u i t r y .

1/23

Rod scram due t o a voltage sag.

1/23

Power reduction t o attempt t o blow out plugs A offgas l i n e , equalizer l i n e s , and main charcoal' beds.

1/23

1/24

Load scram due t o f a l s e low coolant flow indication.

1/25

Rod scram due t o inadvertent actuation of t h e manual s crgm- switch.

Description and Related A c t i v i t i e s

Rod scram due t o instrument malfunction when switching l i n e a r channel ranges.

Fuel and coolant drain due t o plugging i n t h e offgas system and t o i n s u l a t e t h e coolant flow t r a n s m i t t e r l i n e s . While shut down, t h e r e s t r i c t o r i n t h e equal i z e r l i n e (521) and t h e check valve i n t h e vent l i n e (533) t o t h e a u x i l i a r y charcoal bed were removed. The f i l t e r i n t h e main offgas l i n e w a s replaced a i N t h e pressure control'valve (PCV-522) w a s replaced w i t h a hand valve w i t h l a r g e r p o r t s . Work was done on No. 3 control rod d r i v e , component coolant pump No. 1, r a d i a t o r door seals, and motor generator No. 4. The thermal s h i e l d air lock problem w a s investigated.

2/16

10,

During 1966

Fuel drain due t o a f a i l u r e o f t h e drain tank c e l l space cooler motor. While shut down, t h e i n l e t valves t o t h e charcoal beds were replaced and t h e offgas l i n e at t h e f i e 1 pump w a s reamed out. Rep a i r s were made on a damaged e l e c t r i c a l penetration on t h e sampler-enricher and a le& on r e a c t o r c e l l

. Reportable unscheduled rod scram.

4 .ll (continued)

Table

No.

Date

10 ( c o n ' t )

Description and Related A c t i v i t i e s space cooler No. 2 w a s fixed. A s t a t i c i n v e r t e r w a s i n s t a l l e d t o replace motor generator No. 4. The t r e a t e d water corrosion i n h i b i t o r was changed from potassium t o lithium n i t r i t e - t e t r a b o r a t e .

416

Fuel drain due t o a plug at t h e a u x i l i a r y charcoal bed which turned out t o be a poppet from a drain tank vent check valve.

4/12

Rod scram and load scram t o t e s t performance of t h e rods and r a d i a t o r doors.

4/14

Power reduction t o c o l l e c t data f o r noise analysis.

14% 4/16

Rod scram due t o t h e f u e l pump stopping which w a s caused by a f a l s e low salt l e v e l indication.

4/20

Manual rod scram and load scram t o end t h e 2.5-Mw operation.

16*

4/20

Rod scram due t o a f a l s e s i g n a l on s a f e t y channel No. 1 while s a f e t y channel No. 2 w a s being t e s t e d .

4/21

Power reduction due t o charcoal beds plugging.

18% 4/22

Rod scram, load scram, f u e l drain and coolant drain i n i t i a t e d by a f a u l t y pump pressure r e l a y which t r i p p e d instrument power breakers. Some salt froze i n r a d i a t o r because t h e coolant pump could not be s t a r t e d due t o a l u b r i c a t i n g o i l flow switch failure ( FS-754).

19*

4/25

Rod scram due t o i n v e r t e r f a i l u r e .

20*

4/25

Rod scram and load scram due t o f a l s e low fuel-pump l e v e l i n d i c a t i o n while doing periodic instrument check l i s t s .

4/28

Rod scram and load scram due t o TVA power outage caused by a storm.

22%

4/28

Rod scram and load scram due t o spurious s a f e t y chann e l t r i p s possibly caused by an e l e c t r i c a l storm.

4/29

Manual rod scram and load scram t o check e f f e c t of cold doors on r a d i a t o r temperatures. Shutdown w a s necessary t o r e p a i r t h e l a t c h on t h e samplerenricher.

Reportable unscheduled rod scram.

Table 4 .ll (continued) No.

Date

4/29

Description and Related A c t i v i t i e s

Fuel drain necessitated by a f a i l u r e of t h e samplerenricher cable drive motor. While shut down, t h e b e l t s were tightened on component cooling pump No. 1 and attempts were made t o blow out t h e plugs i n the offgas l i n e s .

5/11

Power reduction due t o offgas l i n e f i l t e r plugging.

26*

5112

Rod scram and load scram due t o operator f a i l i n g t o ins e r t a control c i r c u i t jumper w h i l e doing periodical instrument check l i s t s .

5/16

Power reduction t o increase p i t c h on main bZower ,blades t o increase possible maximum power.

28*

5/19

Rod scram and load scram due t o a TVA power outage. While shut down, measured offgas pressure drop and attempted t o blow out plugs.

29*

5/23

Rod scram caused by instrument malfunction and operator e r r o r . Operator had switched t o servo with a 40째F difference between a c t u a l temperature and temperature demand setpoint.

5/25

Power reduction t o check t h e power c o e f f i c i e n t of reactivity.

5/26

Power reduction t o obtain r e a c t i v i t y balance data.

5 126

Power reduction due t o main blower No. 1 stopping and f a i l i n g t o start.

5/26

Power reduction t o c o l l e c t data f o r anoise analysis.

34%

5/26

5/28

scram and load scram t o i n v e s t i g a t e a false f i r e alarm. Fuel drain due t o a f a i l u r e of component coolant pump No. l w h i l e component coolant pump No. 2 w a s valved o f f . A planned drain had been scheduled due t o t h e high indicated c e l l leak r a t e . This w a s found t o be leak from t h e thermocouple headers i n t o t h e c e l l . While shut down, a 20-psig leak t e s t w a s made

Reportable unscheduled rod scram.

36 Table 4.11 ( continued)

No.

Date

35 (can't)

Description and Related A c t i v i t i e s on t h e c e l l s , repairs were made on both component coolant pumps, t h e e l e c t r i c a l load w a s r e d i s t r i b u t e d and attempts were made t o locate t h e water leak i n t o the cells.

6/14

Power reduction due t o a f a i l u r e i n t h e e l e c t r i c a l supply t o t h e nuclear instruments.

6/14

Power reduction due t o failure of a f l e x i b l e coupling on main blower No. 1.

6/19

Power reduction t o empty t h e overflow tank.

39*

6/20

Rod scram and load scram due t o t h e operator inadvert e n t l y i n s e r t i n g a t e s t source at t h e wrong process radiation detector (RE-528 instead of RE-565) while doing periodic instrument check l i s t s .

40*

6/27

Rod scram, load scram, f u e l drain, and coolant drain due t o t h e e l e c t r i c a l supply t o component coolant pump No. 1 shorting out at t h e penetration.

4 1 '

7/14

Rod scram and load scram due t o a TVA power outage.

42*

7/15

Rod scram and load scram due t o a TVA power outage caused by a storm.

7/17

Power reduction and coolant drain due t o a catastrophic failure of main blower No. 1 hub. The f u e l s a l t w a s drained on 7/24/66. While f i l l i n g t h e f u e l system w i t h flush s a l t , t h e fuel-pump bowl w a s accidentally o v e r f i l l e d and some salt froze i n t h e sampler and offgas l i n e s . Heat had t o be applied remotely t o remove t h e plugs. While shut down, t h e core specimens were replaced, and a new p a r t i c l e t r a p w a s ins t a l l e d i n t h e offgas l i n e . Leaks i n reactor c e l l space cooler No. 1 and i n t h e t r e a t e d water heat exchanger were repaired and t h e thermal s h i e l d degass i n g tank w a s i n s t a l l e d . Both component coolant pumps were repaired and piping w a s i n s t a l l e d t o drain t h e condensedwdter from them. Repairs were made on t h e r a d i a t o r door s e a l s , r a d i a t o r heaters, and t h e main blower motors. A l l t h e e l e c t r i c a l switch gear breakers were calibrated. P r i o r t o s t a r t q p , the f u e l and coolant pumps lube o i l w a s chang&d, t h e periodic check l i s t s were completed and t h e c e l l s were leak-tested a t 10 psig.

Reportable unscheduled rod scram.

4.11 (continued)

Table

Date

No.

Description and Related A c t i v i t i e s

lO/10

Power reduction d u e - t o a false s i g n a l from a process r a d i a t i o n detector (RE-528).

10/10

Power reduction due t o an experiment when t h e fuel-pump pressure was rapidly vented from 1 5 t o 3 psig.

46*

10/12

Rod scram due t o 'inverter trouble.

47*

10/16

Rod scram and load scram due t o f a l s e s i g n a l from a process r a d i a t i o n detector ( RE-528).

10/31.

Power reduction t o i n s t a l l a new r o t o r on main blower No. 1.

10/31

Fuel drain due t o upper offgas l i n e (524) being plugged. While shut down, t h e flow r e s t r i c t o r i n t h i s l i n e w a s replaced and heat w a s applied t o main offgas l i n e at t h e f u e l pump t o unplug it.

11/11

Power reduction t o check main blower vibration.

11/12

Power'reduction t o empty t h e overflow tank.

11/15

Power reduction t o empty t h e overflow tank.

11/17

Power reduction t o attempt t o blow t h e plug out of t h e main offgas l i n e at t h e f u e l pump.

11/20

Fuel drain due t o a high indicated c e l l leak rate which proved t o be due t o leaking a i r l i n e disconllects on i n - c e l l valve operators. Rotameters were i n s t a l l e d on these air l i n e s . While shut dawn, t h e plug i n t h e offgas l i n e at t h e f u e l pump w a s reamed out and r e p a i r s were made nn t h e component coolant pumps and t h e component coolant system pressure control valve (pdcv-960). Heaters were i n s t a l l e d on t h e i n l e t s of main charcoal beds IA and B. The c e l l s were leakt e s t e d at 10 psig.

55'

12/23.

Rod scram and load scram.by accidentally opening a c i r c u i t while i n s t a l l i n g jumpers f o r an experiment.

Reportable unscheduled rod scram.

Table 4.11 (continued) Date

No.

Description and Related A c t i v i t i e s

12/24

Parer reduction caused by a pressure r e l e a s e experiment. The r e a c t i v i t y decreased, t h e regulating rod withdrew t o i t s upper l i m i t and t h e operator f a i l e d t o change the l i m i t .

12/26

Power reduction when venting t h e f u e l pump after an attempt t o blow out t h e plug i n t h e main offgas l i n e . The regulating rod reached i t s upper l i m i t and t h e operator f a i l e d t o change t h e l i m i t .

Reportable unscheduled rod scram.

Table 4.12

Description of Interruptions of Operation

During 1967

No.

Date

1/12

1/14

1/16

1/28

1/30-2/2'

2/5

2/26

Description and Related A c t i v i t i e s d load scram due t o a freeze valve (FV-108) r e l a y t r i p p i n g t h e r e a c t o r out of RUN during an investigation of t h e p l a s t i c keepers on t h e relays. Power reduction t o observe xenon decay rake and obt a i n zero-power r e a c t i v i t y balance data. due t o leakage from t h e i n - c e l l a i r l i n e disconnects. The measured leakage w a s s o high t h a t reasonable c e l l leak rate calculations were not possible. The disconnects were replaced. While shut down, t h e main offgas f i l t e r and control. valve (PCV-522) were removed and a new f i l t e r w a s i n s t a l l e d . The t r e a t e d water heat exchanger w a s replaced and maintenance w a s done on t h e component coolant pumps. The c e l l s were leak-tested at 10 psig. Power reduction t o t e s t automatic load control. ion t o take pictures of t h e r a d i a t o r and i n v e s t i g a t e he t r a n s f e r of the r a d i a t o r and heat ade dynamic tests and checked t h e failure of a freeze valve pped the r e a c t o r out of on due t o main blower No. 1 vibration. t o obtain heat-transfer data. t o coolant system offgas f i l t e r due t o e l e c t r i c a l power supply t r o u b l e e t y and rod servo.

3/7

t o a bad bearing on main blower at low p y e r t o observe e f f e c t of ' xenon decay on r e a c t i v i t y balance.

Report able unscheduled rod scram.

Table 4.12 ( c o n ' t ) ~~

No.

Date

Description and Related A c t i v i t i e s

3/13

Rod scram and load scram due t o f a l s e s i g n a l on s a f e t y channel 3 while doing periodic instrument, check l i s t s .

3/19

Rod reverse and load scram due t o false low fuel-pump speed indication while doing periodic instrument check lists.

3/23

Power reduction due t o vibration on main blower No. 3.

4/11

Power reduction t o obtain a s p e c i a l fuel-pump gas sample. Also obtained heat-transfer data.

4/28-4/30

Power reduction t o c6Xlect data f o r noise analysis.

515

Power reduction t o t e s t r e a c t o r response t o load changes without rod changes.

518

Rod scram, l o i d scram, f u e l drain and coolant drain t o remove reacto'r core specimens. While shut down, gamma-scan data were taken, a new source w a s ins t a l l e d , and heaters were i n s t a l l e d on t h e i n l e t s t o main charcoal beds 2A and 2B. Reactor c e l l space cooler No. 2 w a s replaced and r e p a i r s were made on control rod No. 2, r a d i a t o r door brakes, component coolant pumps and t h e main blowers. A boot w a s replaced on t h e sampler-enricher. Prior t o startup, t h e periodic check l i s t s were completed, t h e c e l l s w e r e leak-tested at 20 psig.

6/21

Power reduction t o obtain heat-transfer data.

6/23

Power reduction t o r e p a i r an o i l pressure switch on 3 component coolant pump No. 1.

6/25

Rod scram and load scram due t o a TVA power outage caused by a storm. While shut down, r e p a i r s were made on component coolant pump No. 1 o i l pressure switch, a t r e a t e d water rupture d i s c (844) w a s replaced, and a heavier base w a s i n s t a l l e d on main blower No. 3.

22"

6/30

Rod scram and load scram due t o an operator r e s e t t i n g the wrong channel while doing periodic instrument check l i s t s .

Reportable unscheduled rod scram.

Table 4.12 ( c o n ' t )

No.

Date

Description and Related A c t i v i t i e s

23"

7/12

Rod scram and load scram due t o a TVA power outage caused by a storm.

Power reduction t o obtain heat-transfer data.

Power reduction, f u e l drain and coolant drain due t o t h e sampler-enricher drive cable being severed. During shutdown, t h e l a t c h was r e t r i e v e d but t h e capsule remained i n the f u e l pump. P r i o r t o s t a r t u y , maintenance w a s done on component coolant pump No. 1.

Coolant drain due t o t h e r a d i a t o r door binding i n t h e guides.

Power reduction due t o an o i l leak i n component coolant pump No. 2.

Fuel d r a i n due t o a leak through the discharge valve of component coolant pump No. 2 which prevented repair of t h e o i l leak.

10/1

Power reduction t o c o l l e c t data f o r noise analysis.

30*

10120

Rod scram and load scram due t o a power f a i l u r e due t o an a r c between a switch actuator rod and t h e main power l i n e t o t h e building.

10/2 3

Power reduction t o observe t h e e f f e c t of xenon decay on t h e reactivity.balance.

11/14

Power reduction t o c o l l e c t data f o r noise analysis.

11/16

Power reduction t o r e p a i r sampler-enricher wiring.

34%

11/22

Rod scram while maintenance w a s being performed on t h e t rument at b n .

12/13

Power reduction t o observe t h e e f f e c t of xenon poisoni n g on t h e r e a c t i v i t y balance and t o c o l l e c t data f o r noise analysis.

Reportable unschedule

Table 4.13

Description of Interruptions of Operation During 1968

No.

Date

Description and Related A c t i v i t i e s

1/22

Power reduction t o observe t h e e f f e c t of xenon poisoning on t h e r e a c t i v i t y balance and t o c o l l e c t data f o r noise analysis.

2/21

Power reduction t o t e s t t h e rod servo instrumentation under simulated 3U conditions.

2/29

Power reduction t o observe t h e e f f e c t of xenon poisoning on t h e r e a c t i v i t y balance and t o c o l l e c t data f o r noise analysis.

317

3/22

Power reduction t o make dynamics t e s t s .

3/25

Manual rod scram and load scram t o observe xenon poisoning e f f e c t on r e a c t i v i t y balance.

3/25

Manual rod scram t o end 235U operatfon.

3/26

Fuel drain t o end 235U operation. Shutdown w a s emminent due t o a t a n g l e d drive cable i n t h e samplerenricher. A f t e r t h e .drain, t h e capsule w a s accidentally dropped i n t o t h e f u e l pump when t r y i n g t o untangle t h e cable. Unsuccessful attempts were made t o recover t h i s . While shut down, gamma-scan data w a s taken, the core specimens were removed, and t h e main offgas l i n e at t h e f u e l pump w a s unplugged. The flush and f u e l salts were t r a n s f e r r e d t o t h e f u e l processing p l a n t where t h e 235Uw a s removed. 233U w a s added t o t h e f u e l salt. Repairs were made on two heat exchanger h e a t e r s , control rod No. 2, both component coolant pumps, t h e i n v e r t e r , and t h e r a d i a t o r doors. Main blowers No. 1 and No. 3 bearings were replaced and t h e coolant offgas f i l t e r w a s replaced. P r i o r t o s t a r t u p , t h e periodic check l i s t s were completed and t h e c e l l s were l e a k - t e s t e d ' at 20 psig.

9/14

Fuel drain f o r adding more 233U as p a r t of t h e c r i t i c a l i t y experiment.

Power reduction t o observe t h e e f f e c t of xenon poisoni n g on t h e r e a c t i v i t y balance, t o make dynamics tests and take heat-transfer data.

Table 4.13 ( c o n ' t ) Date

No.

Description and Related A c t i v i t i e s

9/17

Fuel drain f o r adding more 233U as p a r t of t h e c r i t i c a l i t y experiment. .

9/21

Fuel d r a i n - f o r adding more 233U as p a r t of t h e c r i t i c a l i t y experiment.

10121

Fuel drain due t o i n v e r t e r failure. Operator f a i l e d t o r e s e t r e a c t o r c e l l r a d i a t i o n monitor i n time t o prevent d r a i n .

13*

11/27

Rod scram and load scram due t o operator f a i l i n g t o ins e r t a jumper around f u e l pump l e v e l contacts. The fuel-pump pressure was being vented a f t e r attempting t o blow out a plug i n t h e main offgas l i n e .

11/28

Fuel d r a i n and coolant drain t o t h e loop and t h e drain tank. f u e l pump w a s reamed o u t , t h e was cleaned, and r e p a i r s were coolant pumps.

12/17

Fuel drain due t o d i f f i c u l t y with the l a t c h and cable drive of t h e sampler-enricher. While shut down, control rod drive No. 3 w a s inspected and a weight added t o decrease i t s drop time.

Reportable unscheduled rod scram.

mix t h e f u e l salt i n The offgas l i n e a t t h e coolant offgas piping made on t h e component

44 Table

No.

4 .I4 Description of Interruptions of Operations During 1969 .'

Date

Description and Related A c t i v i t i e s

1/15

Power reduction t o r e p a i r lower limit switches on r a d i a t o r doors.

1/23

Power reduction t o i n v e s t i g a t e e f f e c t of power on "blips".

1/26

Power reduction t o i n v e s t i g a t e e f f e c t of stopping t h e f u e l pump on "blips".

1/28

Power reduction t o i n v e s t i g a t e "blips".

217

Power reduction t o c o l l e c t d a t a f o r noise analysis.

2/n

Power reduction t o run t e s t s on n a t u r a l convection.

2/16

Power reduction t o run dynamics t e s t s .

2/23

Power reduction due t o failure of b e l t s on s t a c k fan No. 1.

2/27

Power reduction t o connect t h e f u e l pump t o a variable speed motor generator set.

314

Manual rod scram and load scram due t o failure of t h e variable speed motor generator set.

318

Power reduction t o - a l l o w xenon t o decay i n preparation t o making void f r a c t i o n t e s t s . While shut down, t h e coolant offgas f i l t e r w a s replaced.

3/20

Power reduction t o connect t h e f u e l pump t o t h e normal e l e c t r i c a l supply.

3/25

Power reduction t o connect t h e f u e l pump t o t h e variable frequency motor generator set.

14*

3/25

Manual rod scram and load scram due t o failure of t h e variable speed motor generator set.

4/3

Power reduction t o connect t h e f u e l pump t o t h e normal e l e c t r i c a l supply.

ic . Reportable

unscheduled rod scram.

Table 4-14 (continued) ~

No.

Date

Description and Related A c t i v i t i e s

16*

4/10

Rod scram, load scram, f u e l drain, and coolant d r a i n due t o a burned-out fuse on one phase of t h e main transformer. Drains could have been averted by operating a l t e r n a t e equipment.

4/12

Load scram due t o a disturbance i n t h e 48-v dc power supply '

18*

4/15

Rod scram, load scram, and f u e l drain due t o t h e drain freeze valve thawing. The setpoint on a temperature switch had d r i f t e d o f f t h e setpoint and an inadequate plug had been established.

514

Power reduction due t o a f a i l u r e of t h e coolant stack b e r y l l i ~monitor.

515

Power reduct'ion due t o an e r r o r i n calculating t h e r e a c t i v i t y balance.

5/13

Power reduction t o connect t h e f u e l pump t o t h e variable speed motor generator s e t .

5/19

Power reduction t o connect t h e f u e l pump t o t h e normal e l e c t r i c a l supply.

5/25

Power reduction due t o plug i n overflow tank vent l i n e 1:hich caused d i f f i c u l t y i n emptying t h e overflow

tank.

5/26

Power reduction t o connect the fuel pump t o t h e variable speed motor generator s e t .

25*

5/26

Marpu

5/27

scram and load scram due t o f a i l u r e of t h e variable speed motor generator s e t .

scram and load scram due t o f a i l u r e of t h e variable speed motor generator s e t . scram, load scram, f u e l drain, and coolant drain t o remove core specimens and due t o plugging i n t h e offgas system. Control Rod No. 3 d i d not scram. It w a s replaced during t h e shutdown. Extensive gamma-scan data w a s taken. A h e a t e r w a s installea on t h e main offgas l i n e near t h e f u e l pump which aided i n removing t h e plug. The overflow tank

*Reportable

unscheduled rod scram.

46 Table 4.14 (con't) Date

NO.

27 (con't)

Description and Related A c t i v i t i e s offgas l i n e w a s replaced and t h e coolant offgas syst e m w a s unplugged. Repairs were made on t h e samplerenricher, and on a leaky in-cell air l i n e disconnect. Before s t a r t u p , t h e periodic check lists (modified) were completed and t h e c e l l s were leak-tested at 20 psig. N l power w a s delayed while t e s t s were made at various fuel-pump speeds t o i n v e s t i g a t e bubbles i n t h e loop.

8/20

Power reduction due t o f a i l u r e of t h e variable speed motor generator s e t .

8/20

Power reduction due t o f a i l u r e of t h e variable speed motor generator s e t .

8/20

Power reduction due t o f a i l u r e of t h e variable speed motor generator s e t .

8/20

Power reduction due t o f a i l u r e of t h e variable speed motor generator s e t .

8/20

Power reduction due t o failure of t h e variable speed motor generator set^.

8/21

Power reduction due t o f a i l u r e of t h e variable speed motor generator s e t .

8/21

Power reduction due t o f a i l k e of t h e variable speed motor generator set.

8/28

Manual rod scram and load scram due t o f a i l u r e of t h e variable speed motor generator s e t .

36'

8/29

Manual rod scram and load scram due t o f a i l u r e of t h e variable speed motor generator s e t .

8/29

Power reduction f o r xenon decay i n preparation f o r argon cover-gas experiments.

9/7

Power reduction due t o an e r r o r i n c a l c u l a t i n g t h e r e a c t i v i t y balance.

919

Power reduction t o c o l l e c t data f o r noise analysis, t o make gamma scans and run rod drop t e s t s .

9/16

Power reduction t o connect t h e f u e l pump t o t h e normal e l e c t r i c a l supply.

Reportable unscheduled rod scram.

Table 4.14 ( c o n ' t ) No.

Date

Description and Related A c t i v i t i e s

9/18

Power reduction t o replace a bearing on Main Blower No. 1.

9/23

Power reduction t o replace a bearing on Main Blower No. 1. While shut down, connected t h e f u e l pump t o t h e variable speed motor generator s e t .

9/30

Load scram due t o d i r t y contacts on relays.

9/30

Load scram due t o d i r t y contacts on relays.

9/30

Load scram due t o d i r t y contacts on relays.

9/30

Load scram due t o di'rty contacts on relays.

9/30

Load scram due t o d i r t y contacts on relays.

10/1

Power reduction t o connect t h e f u e l pump t o t h e normal e l e c t r i c a l supply.

10/3

Power Yeduction due t o door being inadvertently l e f t open i n an exclusion area which caused inadequate containment v e n t i l a t i o n .

Power reduction t o take a s p e c i a l fuel-pump gas sample. Power reduction t o take s p e c i a l fuel-pump offgas samples and t o run t e s t s with t h e f u e l pump o f f .

51 52

10/23

Load scram due t o a f a l s e t r i p of a coolant s a l t flow relay while doing periodic instrument check l i s t s .

10/24

Load scram w h i l e t e s t i n g relays a f t e r they were repaired. Operator f a i l e d t o r e s e t . t h e relays.

11/2

Manual rod scram, l o a d scram, f u e l drain and coolant drain t o take gamma-scan data,

12/3

Power reduction t o take gamma-scan data.

12/12

Manual rod scram, load scram, f u e l drain, and coolant drain t o end operation of t h e reactor.

48 The next t h r e e t a b l e s summarize t h e foregoing m a s s of information on

Table 4-15 is a summary by type; Table 4-16, by cause and

interruptions.

work done during t h e i n t e r r u p t i o n .

Because unscheduled scrams of t h e con-

t r o l rods are of s p e c i a l i n t e r e s t , a separate breakdown of these i s given

i n Table 4-17.

It should be noted t h a t never were t h e rods scrammed because

of t h e r e a c t o r power, t h e r e a c t o r period, o r t h e f u e l temperature going out of limits.

4.4

Time Required f o r Operational Tasks

One aspect of t h e o p e r a b i l i t y of a plant (and one which a f f e c t s plant a v a i l a b i l i t y ) i s haw long it takes t o do various tasks required i n t h e operaThe purpose of t h i s s e c t i o n i s t o summarize t h e MSRE experience i n

tion.

t h i s regard. Manpower and t h e approximate number of working hours required f o r various t a s k s are l i s t e d i n Table 4-18.

One must recognize t h a t some of

these figures are subject t o considerable variation.

The time required t o

do a p a r t i c u l a r t a s k depended on chance d i f f i c u l t i e s t h a t were encountered,

LiJ

t h e experience of t h e crew, o t h e r jobs being done concurrently and t h e urgency of speedily c o ~ l e t i n gt h e t a s k , i . e . , whether o r not it w a s given a high p r i o r i t y .

Furthermore, sometimes extra requirements such as s p e c i a l

data-taking caused t h e job t o take longer than usual.

The figures i n Table

4-18 apply t o f a i r l y normal s i t u a t i o n s ; not t h e b e s t nor t h e worst.

Very

b r i e f descriptions of the t a s k s a r e given below, w i t h references t o t h e s e c t i o n of t h e MSRE Operating ProceduresP2 followed i n each case, i f applicable.

4.4.1

Auxiliary Systems S t a r t u p Check L i s t s During a shutdown, much of t h e equipment remained i n operation.

However,

a f t e r long shutdowns, t h e a u x i l i a r y systems s t a r t u p check lists (Section 4) were completed t o assure that nothing had been overlooked.

A l l valves and

switches were checked, equipment w a s put i n t o operation and standby equipment

w a s tested.

These were done as l a t e i n t h e shutdown as p r a c t i c a l .

- Electrical

Startup Check L i s t assured t h a t all main breakers were

closed so t h a t equipment and heaters could be started from t h e control boards.

Table 4-15

Summary of Interruptions by Type

Number of Interruptions Type of Interruption Power Reduction: Rod Scram: Load Scram:

Fuel Drain Coolant Drain

Total

1966

Automatic Manual

3 23

Aut omat ic Manual

Automatic Manual

1967” ’

1968

1969

Total

0 23

7 1

3 :.

7 3

14 3

Aut omat ic Manual

Aut omati c Manual

1 3

2 10

225

Table 4-16.

Summary of Interruptions by Cause and

Work Done During Interruption

Number of InterruDtions- Total 1969 .1967 1968 P S W P S W P S W P sw

o o

14 0'

1966 Category Planned

12 0'

o o 58

0' 0 .

Unplanned

0 0

0 0 0

5,0

1 2 0 ' 0 0 0

Human Error

i o

2 3 0 1 1 6 3

0 0 0 1 3 0 0 1 6 0 0 ~

1 1 0

, 1 8 0

Fuel and Coolant S a l t Systems

0 0

0 0 0

Radiator o r Blowers

1 5

4 0 5

0 0 1

2 0 3 1 2 1 1 3

0*12

1 1 2 1 2 1 1 8

o o o

Offgas Bsytem

But Due t o an Ekperimen t Check L i s t s

Water System Component Coolant Syst e m

0 6

Containment and Ventilation

TVA Power Outage

I 3

E l e c t r i c a l and Heaters

0 . 4 .

Instrument A i r System

Instrumentation

0 0 0

1 1 7

4 0 1 0 ' 2

0 0 0 I

0 1

o o

0 0 0 ; o 0

g 30 0 19

1 , 2 1 3 1

I l

Control Rods

0 31 7 0 7 0 2 ' 0 0 2

Freeze Valves

0 0111 0

2io

0O 0 0 O 0

3 ' 2.0 3

0 1 O 0

Samplers

Periodic Containmen t Tests

0 3 f O O 2

0 0

1 l O

Core Specimens

0 0

NOTE:

1 1 2 1 0

0 1 0

5 1 9

l i l 0

= Primary cause of interruption. S = Secondary cause of interruption.

P W

*No

= Worked on during period following the i n t e r r u p t i o n (includes preventive maintenance).

interruption of operations during this e n t i r e period was caused by the cover-gas system, lube-oil system, o r leak-detector system.

Table 4.17

MERE Cumulative Cycle History

ul w

*These

f.igures are based on the original calculations. cycle t e s t s , the usage factors would be 23.04s and 13.37s.

If they were based on freeze flange thermal

. .

. ..

...

....

__...

..............I

... .... ...

Table 4-18.

Operating Procedures Section Number

-~ -~~

" . . . I

. _ ._ ..

l_l

Time t o Complete (hrs)

Operation

- 15

1-T

Instrument a i r system s t a r t u p check l i s t

4c 4D 43

Water system s t a r t u p check l i s t Component cooling system s t a r t u p check l i s t Shield and Containment s t a r t u p check l i s t s Leak test of containment valves Pressurize the c e l l s ( t o 20 p s i g ) Cell l e a k rate a t pressure Evacuate c e l l s and purge with nitrogen F i r s t c e l l leak r a t e a t -2 p s i g V e n t i l a t i o n system s t a r t u p check l i s t h a k d e t e c t o r system s t a r t u p check l i s t Instrumentation s t a r t u p check l i s t

Purging oxygen and moisture from the f u e l c i r c u l a t i o n system

Startup of cover-gas and offgas systems

2 - 4

1-T*

Startup of lube o i l systems

2 - 4

1-T*

Heatup of fuel' and coolant systems

Routine pressure test

4-8 4-8 1 - 2

- 160 15 - 18 >60

130

18 24 >I70 4 - 8

2 - 4 160 - 200 30

- 40 - 50 - 30

1-T* 1-T* H

1-E, 1-T, 1-P *H

w iwt

1-T* 1-T*

1-E, 1-T, 1-1 )Hm

w w

on s h i f t . Most of t h i s *as done on day s h i f t w i t h some a s s i s t a n c e from the s h i f t s ,

1-T*

*Done w

MANPOWER E= Engineer T= Technician P= Pipef itter I= Instrument Mechanic

E l e c t r i c a l s t a r t u p check l i s t

..,.

APPROXIMATE TlMES RE&UIRFD TO PERFORM VARIOUS OPERATIONS

I:\ 81

. " _ ~.I

This d i d n o t r e q u i r e much a t t e n t i o n . This required the a t t e n t i o n of most of the s h i f t operating crew.

c Table 4-18.

APPROXIMATE TIMES REQUIRED TO PERFORM VARIOUS OPERATIONS (continued)

Operating Procedures Section Number

MANPOWER Enaineer T= Technician P= Pipef itter I= Instrument Mechanic

Operation

Time t o Complete (hrs)

F i l l i n g the f u e l and coolant systems

F i l l i n g the c o o l a n t system F i l l i n g the fuel system w i t h flush s a l t Drain of flush s a l t and secure Prepare f o r fuel f i l l

Fuel s a l t sampling o r enriching Coolant salt sampling

12 12

-x-w M

3-5

-x-w

8 - 9

1 - 2 1

3 - 4

2-T* 1-T*

Preparation for c r i t i c a l i t y and power operation S u b c r i t i c a l t o f u l l power

- 12

1-T

P e r i o d i c instrument checks

8A 8B 8D

5 - 7

Neu+xon l e v e l instruments P r o c e s s . r a d i a t i o n monitors Safety ' c i r c u i t checks Emptying the overflow tank

91 12A 12A-2A

m3s F i r s t c o n t r o l room log on each s h i f t

*Done

2 - 4

on s h i f t . Most of this was done on day s h i f t with some a s s i s t a n c e from the shifts.

*This

.x-x-H-

d i d n o t r e q u i r e much a t t e n t i o n .

This required the a t t e n t i o n of most of the s h i f t operating crew.

-1

2-T*

- 1-1/4

1-T

112

3/4

2-I* 3-T* 2 t o 3-T

... .

. . ..

Table 4-18.

Operating Procedure s Section Number

.. .

. .

.. .

_,____I

....

_...__I__...____-...

...

APPROXDlA!l’E TlMES ReQUIRED TO PERFORM VARIOUS OPERATIONS (continued)

MANPOWER Engine e r TI Technician PI Pipef itter I= Instrument Mechanic

E= Time t o Complete (hrs)

Operation

12A

Lags (continued)

12A-2A 12A-2B

12A-2B m-2B

116

Second c o n t r o l room l o g on each s h i f t F i r s t b u i l d i n g log each day F i r s t b u i l d i n g l o g on o t h e r s h i f t s Second b u i l d i n g log on each s h i f t

-- 21

0 - 1

1-T, 1-T* 1-T

1-112 112

113

F u l l power t o s u b c r i t i c a l Fuel s a l t d r a i n and secure Coolant salt d r a i n and secure Cool down t o 400’F

3 - 5

2 - 4 48 60

R e s t a r t i n g equipment after an e l e c t r i c a l power outage

Return to power after a d r a i n Drain to both d r a i n tanks ( t r a n s f e r through t r a n s f e r l i n e s ) Drain t o both d r a i n tanks ( t r a n s f e r through f i l l l i n e s ) Drain t o one d r a i n tank

*Done

- 60 - 30 io - 1 5

50 20

on s h i f t .

Most of this was done on day s h i f t w i t h some a s s i s t a n c e from t h e s h i f t s . w This d i d n o t r e q u i r e much a t t e n t i o n .

This required the a t t e n t i o n of the s h i f t operating crew.

116

R e t u r n t o power after a load and rod scram

- 112

Shutdown of t h e Reactor

l-T* 1-T* l-T* 1-T

1-T*

- Instrument

A i r System Startup Check L i s t assured t h a t t h e corn-

pressors and d r i e r s were operating properly, t h a t t h e standby compressor would automatically s t a r t i f needed, and t h a t emergency nitrogen cylinders were i n s t a l l e d .

- Water

System Startup Check L i s t assured t h a t all pumps and tower

fans were operable and t h a t proper water flow w a s established on all equipment. 4D

- Component

Cooling Systems Startup Check L i s t assured t h a t a l l

valving w a s s e t properly and t h a t all blowers were operable. 4E

- Shield

5 parts.

and Containment Startup Check L i s t i s divided i n t o

P a r t (1)checked t h a t t h e primary and secondary block and check

valves, containment enclosures, and l i n e s were leak-tight t o assure adequate containment.

Removable shielding .was a l s o checked t o assure t h a t it

had been r e i n s t a l l e d .

This check l i s t was assigned t o an operating engi-

neer on day s h i f t with technician and p i p e f i t t e r help as needed. Part ( 2 )

b.i

- this

i s t h e time required t o pressurize t h e c e l l s t o 20 psig

by adding instrument a i r .

Part ( 3 ) involved taking pressure and tempera-

t u r e d a t a periodically t o e s t a b l i s h t h e c e l l leak r a t e . valves and l i n e s were checked while at pressure.

Some containment

The time required f o r

checking these i s included i n P a r t (1). P a r t ( 4 ) i s t h e time required t o vent t h e c e l l s , evacuate t o -2 p s i g , and purge w i t h nitrogen.

a l s o involved taking periodic data.

Part ( 5 )

Since t h e r e w a s a s m a l l water leak i n

the c e l l , it took about 7 days f o r t h e atmosphere t o reach equilibrium. 4F

- Ventilation

Syste

t a r t u p Check L i s t assured t h a t a s t a c k fan

w a s i n operation, t h a t t h e other one would automatically s t a r t , and t h a t

v e n t i l a t i o n w a s adequate i n a l l areas. 4G

- Leak

Detector Systems S t a r t u p Check L i s t assured t h a t a l l valves

were open t o leak-detected flanges and that t h e o v e r a l l leak-rate w a s a t i s f actory

- Instrumentation

S t a r t u p Check L i s t assured t h a t each instrument

and c i r c u i t functioned properly.

Abnormal conditions were simulated and

where possible t h e control action w a s allowed t o occur.

One operating

engineer on day s h i f t was assigned r e s p o n s i b i l i t y f o r completing t h i s check l i s t with technician and instrument mechanic assistance as required.

Some of t h e t e s t s were done on s h i f t . \

4.4.2

Reactor Startup The manipulations involved i n a c t u a l s t a r t u p of t h e r e a c t o r

(Section 5) are described below.

5A -Purging Oxygen and Moisture f r o m t h e Fuel Circulating System involved s e t t i n g t h e helium purge at' maximum rate and operating t h e f u e l pump t o c i r c u l a t e it. The t i m e of purge w a s a calculated number. 5B - S t a r t u p of Cover Gas and Offgas System involved s e t t i n g t h e valves and checking t h e helium t r e a t i n g s t a t i o n .

'5E S t a r t of Lube O i l Systems involved checking t h e supply tank, draining t h e o i l catch tank, s e t t i n g valves and checking t h a t t h e standby pumps would start automatically. 5F

- Heatup of

Fuei and Coolant Systems involved r a i s i n g t h e heater

s e t t i n g s and maintaining a reasonable temperature d i s t r i b u t i o n . and coolant systems could be heated si&taneously.

The f u e l

The t i m e given i s

f o r heating t h e coolant system from room temperature and the r e a c t o r syst e m from 400째F (normal shutdown condition) t o 1200째F and reaching

equilibrium.

5H -Routine Pressure T e s t involved r a i s i n g t h e f u e l system pressure t o 60 p s i g (usually w i t h f l u s h salt c i r c u l a t i n g ) , then lowering t h e f u e l system pressure and increasing t h e coolant system pressure t o 60 psig. A l l pressure switches and t h e i r c o n t r o l o r alarm actions w e r e checked.

51 - F i l l i n g t h e Fuel and Coolant Systems is divided i n t o 4 p a r t s . P a r t (1)involved thawing t h e freeze valves , f i l l i n g t h e coolant loop, checking thaw times of t h e freeze valves, r e f i l l i n g and s t a r t i n g t h e coolant pump.

P a r t ( 2 ) involved thairing t h e freeze valve, f i l l i n g t h e

f u e l loop w i t h f l u s h salt and s t a r t i n g t h e f u e l pump.

P a r t ( 3 ) involved

draining t h e flush s a l t , emptying t h e overflow tank and freezing t h e freeze valve.

P a r t ' , ( 4 ) involved thawing t h e freeze valve, adjusting t e m -

peratures, running rod drop times and various other s a f e t y checks and taking base l i n e countrate data.

P a r t (5) involved f i l l i n g t h e f u e l

loop w i t h f u e l salt, stopping at s i x l e v e l s t o take count r a t e data, freezing t h e drain freeze valve and s t a r t i n g t h e f u e l pump.

5J - C r i t i c a l i t y

and Power Operation i s divided i n t o 2 p a r t s .

Part (1)involved running rod drop and o t h e r tests.

The time does not

include check l i s t s 8A, 8B, o r 8D which sometimes had t o be done before c r i t i c a l i t y (see 4.4.3).

P a r t ( 2 ) involved manipulating t h e rods and

heat-removal system components t o reach f u l l power.

4.4.3

Operation Various jobs were done while t h e r e a c t o r w a s operating.

These are

described below.

6A3 o r 6 ~ 4 Fuel System Sampling o r Addition of Enriching Capsules involved a series of manipulations t o i n s e r t and w i t h d r a w a capsule w h i l e maintaining containment. '

Two samples could be delivered t o the laboratory

per s h i f t i f no difficulties were encountered.

This rate could not be

sustained due t o decontamination of c a r r i e r s , etc.

- Coolant

S a l t Sampling involved a less complicated series of

manipulations t o i n s e r t and w i t h d r a w a sample from t h e coolant pump.

8 - P e r i o d i c Instrument Checks involved t e s t i n g as much of t h e c r i t i c a l equipment as possible without i n t e r r u p t i n g operation of the reactor.

9I - m p t y i n g t h e Overflow Tank involved pressurizing it w i t h helium and venting when it w a s nearly empty.

This took about twice as long when

t h e FP offgas l i n e w a s plugged.

12A -Control Room and

l d i n g Logs were taken twice p e r shift.

Some adjustments were usually required, water treatment w a s added and other odd jobs were done as part of t h e logs.

4.4.4

Shutdown of t h e Reactor ibed below.

P a r t (1)involved reducing t h e power.

It i s divided i n t o

4 parts.

This could be done by scramming

t h e rods and t h e load o r by i n s e r t i n g t h e rods and lowering t h e r a d i a t o r

doors. valves.

P a r t (2) involved draining the fuel system and freezing t h e freeze P a r t ( 3 ) involved

freeze valves.

4.4.5

Part

ning t h e coolant system and freezing t h e

( 4 ) involved codling

t h e systems t o 400'F.

Recovery from Unplanned Shutdowns A nuuiber of times during operation, power outages and o t h e r unex-

pected events caused shutdowns. operation as soon as possible.

Often there w a s a desire t o r e t u r n t o Some of these are described below.

...~

. ..

....

. .~~,~

. . . ...

- -1

ORNL-DWG 73-596 DAYS

30 ,-25 ,-M ,-15 ,-la -5 I

, I

DAYS 8

Repin. modifiutiom. prvsntatiw maintlruncs a d w m specimen mplawmant

Instrumentationstartup c h k list. Contdnmat N r m p duck M8t.

Other sysmm sunup check Iim

Lac ctoum of prinwy systm.

Ruor of Pimvy rymm. Weld nnmbmrn and install Mocks on d k Rarurizr mlh to 20 pig.

--art

d h n 20 pdg.

EVUUM d h md pu01with nitrogen. H n i up fuel md ~001mtlwph

MO~S~UN ruchingquilibium in mllr

Fim (411 I n k rata I( VIEUU~. Fill molmt s m . Fill fuel system with flush salt

& m y e d r n rah

srmpk flush salt Dmin fhnh salt R.pmf a fuel Ylt fill.

Fill fuel system with fuel ok. &mph fuel salt Prepmtion f a criticality and power opsntion. Critiulity md taking tho ramor to full power.

Fig

4.6

Return to Full-Power Operation (Typical Schedule)

1 1 2

1 1 3

Lli4

9A-1 - R e s t a r t

Equipment After a Power Outage involved s t a r t i n g t h e

d i e s e l s , r e s t a r t i n g equipment and r e s e t t i n g t r i p p e d instruments.

55-1-

Return t o Power After a Load and Rod Scram involved manipu-

l a t i n g t h e rods and heat-removal system components t o return t o power.

no instrument o r equipment t e s t s were necessary, t h i s could be done i n l e s s than an hour. l l A , 51, and 5J

3 parts.

- Return t o Power After

a Drain i s divided i n t o

Part (1)assumes a drain t o both drain tanks followed by a

t r a n s f e r through t h e t r a n s f e r l i n e s , r e f i l l and return t o power ( t h i s is t h e design method of t r a n s f e r ) .

P a r t ( 2 ) assumes a drain t o both drain

tanks followed by a t r a n s f e r through t h e f i l l l i n e s , r e f i l l , and r e t u r n t o power.

(This method of t r a n s f e r w a s used t o save time.

interlocks w a s necessary.)

Jumpering of

Part ( 3 ) assumes a drain t o only one drain

tank followed by a f i l l and r e t u r n t o power.

4.4.6

Reactor Startup After a Major Shutdown Each s t a r t u p w a s d i f f e r e n t

requiring scheduling between t h e com-

p l e t i o n of maintenance and start of operation.

The c e l l membrane w a s

usually i n s t a l l e d s h o r t l y a f t e r t h e f i n a l closure of t h e primary system. Assuming t h i s , and t h a t maintenance on out-of-cell

components w a s com-

pleted by t h e time they were needed, a somewhat t y p i c a l s t a r t u p would be

as shown i n Fig. 4-6.

4.5

Changes Made i n t h e Plant

The number and kinds of changes made i n a p l a n t a f t e r it begins operation a r e influenced by two d i f f e r e n t things: changing a c t i v i t i e s .

o r i g i n a l design and

I n some single-purpose systems t h e number of changes

may simply r e f l e c t how well t h e o r i g i n a l design met i t s goals.

I n an ex-

perimental plant such as t h e MSRE, most of t h e changes may be required f o r experimental purposes o r changes i n t h e mode of operation.

I n any event,

t i m e spent i n making changes a f f e c t s t h e a v a i l a b i l i t y of t h e plant. During t h e course of t h e MSRF: operation many changes were made, but only a few caused s i g n i f i c a n t delay i n t h e program.

A f t e r t h e end of construction and t h e non-nuclear checkout of t h e

MSRE, it w a s required t h a t a change request be formally i n i t i a t e d , reviewed and approved before any modification w a s made.

(Section 13 of

Reference 22, which prescribes t h e procedure, defines a modification as tta change i n t h e physical plant which produces a s i g n i f i c a n t l y d i f f e r e n t

c h a r a c t e r i s t i c o r function i n any component o r system.")

I n t h e 55 months

from June 1965 through December 1969, a t o t a l of 633 requests f o r changes i n t h e reactor system were i n i t i a t e d , of which 512 were approved.

For t h e

chemical processing p l a n t , 113 change requests were i n i t i a t e d of w h i c h 87

were approved. Table 4-19 summarizes f o r each 6-month period t h e number of requests i n i t i a t e d and approved f o r t h e reactor.

The table a l s o categorizes t h e

approved changes as t o t h e reason f o r making t h e change and t h e type of

As indicated, most of t h e changes were aimed a t f i l l i n g t h e needs of normal. operation and over h a l f were changes i n e i t h e r instruchange t h a t i t - w a s .

mentation o r setpoints. Table 4-20 summarizes t h e change requests f o r t h e chemical processing

plant. Although w e know how many changes w e r e made, we cannot accurately sum up t h e times required f o r making t h e changes.

We can s a y , however, t h a t i n

t h e case of t h e r e a c t o r change requests, t h e vast majority e i t h e r took

very l i t t l e t i m e t o execute o r w e r e made while other work w a s going on. Exceptions include t h e work on the f u e l offgas system i n the spring of

k966.

I n the-processing plant, t h e summer of 1968 w a s spent i n t e s t i n g

and modifications, p a r t i c u l a r l y of the fluorine disposal system.

c Table 4-19

Number of Requests

Period

1966

1967

1968

1969

Summary of Reactor

Reason for

System Requests

Change Requests

12 $: d

3 2 1965

ti c 0’

lsta

'27

0’

2nd

126

115

1st

173

J-37

104

2nd

103

lst

2nd

-0

1st

2nd

lst

2nd

373

100

234

Total

633 'After

institution

512

of change request

22 procedure

on June 21, 1965.

Suminaryof ChemPlant Change Requests

Table 4-20 Number c Request

Period

iP) 2 g u8

9 8 1965

1966

1967

1968

1969

'otal

1 0

0 2

0 0

5 0

1st 2nd

'25

0 0

1st 2nd

0 0

1 0

113

1st

2nd

1st 2nd 1st 2nd

2 3 8 II

b g 1

0 ,O

0 0

.O 2

0 0

1 1

0 2

1 3

0 0

17 10

1.6

1 0

0 0

1 0

0 0

0 1

0 0

' 0

0 0

1 0

0 0

1 0

0 0

1 0

0 0

0 i;!

t c_.

Type Change

Reason for Change Requests

4.6

Tabulation of Recorded Variables a t F u l l power

It seemed d e s i r a b l e t o record a set of readings of a l l of t h e various

r e a c t o r v a r i a b l e s a t normal conditions.

However, even though t h e MSRE r a n

a t f u l l power much of t h e t i m e , conditions were constantly changing due mainly t o t h e various-experiments which were performed. Also some of

t h e v a r i a b l e s were not recor

select t h e b e s t t i m e .

r e g u l a r l y thus making i t d i f f i c u l t t o

After consideration of these f a c t o r s , t h e 12 t o 8

s h i f t on 10/12/69 was selected as a period a t f u l l power when conditions were f a i r l y normal and considerable information w a s available. Values

were taken from the various recorder charts, r o u t i n e c o n t r o l room and building logs and t h e weekly, monthly, and o t h e r check lists. These are tabulated i n Tables 4.21 and 4.22.

A s n a p s h o t . l i s t i n g a l l of t h e computer

inputs w a s also r e t r i e v e d from the computer tapes f o r 0400 on 10/12/69. The l o c a t i o n of t h e sensing element These values are l i s t e d i n Table 4.23. is described b r i e f l y i n t h e d e s c r i p t i o n column.

The number and l e t t e r s i n

t h e i d e n t i f i c a t i o n column correspond t o those used i n t h e design drawings and o t h e r design documents. They can be used t o f u r t h e r - i d e n t i f y the variable.

64 Table 4.21.

Identification

Tabulation of recorded variables . On 10/12/69 't-

Description

Reading 10/12/69

Line 100 (Reactor butlet) temperature

1205'F 1169'F

TE-101-2A

Line 100 temperature Line 101 temperature

TE-101-3A

Line 101 temperature

1215'F

TE-102-3A

Line 102 temperature

1160'F

TE-102-B4A

Line 102 temperature

1152'F

TE-102-5A

Line 102 (Reactor inlet) temperature

1173'F

TE-103-AlB "E-103-B1

Line 103 temperature Line 103 temperature

908'F 798'F

TE-103-6

Line 103 temperature

1150'F

TE-103-8

Line 103 temperature

TE-103-Bll

Line 103 temperature

1223'F 983'F

TE-103-14B

Line 103 temperature

1200'F

TE-104-5B

Line 104 temperature

918'F

TE-106-5B

Line 106 temperature &ine 108 temperature

2080'F 222'F

TE-109-7 TE-200-B7B

Line 109 temperature

227'F

Line 200 temperature

1018'F

TE-200-D7B

Line 200 temperature

1025'F

TE-200-A8B

Line 200 temperature

1035'F

TE-200-BSB

Line 200 temperature

lOlO'F

TE-200-C8B

Line 200 temperature

1015'F

TE-200-AgB

Line 200 temperature

1010' F

TE-200-BgB

Line 200 temperature

TE-200-14A TE-200-16B

Line 200 temperature Line 200 temperature

1025'F 1018'F

TE-200-19A

Line 200 temperature

1023'F

TE-200-AS-AlB

Line 200 penetration temperature

965'F

TE-200-AS-BlA

Line 200 penetration temperature

642'F

TE-100-A5 TE-100-A6

TE-108-7

1200'F

1015'F

Table 4.21.

Description

Identification

Reading 10/12/69

TE-200-AS-CIA

Line 200 penetration temperature

332°F

TE-2 00-AS-2A

Line Line Line Line Line

200 penetration temperature

363°F

201 temperature 201 temperature

1072°F 1062°F

201 temperature

1068°F

201 temperature

1078°F 1045°F

TE-201-B1OB

Line 201 temperature Line 201 temperature Line 201 temperature

TE-201-C1OB

Line 201 temperature

1086°F

TE-201-A11B

Line Line Line Line

201 temperature

1078°F

201 temperature

1083°F

201 temperature

1075°F

201 penetration temperature

706°F

Line 201 penetration temperature Line 201 penetration temperature

772°F

TE-201-2B TE-201-5A TE-201-7B TE-201-AgB TE-201-BgB TE-201-AlOB

TE-201-CIlB

(continued)

TE-201-D11B TE-201-AS-AIB TE-201-AS-BIA TE-201-AS-C1B TE-2 01-AS-2A TdI-201A Xpr 201 FR-201

1073°F 1059°F

350°F 350°F

Line 201 penetration temperature Radiator AT (salt)

59°F

Radiator power

7.1 MW

Bolant salt flow

860 gpm

TE-2 03-1

870°F

TE-203-2

107°F

TE-204-1A

Line 204 temperature

1135°F

TE-204-2A

204 temper

1155°F

204 tempe

1205°F

204 temper

1200°F

204 temper

1193°F

TE-204-6A

Line Line Line Line Line

204 temp

1170°F

TE-204-A7B

Line 204 temperature

1178°F

TE-204-B7B

Line 204 temperature

1170°F

TE-204-8B

Line 204 temperature

1130°F

TE-2 04-3A TE-204-4A TE-204-5A

66 Table 4.21

Description

Identification

(continued)

Reading 10/12/69

TE-206-lA

Line 206 temperature

1073°F

TE-206-2A

Line 206 temperature

1124°F

TE-206-3A TE-206-4A

Line 206 temperature Line 206 temperature

1147°F 1163°F

TE-206-5B

Line 206 temperature

1150°F

TE-206-6A

Line 206 temperature

1073°F

PI-407

Leak detector pressure

100 psig

RM-500

Cover-gas supply radiation

0.3 mr/hr

PI-500A

Helium header pressure

223 psig

FIC-500A

H e l i u m f l o w rate

6 litershin.

PI-5OOG2

Helium treating station pressure

250 psig

PI-500M

Reduced helium pressure

35 psig

PI-501A

Reduced helium header pressure

PI-51OA

Oil tank No. 2 pressure

PR-51lD

Coolant drain tank pressure

TE-512-1

Line 512 temperature

34.5 psig 7.5 psig 4.8 psig 96'F

FI-512A PI-5 13A

Coolant pump purge gas flow

0.6 liters/min

Oil tank No. 1 pressure

8 Psig

TE-516- 1

.Line 516 temperature

107°F

PR-516

Fuel pump pressure

FI-516B

Fuel pump purge gas flow

5 Psig 2.4 liters/min.

PR-522A

Fuel pump pressure

5.2 psig

TE-524-2

Line 524 temperature

93°F

FI-524B

Fuel pump upper off-gas flow rate

47%

LI-524C

Oil citch tank No. 1 level

FI-526C

Coolant pump upper-off-gas flow rate

LI-526C

Oil catch tank No. 2 level

RTI-528

Coolant gas supply radiation

PR-528A

Coolant pump pressure

14% 0.04 liters/min. 18% 1.5 mr/hr 5.1 psig

(I'

67 Table 4.21

tJ Identification

Reading 10/12/69

Description

A 021 548

Helium oxygen content

0.4 ppm

A HpO I 548 FI-548A ,

H e l i u m moisture content Helium oxygen analyzer flow

2 PPm 100 cc/min

FI-548BI

H e l i u m moisture analyzer flow

100 cc/miil

PdI-556A

Main charcoal bed AP

3.4 p s i

RBI-557

Charcoal beds off-gas radiation

0.1 mr/hr

RM-565

C e l l a i r radiation

1.5 mr/hr

FI-566

Reactor c e l l a i r oxygen analyzer flow

100 cc/min

A01 I 566

Reactor c e l l a i r oxygen content

2.7%

PR-572B

Fuel drain tank No. 1 pressure

5.9 p s i g

PR-574B

Fuel drain tank No, 2 pressure

5.5 psig

PR-576B

Fuel f l u s h tank pressure

PI-589

Werflow tank pressure

5.1 p s i g

FI-589

Overflow tank bubbler flow rate

27 psig

PI-592

Fuel pump pressure

5.4 p s i g

FI-592

(continued)

5.0 psig

25 psig

Fuel pump bubbler flow rate

FI-593

Fuel pump bubbler flow rate

23 p s i g

LR-593C

Fuel pump l e v e l

53%

FI-594

Coolant pump bubbler flow rate

24.8 psig

FI-595

Coolang pump bubbler flow rate

25.5 p s i g

LR-595C

5 7%

FI-596

25.2 p s i g

RM-596

0.1 mr/hr 24.0 p s i g

FI-598

Coolant pump bubbler flow rate

FI-599

Overflow tank

LI-599B

Overflow tank

FI-600

Overflow tank

LI-600B

Overflow tank

24%

PI-7OlA

Fuel 011 pump

10 psig

PI-702A

Fuel o i l pump No. 2 pressure

"E-702-lA

Line 702 temperature

ler flow rate

27.0 p s i g 2 3%

ler flow rate

26.5 psig

64 p s i g 135'F

68 Table 4.21

Identification

(continued) Reading 10/12/69

Description

FI-703

Fuel pump lube oil flow rate

3.8 gpm

FI-704

Fuel pump coolant oil flow rate

8.2 gpm

PI-751A

Coolant oil pump No. 1 pressure

PI-752A

Coolant oil pump No. 2 pressure

8 P a 56 psig

TE-752-lA

Line 752 temperature

FI-753

Coolant pump lube oil flow rate

FI-754

Coolant pump coolant oil flow rate

LI-806A

Steam dome water level (Pp-1)

LI-807A

Steam dome water level (FD-2)

FI-810 FI-812

40 gpm 40 gpm

FI-817

Condenser water flow rate (-1) Condenser water flow rate (-2) Offgas particle trap water flow rate

TI-820-1

OT-1 water outlet temperature

87°F

TI-821-1

OT-1 water inlet temperature

80°F

FI-821A TI-822-1

OT-1 water flow rate OT-2 water outlet temperature

TI-823-1

OT-2 water inlet temperature

8.4 gpm 81°F 77°F

FI-823A

OT-2 water flow rate

TI-826

Treated water cooler (TW out) temperature

RM-827

Process water radiation

TI-829

Treated water cooler (TW in) temperature

PI- 829A FI-830

Treated water pump pressure

FI-832

Coolant pump motor cooling water flow rate

3 mr/hr 108°F 70 psig 4.6 gpm 4.8 gpm

FI-836A

Drain tank cell space cooler water flow rate

63 gpm

FI-838A

Reactor cell space cooler No. 1 water flow rate

FI-840A

Reactor cell space cooler No. 2 water flow rate

FI-844A

Thermal shield water flow rate

53 gpm 59 gpm 50 gpm

PI-85U

Cooling tower water pump pressure

34 psig

FI-85lC

Cooling tower water to cooler flow rate

273 gpm

Fuel pump motor cooling water flow rate

130°F 4.0 gpm 6.7 gpm 0% 0%

8.6 gpm 98°F

69 Table 4.21

Identification

Reading 10/12/69

Treated water cooler (CTW out) temperature Cooling tower water temperature

84°F 80°F

FI-859 FI-862A FI-864A

Thermal shield slide water flow rate Coolant cell space cooler (west) flow rate Coolant cell space cooler (east) flow rate CCP gas cooler water flow rate CCP No. 1 and 2 oil cooler water flow rate Air compresser cooler outlet water temperature Air compresser head outlet water temperature

7 gpm 20 gpm

PI-927C

CCP AP Ventilation filter suction Ventilation stack filters AP Ventilation stack fan suction

PdI-937A PdI-938A

TR to 840 level AP SESA to TR AP

RI4-6000-1 RM-6000-2 RM-6000-3

Reactor cell radiation

PdI-9OOA PI-927A PdI-927B2

RM-6000-4

RM-6000-5 RM-6000-6 RM-6010 RR-8100 RR-8200 FI-9000 PIC-9006-1 FI-9006 PdI-AD2-A2

Lid

Description

TI-854 TI-858

FI-873 FI-875 TI-881-2A TI-881-2B

(continued)

Reactor cell radiation Reactor cell radiation Drain tank cell radiation Drain tank cell radiation Drain tank ce Coolant cell r Reactor power Log reactor p Instrument air f Emergency N2 setpoint Instrument air flow (emergency header) Radiator air pressure drop

21-AD2

Bypass damper position

FI-AD3A

Radiator stack flow rate

TI-AD

Radiator air inlet temperature

TI-AD3- 8A

Radiator air outlet temperature

20 gpm 15 gpm 6.3 gpm 86°F 92°F

8 psi 2.2 in. H a 0 3.3 in. H a 0 5.5 in. H a 0 0.05 in. HzO 0.04 in. HzO 50,000 R/hr 30,000 R/hr 50,000 R/hr 10 R/hr 60 R/hr 500 R/hr 100 R/hr 8.5 MW 7Mw

20% 65 psig 44% 67%

10% 70% 71°F 190°F

70 .L

Tab be 4 21 (continued)

Identification

Description

Reading 10/12/69

PI-AR-1

Instrument air receiver tank pressure

TE-CC-1

Coolant cell ambient temperature

TE-CC-2

Coolant cell ambient temperature

88 psig 88°F 90°F

TE-CC-3

Coolant cell ambient temperature

78°F

TE-CC-4

Coolant cell ambient temperature

80°F

TE-CC-5

Coolant cell ambient temperature

76°F

TE-CC-6

Coolant cell ambient temperature

115°F

TE-CC-7

Coolant cell ambient temperature

105OF

TE-CC-8 TE-CDT-IA

Coolant cell ambient temperature Coolant drain tank top temperature

TE-CDT-3A TE-CDT-5A

Coolant drain tank top temperature Coolant drain tank middle temperature

103°F 1195°F 1182°F 1200°F

WR-CDT-C1

Coolant drain tank weight

EiI COP-1 EiI COP-2

Coolant oil pump No. 1 current Coolant oil pump No. 2 current

0 amps

TE-CP-1B

Coolant pump flange-neck temperature

8.7 amps 245°F

TE-CP-B2A

Coolant pump bowl-neck temperature

786°F

TE-CP-C2A

Coolant pump bowl-neck temperature

795°F

TE-CP- 3B

Coolant pump bowl top temperature

957°F

TE-CP-4A

Coolant pump bowl top temperature

TE-CP-5 A

Coolant pump bowl top temperature

965°F 965°F

TE-CP-6A

Coolant pump bowl bottom temperature

1042°F

TE-CP-7A

Coolant pump bowl bottom temperature

1020°F

TE-CP-8A

Coolant pump bowl flange top temperature

107°F

TE-CPM-1B

Coolant pump motor temperature

100°F

SI-CP

Coolant pump speed

1780 rpm

EiI-CP

Coolant pump current

51 amps

EWI-CP

Coolant pump power

56.5 kW

LR-CPA

Coolant pump level (float)

TE-CPLE-A2

CP level element pot (lower) temperature

4.8 in. 1140°F

71 Table 4.21

Description

Identification

CP level element pot (upper) temperature CP level element pipe (top) temperature

1097°F 1102°F

CP level element sensor (lower) temperature

515°F

TE-CPLE-A7 TE-CR-121 TE-CR-123

CP level element sensor (upper) temperature Coolant radiator inlet pipe temperature Flow venturi pipe temperature

372°F 1060°F 1020°F

LI-DC LI-DTC

Decontamination cell sump level Drain tank cell sump level Fuel drain tank No. 1 (upper) temperature

0 in. 0 in. 1115'F

Fuel drain tank No. 1 (middle) temperature Fuel drain tank No. 1 bayonet (top) temperature

1170°F

TE-FD1-18B TE-FD1-19A TE-FD 1-2OA WR-FD1c TE-FD2-2A TE-FD2-13A

Fuel Fuel Fuel Fuel

drain drain drain drain

tank tank tank tank

No. 1 bayonet (lower) temperature 1161°F No. 1 bayonet (bottom) temperature 1161°F No. 1 weight 0% No. 2 (upper) temperature

Fuel drain tank No. 2 (middle) temperature. No. 2 bayonet (top) temperature

TE-FD 2- 17A TE-FD2-18B

No. 2 bayonet (upper) temperature

TE-FF-102-2 TE-FF-200-2 TE-FF-201-2

1130°F

Fuel drain tank No. 1 bayonet (upper) temperature 1146°F No. 1 bayonet (center) temperature 1158°F Fuel drain t

TE-FD2-16A

TE-FD2-19A TE-FD2-20A WR-FD2c TE-FF-100-2

Reading 10/12/69

TE-CPLE-A4 TE-CPLE-A5 TE-CPLE-A6

TE-FD-1-2A TE-FD1-13A TE-FD1-16A TE-FD1-17A

(continued)

1140°F 1150°F 1115°F 1130°F 1140°F 1140°F 1140°F

Fuel drain tank No. 2 bayonet (center) temperature Fuel drain tank No. 2 bayonet (lower) temperature 0 . 2 bayonet (bottom) temperature Fuel drain tank No. 2 weight 5% 885°F ter temperature 974°F ter temperature ter temperature ter temperature

754°F 806'F

TE-FFT-SA

Fuel flush tank (top) temperature

1132°F

TE-FFT-7B

Fuel flush tank (middle) temperature

1174°F

72 Table 4.21

Identification

. .

(continued)

Description

Reading 10/12/69

WR-FFT-C

Fuel flush tank weight

69.6%

PdI-FI-AI.

Vent. system roughing filters AP

0.6 in. HzO

PdI-FI-A2

Vent. system absolute filters AP

1.6 in. HzO

EiI FOP-1 EiI FOP-2

Fuel oil pump No. 1 current

0 amps

Fuel oil pump No. 2 current

0.2 amps

TE-FP-1B

Fuel pump neck-flange temperature

305°F

TE-FP-2B

Fuel pump neck temperature

559°F

TE-FP-3B

Fuel pump neck-bowl temperature

TE-FP-4B TE-FP-5B

Fuel pump bowl top temperature Fuel pump neck-bowl temperature

962°F 992'F 990°F

TE-FP-6A TE-FP-9B

Fuel pump bowl top temperature

1030°F

Fuel pump neck-bowl temperature

TE-FP-1OB

Fuel pump bowl top temperature

955'F 1002'F

TE-FP-11B TE-FP-12B

Fuel pump flange top temperature Fuel pump bowl center temperature

"E-FPM- 1B

Fuel pump motor temperature

998°F 120°F

SI-FP

Fuel pump speed

1176 rpm

Ei-FP

Fuel pump current

44.5 amps

Ew-FP

Fuel pump power

34 kW

LI-FSC

Fuel storage cell sump level

1.1 in.

FT-2olA-lA

Flow element 20lA pipe temperature

1130'F

FT-20lA-2A

Flow element 20lA top flange temperature

1210'F

FT-2 0lA-3A FT-2 0lA-4A FT-ZOlA-SA FT-2OlA- 6A FT-201B-lA FT-201B-2A FT-201B-3A FT-201B-4A

Flow element 20l.A bottom flange temperature Flow element 20l.A pipe temperature

1160°F 1130'F

Flow element 20lA top flange temperature

1210°F

Flow element 2OlA bottom flange temperature

1250'F

Flow element 201B pipe temperature

970°F

Flow element 201B top flange temperature

1180°F

Flow element 201B bottom flange temperature

1170°F

Flow element 201B pipe temperature

1130°F

150'F

73 Table 4.21

(continued) ,_

Reading 10112169

Flow element 201B top flange temperature

1200°F

FT-201B-6A

Flow element 201B bottom flange temperature

1130°F

TE-FV-103-1B

Freeze valve 103 shoulder temperature

1000°F

TE-FV- 103-2A

Freeze valve 103 center temperature

390°F .

TE-FV-103-3B

540°F

TE-FV-104-3B

Freeze valve 103 shoulder temperature Freeze valve 104 shoulder temperature

TE-FV-104-B4

Freeze valve 104 adjacent pipe temperature

TE-FV-104-5B TE-FV-104-6B

Freeze valve 104 pot temperature Freeze valv pipe temperature

TE-FV-105-2A

Freeze valve

TE-FV-105-A4A

Freeze valve 105 adjacent pipe temperature ' .

1160°F

Freeze valve 105 adjacent pipe temperature

center temperatur e

4500F 450°F 590°F

650°F 1250°F

TE-FV-105-5B

Freeze valve 105 pot temperature

1190'F. 108D.">F .

TE-FV-105-6B TE-FV-106-2A

Freeze valve 105 pipe temperature

1120°F

Freeze valve 106 center temperature

1215°F

TE-FV-106-A4A

Freeze valve 106 adjacent pipe temperature

1140°F

TE-FV-106-B4A

Freeze valve 106 adjacent pipe temperature

1190°F

TE-FV-106-5B TE-FV-107-lA TE-FV-107-3B

Freeze valve 106 pot temperature Freeze valve 107 shoulder temperature oulder temperature Freeze valv adjacent pipe temperature

1120°F

TE-FV-107-5B TE-FV-107-6A Freeze valve

TE-FV-108-3B TE-FV-108-5B

TE-FV-108-6A

500°F 499°F 530°F

ot temperature

610°F

temperature

540'F 440°F

oulder temperature temperature

450°F; 640°F

temperature

543°F

acent pipe temperature

TE-FV-108-B-4

(Ilr

FT-201B-5A

TE-FV-J.O7-A4 '

..-

Description

Identification

. .

TE-FV-lOg-lB

Freeze valve 109 shoulder temperature

TE-FV-109-6A

Freeze valve

TE-FV-204-1B

Freeze.valve 204 shoulder temperature

pot temperature

470°F 595°F 790°F

74 Table 4.21

Identification

(continued)

Description

Reading 10/12/69

TE-FV-204-2A TE-FV-204-3B

Freeze valve 204 center temperature Freeze valve 204 shoulder temperature

235°F 830°F

TE-FV-206-1B

Freeze valve 206 shoulder temperature

710'F

TE-FV-206-2A TE-FV-206-3B

Freeze valve 206 center temperature

300°F

Freeze valve 206 shoulder temperature Heater 103 temperature

830°F 908'F

High bay pressure

-0.19 in. HzO

TE-H 103 PI-HB-A

TE-HX-lA

Heat exchanger coolant out temperature

1060°F

TE-HX-4A

1180°F

TE-EIX-7A

Heat exchanger coolant !Ln temperature Heat exchanger shell (center) temperature

21-ID-A

Inlet radfator door position

82.5%

TE-OFT-1

Overflow tank pipe temperature Overflow tank top temperature

934'F

TE-OFT-2B

1165°F

Overflow tank side temperature

1202°F 1190°F

LI-OTlA3

Outlet radiator door,position Fuel oil supply tank level

79.5% 64%

RM-OT1

Fuel oil supply tank radiation

1.7 mr/hr

LT-OT2A3

58%

RM-OT2 TIC-02 Rl-1

Coolant oil supply tank level Coolant oil supply tank radiation

0.1 mr/hr

Helium oxygen removal No. 1 temperature

790°F

TIC 02 R2-1

Helium oxygen removal No. 2 temperature

1235'F

TIC

R1-2

Helium oxygen removal No. 1 wall temperature

513°F

TIC 02 R2-2 TIC PH 1

Helium oxygen removal No. 2 wall temperature Helium preheater No. 1 temperature

860°F 790'F

TIC PH 2

Helium preheater No. 2 temperature

800°F

TE PT-1

Particle trap temperature

360°F

TE PT-2

Particle trap temperature

360°F

TE PT-3

Particle trap temperature

360'F

TE R-5A

Reactor top temperature

1206°F

TE R-6A

Reactor top temperature

1210°F

TE-OFT-4 ZI-OD-A

Table 4.21

Identification

(continued)

Description

Reading 10/12/69 .

TE R-7A

Reactor neck temperature

TE R-8A

Reactor neck temperature

799°F 680°F

TE R-9 TE R-10

Reactor neck temperature

600°F

Reactor neck temperature

534°F

TE R-17

Reactor side temperature

1190°F

TE R-23A

Reactor side temperature

1185°F

TE R-32A

Reactor bottom temperature

1170°F

TE R-34

Reactor neck flange temperature

233°F

TE R-35

Reactor neck flange temperature

198°F

TE R-36A

ReacGor contr

rod No. 1 (upper) temperature

449°F

TE R-37A

Reactor control rod No,. 2 (upper) temperature

431°F

TE R-38A

Reactor control.rod No. 3 (upper) temperature

TE R-39A

Reactor control rod No. 1 (lower) temperature

460°F 855°F

TE R-40A

Reactor control rod No. 2 (lower) temperature

786°F

TE R-41A TE R-43B

Reactor control rod No. 3 (lower) temperature Reactor graphite tube (lower) temperature

689°F 1025°F

TE R-44A

Reactor neck (bottom) temperature

1218°F

TE R-46A TE R-47

Reactor neck (upper) temperature Reactor neck (upper) temperature Reactor neck (upper) temperature rmal well temperature

250°F

LI RC-C PI RC-A

1 sump level Reactor cell pressure

215°F 212°F 810°F

0 in. -2.5

psig

FI-S1 RM-SlA

75% 100 cpm

RM-SIB RM-s1c LI SC-A

Containment stack iodine

11 cpm 530 cpm 1.0 in. H a 0

LI TC-A

Spare c e l l sump level.

0.4 in. Ha0

PI VT-1

Vapor suppression tank pressure

LI WT-A LI WTC-A

Waste tank level

0 . 4 psig 107 in. H2O

Waste tank cell level

2.6 in. H a 0

76 Table 4.21

Descr3p t ion

Identification

(continued)

Reading 10/12/69

Main blower vibration

e1 mil

OACOT

Official average coolant outlet temperature

1011.5," F

OAFOT

Official average fuel outlet temperature

1208.3'F

Control rod No. 1 position

35.5 in.

Control rod No. 2 position

44 in.

Control rod No. 3 position

43.1 in.

Fission chamber No. 1 position

60 in.

Fission chamber No. 2 position

65.7 in.

Fission chamber No. 1 count rate Fission chamber No. 2 count rate

io4 cps

Control rod No. 1 clutch current

143 ma

Control rod No. 2 clutch current

144 ma

Control rod No. 3 clutch current Motor generator 2 current

147 ma 28 amps

Motor generator 3 current Motor generator 2 voltage

32 amps 52 volts

Motor generator 3 voltage

52 volts

Battery voltage Motor generator No. 1 current

50.5 volts 260 volts 130 amps

Inverter voltage

206 V

Inverter current

89 amps 260 amps 280 amps 0 amps 88 amps 12 cfm

Motor generator No. 1 voltag-

Main blower No. 1 current Main blower No. 3 current Component coolant pump No. 1 current Component coolant pump No. 2 current Instrument air dryer purge rate

104cps

Table 4.22,

Reading 10/12/69

.Description

Reading 10/12/69

Description

Reading 10/12/69

H-CR-1

H203-2

gCH-5

H-CR-2

H204-lA

RCH-6

H-CR-3

LE-CP-1

RCH- 7

H-CR-4

LE-CP-2

H102-2

H-CR-5

FV204-1

2.5

R- 1

H-CR-6

Fvj204-2

R-2

H-CR-7

FV206-1

R-3

H-CR-8

FV206-1A

1.5

HX- 1

200-13

H204-1

Hx-2

201-12

H206-1

Hx-3

202-2

CDT-1

FP-1

200-14

CDT-2

FP-2

200-15

CDT-3

RAN-1

201-10

CP-1

RAN-2

201-11

CP-2

200-16

201-13

H200-1

201-14

202-1

H200-11

102-1

204-2

H200-12

522

205-1

H201-1

102-4

204-3

H201-2

102-5

FT20lAl

H201-9

103

FT201A3

H100-1

FV-103

FT20lA2

RCH-

H-104-1

FT201Ar

RCH-

FV-104-1A

FT-201B1

RCH-3

FV-105-1A

FT201B3

FFT-1

m201B2

FFT-2

FT201B4

H101-2

-1-1

H203-1

H101-3

FD-1-2

. Description *

Tabulation of recorded h e a t e r current on 10/12/69 (average of 3 phases)

‘78 Table 4.22

(continued)

Description

Reading 10/12/69

Description

Reading 10/12/69

Description

Reading 10/12/69

FD-2-1

FV-104-3

FV-108-3

-2-2

FV-104-4

FV-108-1

FV-104-1

FV-105-2

FV-108-3

FV-104-3

FV-105-3

FV-109-1

H-104-5

FV-104-7

FV-109-2

H-104-6

FV-106-2

FV-109-3

FV-105-1

FV-106-3

FV-109-1

FV-105-3

FV-110-2

FV-109-3

FV-105-1

FV-110-3

FV-110-1

FV-105-4

FV-107-1

FV-106-1

FV-107-2

FV-106-3

FV-107-3

FV-106-1

FV-107-1

FV-106-4

FV-107-3

FV-106-IA

FV-108-1

FV-104-2

FV-108-2

Table 4.23. Computer snapshot taken at 0400 on 10/12/69

Identification

Scan Seq.

Description

No. EWM-CP-D

256

Coolant pump power

EWM-F'P-D

255 349

Fuel pump power Reactor cell evacuation flow Radiator stack flow

Fq I-569 FT-AD3-A FT-S1-A FT-20 1-A ET-201-B FT-512-A FT-516-B FT-524-B FT-526-C FT-703-A FT-704-A FT-753-A FT-754-A LE-CP-A

40 242 15 28 236 2 34

Containment stack flow Coolant salt flow Coolant salt flow Coolant pump purge gas flow Fuel pump purge gas flow Fuel pump upper offgas flow

235 237 238

Coolant pump upper offgas flow Fuel pump lube oil flow

239 &O

Fuel pump coolant oil flow lube oil flow Coolant pump coolant oil flow

LT-OT-1-A

241 65 248

LT-OT-2-A LT-5244

250 251 252

Coolant pump level Fuel oil tank level Coolant oil tank level Oil catch tank No. 2 level

LT-595-C LT-596-B LT-598-C LT-599-B LT-600-B PDT-AD2-A PDT-556-A

61 54 62 57 58 24 228

Coolant pump level

Reading 10/12/69 31.8 kW 34.8 kW 1.27 liter/min 195,000 cfm 20,200 cfm 849 gpm 840 gpm 0.65 liter/min 2.39 liter/min 133 cc/min 83 cc/min 3.75 gpm 8.04 gpm 3.87 gpm 6.54 gpm 4.6 in. salt 12.4 in. oil 11.1 in. oil 11.3 in. oil 16.0 in. oil 6.1 in. salt 5.6 in. salt 5.1 in. salt 5.5 in. salt 5.6 in. salt 6.3 in. salt

9.1 in. H20 Main charcoal bed pressure drop

3.4 psi

80 T a b l e 4.23

&continued)

Identification

Scan Seq. No

PDT-960-A

230

Component coolant pump LIP

8.0 p s i

PT-HB-A

233

High bay pressure

-.07

PT-RC-A

348

Reactor c e l l pressure

-1.98 p i g

PT-500-A

223

H e l i u m header pressure

220 p s i g

PT-510

O i l tank No. 2 pressure

6.8 p s i g

PT-511-C

225

Coolant drain tank pressure

-- p s i g

PT-511-D

218

Coolant drain tank pressure

5.7 p s i g

PT-5 13-A

226

O i l tank No. 1 pressure

7.7 p s i g

PT-516

347

Fuel pump pressure

5.0 p s i g

PT-517-A

224

Drain tank supply pressure

PT-522-A

Fuel pump pressure

8.5 p s i g 5.0 p s i g

PT-528

Coolant pump pressure

Description

Reading lo/ 12/69

in. HzO

4.6 p s i g

PT-572-B

219

Fuel drain tank No. 1 pressure

5.2 p s i g

PT-574-B

220

Fuel d r a i n tank No. 2 pressure

5.1 p s i g

PT-576-B

221

Fuel f l u s h tank pressure

5.3 p s i g

PT-589-A

Overflow tank pressure

7.2 p s i g

PT-592-B

Fuel pump pressure

5.6 p s i g

PT-608-B

222

Fuel storage tank pressure

-1.2

RE-NLC1-A

Reactor power

8.6 MW

RE-NLC2-A

Reactor power

8.6 MW

RE-OT-1-B

262

O i l tank No. 1 r a d i a t i o n

1.8 mr/hr

RE-OT-2-B

263

O i l tank No. 2 r a d i a t i o n

0.09 mr/hr

Reactor Power

8.5 MW

RE-SCl-AI.

LiJ

psig

RE-SlA

277

Containment stack alpha

14%scale

RE-S1B

278

Containment stack beta-gamma

7% scale

RE-s1c

270

Containment stack iodine

21% scale

RE-500-D

261

Cover gas supply r a d i a t i o n

0.3 mr/hr

RE-528-B

275

Coolant gas supply r a d i a t i o n

1.5 mr/hr

RE-528-C

276

Coolant gas supply r a d i a t i o n

2.'1 mr/hr

RE-557-A

273

Offgas from charcoal beds r a d i a t i o n

0.1 mr/hr

81 Table 4.23

(continued) ~~

Identification

RE-557-B RE-565-C RE-596-A RE-596-B RE-596-C RE-675-A RE-675-B

283 284 285 286 287

RE-678-C

RE-827-C RM-NCC1-A6 RM-NCC2-A6

264 265 266 259 260

Description

Offgas from charcoal beds radiation Cell Cell Fuel Fuel

air radiation air radiation pump gas supply radiation pump gas supply radiation

Fuel pump gas supply radiation Sampler cold offgas Sampler cold offgas Sampler hot offgas Sampler hot offgas Process water radiation Process water radiation Process water radiation Reactor count rate Reactor count rate

RM-NCC1-A7

Reactor power

RM-NCC2-A7

Reactor power

RM-NCC1-A9

RM-NCC2-A9

SE-CP-G1-A

SE-F"-E l-A TE-AD1-1A TE-AD3-4 TE-AD3-SA TE-AD3-6 TE-AD3-7A

11 184

Reactor period Reactor period speed eed

187 188

TE-AD3-8A

190

TE-CDT-2A

181 182 127

TE-CDT-8

'L,

274 271 272 280 282

RE-565-B

RE-678-D RE-827-A RE-827-B

Scan Seq. No

TE-CP-A2B

185 186

Reading 10/12/69

0.1 mr/hr 1.0 mrlhr 1.4 mr/hr 0.1 mr/hr 0.1 mr/hr 0.1 mr/hr 1.1 mr/hr 4.2 mr/hr 2600 mr/hr 4000 mr/hr 25 mr/hr 26 mr/hr 34 mr/hr 10,000 cps 10,000 cps' 9.6 Mw 9.2 MW . ' -300 sec '

-150 sec

Radiator inlet air temperature Radiator outlet duct wall temperature

1775 rpm 1190 rpm 67°F 107°F

Radiator outlet duct wall temperature tlet duct wall temperature tlet duct wall temperature

120°F 129°F 173°F

tlet air temperature Coolant drain tank bottom temperature Coolant drain tank middle temperature

178°F 1210°F 1203°F

Coolant pump bowl-neck temperature

804°F

82 Table 4.23

(continued)

Identification

scan Seq. No

TE-CP-lA

126

Coolant pump flange-neck temperature

247°F

TE-CP-3A

128

Coolant pump bawl top temperature

955°F

TE-CP-8B

125

Coolant pump flange top temperature

102°F

TE-CP-SA TE-CPM-IA

129 124

Coolant pump bowl middle temperature

1018°F 96°F

TE-CTS-D

Coolant pump motor temperature Surveillance rig top zone temperature

TE-CTS-E

88 87 267 137 138 303 304 305 306 307 309 166 163 164 167 171 168 169 172 106 102 108 109 110

Surveillance rig mid zone temperature Surveillance rig bottom zone temperature

1230°F 1220°F

Diesel house ambient temperature Computer room ambient temperature

76°F 71"F

TE-CTS-F TE-DH-1 TE-DL-1

TE-DL-2 TE-DTC-1 TE-DTC-2 TE-DTC-3 TE-DTC-4 TE-DTC-5 TE-DTC-6 TE-FD1-lA TE-FD1-3A TE-FD1-12A TE-FD1-18A TE-FD2-lA TE-FD2-3A TE-FD2-12A

TE- FD2- 18A TE-FF100-4 TE-FF100-5 TE-FF100-6 TE-FF101-4 TE-FF101-5

Description

Reading 10/12/69

1200°F

Computer reference thermal plane temperature 69°F Drain tank cell ambient temperature

148°F

Drain tank cell ambient temperature Drain tank cell ambient temperature

144°F 151°F

Drain tank cell ambient temperature

145°F

Drain tank cell ambient temperature D r a h tank cell ambient temperature

149°F 150°F

Fuel drain tank No. 1 top temperature

1073°F

Fuel drain tank No. 1 bottom temperature

1149°F

Fuel drain tank No. 1 middle temperature

1178" F

Fuel drain tank No. 1 bayonet temperature

1158°F

Fuel drain tank No. 2 top temperature

1062°F

Fuel drain tank No. 2 bottom temperature

1130°F

Fuel drain tank No. 2 middle temperature

1159°F

Fuel drain tank No. 2 bayonet temperature Freeze flange 100 inner temperature

1138°F 904°F

Freeze flange 100 middle temperature

648°F

Freeze flange 100 outer temperature

559°F

Freeze flange 101 inner temperature

824°F

Freeze flange 101 middle temperature

585°F

83 Table 4.23

(continued)

_. Scan Identification Seq.

Description

Reading 10/12/69

NO TE-FF101-6

111

Freeze flange 101 outer temperature

TE-FF102-4

112

Freeze flange 102 inner temperature

868°F

TE-FF-102-5

113

Freeze.flange 102 middle temperature

631°F

TE-FF-102-6

114

Freeze flange 102 outer temperature

541°F

TE-FF200-4

115

Freeze flange 200 Inner temperature

769°F

TE-FF200-5

116

Freeze flange 200 middle temperature

546°F

TE-FF200-6

117

477°F

TE-FF201-4

118

Freeze flange 200 outer temperature Freeze f4ange 201 inner temperature

TE-FF201-5

120

Freeze flange 201 middle temperature

605°F

TE-FF201-6

121

Freeze flange 201 outer temperature

495°F

TE-FFT-1A

175

Fuel flush tank top temperature

1134°F

TE-FFT-2A

173

Fuel flush tank bottom temperature

1157°F

TE-FFT-10

Fuel flush tank middle temperature

1174°F

TE-FP-1A

174 92

Fuel pump neck-flange temperature

303°F

TE-FP-2A

Fuel pump neck temperature

TE-FP-3A

Fuel pump neck-bowl temperature

518°F 962°F

TE-FP-4A

Fuel pump bowl top temperature

1004°F

TE-FP-SA TE-FP-7B

Fuel pump neck-bowl temperature

984°F

er temperature

454°F

796°F

.1212"F

bot t-om temperature

1208°F

TE-FP-1OA

bowl temperature top temperature

950°F 992°F

TE-FP-11A

e top temperature

146°F

TE-FP-8B TE-FP-9A

TE-FST-10

117°F 81°F

TE-FV103-ZB

407°F

TE-FV104-1B

462°F

r temperature

TE-FPM-lA

center temperature

TE-FV105-2B

1214°F

TE-~~106-2 B TE-FV107-1B

1236°F

144

Freeze valve 107 shoulder temperature

102°F

84 Table 4.23

(continued)

Identification

scan Seq. No.

TE-FV107-2B

145

Freeze valve 107 center temperature

TE-FV107-3B TE-FV108-1B

146

Freeze valve 107 shoulder temperature

490°F 90°F

147 148 149 150 151 152 153 154 155 156 157 158 160 161 162 17 43 176 34 67 91 134 133 132 131 345 105 183

Freeze valve 108 shoulder temperature

77°F

Freeze valve 108 center temperature

446°F

Freeze valve 108 shoulder temperature

76°F

Freeze valve 109 shoulder temperature

103°F

478OF

Tdeeze valve 109 shoulder temperature

108°F

Freeze valve 110 shoulder temperature Freeze valve 110 center temperature

106°F 77°F

Freeze valve 110 shoulder temperature Freeze valve 111 shoulder temperature

88°F 88°F

Freeze valve 111 center temperature

80°F

Freeze valve 111 shoulder temperature

88°F

Freeze valve 112 shoulder temperature

89°F

Freeze valve 112 center temperature

83'F

Freeze valve 112 shoulder temperature

89OF

Freeze valve 204 center temperature Freeze valve 206 center temperature

254°F 303OF

High bay ambient temperature

83°F

HX fuel outlet nozzle temperature

1170°F

HX fuel inlet nozzle temperature HX shell center temperature

1207°F 1185°F

Main blower No. 1 bearing temperature

86°F

Main blower No. 1 bearing temperature

68°F

Main blower No. 3 bearing temperature Main blower No. 3 bearing temperature

93OF 68°F

Nuclear instrument penetration temperature

139°F

Overflow tank bottom temperature

1172°F

Particle trap temperature

203°F

TE-FV108-2B TE-FV108-3B TE-FV10 9-18 TE-FV109-2B TE-FV109-3B TE-FV110-1B TE-FV110-2B TE-FV110-3B TE-FV111-1B TE-FV111-ZB TE-FV111-3B TE-FV112-1B TE-FV112-2B TE-FV112-3B TE-FV204-2B TE-FV206-2B TE-HB-1 TE-HX-2A TE-HX-3A TE-HX-9A TE-MB1-1 TEiMB1-2 TE-MB3-1 TE-MB3-2 TE-NIP-2 TE-OFT-6B TE-PT2YM

Description

Freeze valve 109 center temperature

Reading 10/12/69

able 4.23 Scan Identification Seq. No. TE-PT-2FM

165

TE-PT-2FF

170

Description -

Reading -10/12/69

Particle trap temperature

167°F

Particle trap temperature

94°F

Reactor top temperature

1189'F

TE-R-2

TE-R-4A

TE-R-15A

Reactor side temperature

1181°F

TE-R-18A

Reactor side temperature

1196°F

TE-R-2OA

80 82

Reactor side temper

1182'F

Reactor side temper

1178°F

Reactor bottom temperature

1182'F

Reactor bottom temperature

1182°F

TE-R-26A

(continued)

1209°F

TE-R-2 7A

55 83

TE-R-28A

Reactor bottom temperature

1181°F

TE-R-29A TE-R- 30A

56 85

Reactor bottom temperature

1180°F

TE-R-31A

Reactor Bottom temperature

TE-R-42A TE-R-45A

Reactor graphite tube temperature Reactor neck upper temperature

431°F 282°F

Reactor top temperature,

1185°F . .

Reactor top temperature

1191°F 1185°F

Reactor bottom temperature

1181°F 1180'F

i TE-R-49

TE-R-50 TEkR-51

TE-RC-1

290

ient temperature

139'F

TE-RC-2

291

nt temperature

143'F

TE-RC-3

292 294

ient temperature

149'F 135°F

295

nt temperature

138°F

296

Reactor ce

127°F

TE-RC-7

297

142°F

TE-RC-8

298

Reactor cell ambient temperature Reacto ient temperatur

TE-RC-9

299

Reacto

ient temperature

149°F 139°F

Reactor cell ambient temperature

153°F

Special equipment room ambient temperature

101°F

TE-RC-10 TE- SER-1

301 178

Table 4.23 Scan Identification Seq. NO

(continued)

Description

Reading ; 10/12/69

TE-TRM-1

177

Transmitter room ambient temperature

81°F

TE-VH-1

180

Vent house ambient temperature

76°F

TE-VT-1

143

Vapor tank water temperature

63°F

TE-VT-2

289

Vapor tank air temperature

66°F

TE-100-Al

Line 100 temperature

1207°F

TE-10042

Line 100 temperature

1207°F

TE-100-A3 TE-100-lA

Line 100 temperature

1208°F

Line 100 temperature

1209°F

TE-100- 3A

1208°F

TE-101-SA

Line 100 temperature Line 101 temperature

TE-102-A4A

Line 102 temperature

1167°F

TE-102-lA

Line 102 temperature

1165°F

TE-102-5D

Line 102 temperature Line 200 temperature

1167°F

64 22

Line 200 temperature

1022°F

Line 201 temperature

1068°F

Line 201 temperature

1068°F

TE-201-A2B

20 18

Line 202 temperature

i0ll"F

TE-20l-MC

Line 202 temperature

i0lO"F

TE-201-B11B

123

Line 201 temperature Line 201 temperature

1068°F

TE-200-C7A TE-200-20A TE-2 01-AlB TE-201-Alc

122

1210°F

1014°F

TE-201-1B

TE-202-Al

Line 202 (well) temperature

997°F

TE-202-B1

Line 202 ( w e l l ) temperature

1007OF

TE-202-D1

Line 202 (well) temperature

1007°F

TE-522-1

135

Line 522 temperature

84°F

TE-524-1

136

Line 524 temperature

101°F

TE-556-lA

201

Line 556 temperature

94°F

TE-702-1B

192

Line 702 temperature

132°F

TE-705-lA

193

Line 705 temperature

142°F

TE-707-1A

194

Line 707 temperature

140°F

TE-75 2-1B

195

Line 752 temperature

123°F

1072°F .. '

87 Table 4.23

a-,

(continued)

TE-755-lA

196

Line 755

126°F

TE-575-lA

197

Line 575 temperature

1270F

TE- 791-1

140

Line 791 temperature

102°F

TE-795-1

Line 795 temperature

146°F

TE-804-1

141 215

Line 804 temperature

100°F

TE-805-1

216

Line 805 temperature

104°F

TE-811-1

211

Line 811 temperature

81°F

TE-813-1 TE-826-1

212

79°F

202

Line 813 temperature Line 826 temperature

99°F

TE-831-1

206

Line 831 temperature

102°F

TE-833-1

213

Line 833 temperature

99°F

TE-837-1

203

Line 837 temperature

102°F

TE-841-1 TE-845-1

205

Line 841 temperature

106°F

TE-846-1

207 204

Line 845 temperature Line 846 temperature

124°F 107°F

TE-85 1-1

210

Line 851 temperature

79°F

TE-874-1

208

Line 874 temperature

118°F

TE-876-1

214

Line 87

106OF

TE-916

200

Line

335°F

124°F

TE-917 TE-922

199

Line 922

118°F

WM-CDT WM-FDl WM-FD2

246 243

Coolant d Fuel dra

152 lbs 0 lbs

244

Fuel d

468 lbs

W-FFT

245

Fuel f

8800 lbs

WM-FST

247

Fuel s

XPM-201

268

Reactor

0 lbs 7.6 MW

ZM-FC1

257

Fission ch

ZM-FC2

258

ZM-NCRl

Fission chamber No, 2 position Compensated ion chamber No. 1 position

No. 1 position

61 in. 66 in. 36 in.

Table 4.23 .

Identification

Scan Seq. No.

.'., w

(continued)

Description

Reading 10/12/69

Compensated ion chamber No. 2 positian Compensated ion chamber No. 3 position

44 i n .

ZM-NCR3

21 23

ZT-ID

Radiator i n l e t door position

89 i n .

ZT-OD

Radiator o u t l e t door position

85 i n .

ZM-NCR2

44 i n .

FUEL SYSTEM

J. L. Crowley R. H. Guymon

5.1

C. H. Gabbard J. K. Franzreb

. Description

The fuel-circulating loop consisted of a graphite-moderated r e a c t o r ,

a c e n t r i f u g a l type f u e l pump with an overflow tank, and a shell-and-tube h e a t exchanger, all connected by 5-inch Hastelloy-N piping.

The normal

operating temperature was about 1200°F and f u e l o r f l u s h salt w a s c i r c u l a t e d

at 1200 gpm.

When t h e r e a c t o r w a s not i n operation, t h e s a l t was drained

t o one o r both of t h e f u e l drain tanks o r t o t h e f u e l f l u s h tank.

Inter-

connecting salt piping and freeze valves permit f i l l i n g t h e r e a c t o r o r t r a n s f e r r i n g salt between tanks by manipulating valves i n t h e helium supply and vent l i n e s . 5.2

Purging Moisture and Omgen from t h e System

I n t h e f a l l of 1964, before s a l t was charged i n t o t h e d r a i n tanks, oxygen and moisture were purged from t h e f u e l c i r c u l a t i n g system and t h e drain tanks.

This w a s accomplished by pressurizing t h e system with helium

t o assure t h a t it was l e a k - t i & t and purging with dry h e l i

followed by a combination of evacuation

efore and during t h e heatup.

Details of t h e

. heatup a r e given i n Chapter 17. The helium w a s introduced a t t h e f u e l - pump which was i n operation t o provide c i r c u l a t i o n i n t h e loop. The system w a s vented o r evacuated at t h e normal f u e l pump offgas l i n e (518) o r at t h e provided a longer flow path Figure 5-1 gives t h e i n t h e e f f l u e n t gas were The system w a s evacuated i n g t o 1130’F d i d not c

Later analyses of salt s effective.

a more e f f e c t i v e purge. f operations used.

Peaks of moisture

about 250°F and again at about. 650OF.

r each of -these moi.sture peaks.

Further heat-

s i g n i f i c a n t addikional moisture releases. a t e d t h a t t h e purging’had been very

. .

.E .-cE c

._ 3

‘

c- 4000 5 5 a

750

,....

’A

I -

-1

REACTORVESSEL TEMPERATURE

500

/ ’ ,

250

! A

0 20

26 OCT 1964

Fig. 5 . 1

. . ,

I’

9 NOV 4964

I n i t i a l Removal of Moisture from Fuel System

. _ .

5.3

Fuel-Circulating-System Volume Calibration

After t h e f l u s h s a l t had been added t o t h e d r a i n tanks and t r a n s f e r s made t o f i l l t h e freeze valves (see Section 5.11), system was f i l l e d and operated f o r 8 days.

the f u e l circulating

It was then drained and f i l l e d

s e v e r a l times t o check t h e d r a i n times and c a l i b r a t e the system.

The C a l i b r a t i o n w a s done by increasing t h e d r a i n tank pressure i n increments and recording t h e d r a i n tank weight and t h e d i f f e r e n t i a l pressure between t h e d r a i n tank and t h e f u e l pump.

This i s p l o t t e d i n Fig. 5-2.

The system

w a s purposely o v e r f i l l e d t o determine t h e p o s i t i o n of t h e fuel-pump overflow l i n e and t o test t h e overflow-tank l e v e l indicators.

Overflow oc-

curred a t readings on the two fuel-pump bubblers of 89.5 and 91%. A t o t a l of 105 l b of s a l t t r a n s f e r r e d t o t h e overflow tank produced a reading on t h e l e v e l instrument of 11.5% o r 4.2 inches o f g a l t , i n acceptable agreement with t h e calculated response.

The temperature d i s t r i b u t i o n i n t h e d r a i n valve (FV-103) w a s cont r o l l e d so t h a t it would thaw i n 9 t o l l m i n u t e s a f t e r an emergency d r a i n

signal.

(See Chapter 20.)

The t i m e required t h e r e a f t e r f o r t h e s a l t t o

d r a i n from t h e loop depended on which valves were open i n t h e s a l t and gas lines.

Normally the f r e e z e valves t o both d r a i n tanks w e r e kept thawed,

but f o r a while ' a f t e r a s a l t f i l l one valve would s t i l l be frozen. emergency d r a i n s i g n a l acted t

t h e f r e e z e valves and t o open valves

i n d r a i n tank ven

equalizer l i n e s between t h e gas i n t h e

f u e l pump and i n t h e d r a i n t

t i t was considered possible t h a t one

o r more of these va

en.

t o measure d r a i n t i

On a normal a f t e r FV-103 thaw

Tests were conducted, therefore,

nceivable combinations ..of valving. loop drained completely i n 9-11 min. zer valves open but one f r e e z e valve

kept frozen, t h e d r a i n t i m

about 30 minutes, regardless of whether o r

not t h e vent valv

With t h e equalizers closed, one f r e e z e

valve frozen and

t i m e s were deemed acceptable.

d r a i n time was 41 min.

These d r a i n

ORNL-DWG 73-597

24 U v)

e.20

4a.

a Y

f za

2i 8

6 0

2000

4Ooo

6000

8000

10,m

POUNDS OF SALTTRANSFERRED

Fig. 5.2

Fuel Circulating System Calibration

5.5

Mixing of Fuel and Flush S a l t s

During 235U f u e l operation, t h e uranium.concentration of t h e f l u s h

salt increased an average of about 215 ppm each time it was-used after f u e l s a l t operation.

D u r i ~ ~ g operation, - ~ ~ ~ U w i t h a lower uranium concen-

tration i n the fuel, the

39 ppm.

responding increase should have been only

"he a c t u a l increases during t h e t h r e e flushes a f t e r 233U fuel c i r -

culation were 36, 42, and 39 ppm.

Using these figures, approximately

40 lbs of f u e l salt mixed w i t h t h e flush salt during each flushing opera.tion.

(For more d e t a i l s on s a l t . a n a l y s i s and i n t e r p r e t a t i o n of r e s u l t s ,

see Reference 23.) System Leak During operation t h e c e l l - a i r a c t i v i t y w a s continuously monitored t o detect any leaks from t h e primary system.

No leaks were detected u n t i l after t h e f i n a l fuel salt drain i n December, 1969, A t t h i s time t h e c e l l -

air a c t i v i t y d i d increase which indicated a leak.

Subsequent t e s t s showed

t h a t t h e leak w a s at o r near one of t h e drain-tank freeze valves (FV-105). The a c t i v i t y w a s mostly xenon w i t h some iodine, krypton, and noble metals.

Four days a f t e r t h e first r e l e a s e , t h e r e w a s less than 25 curies of xenon and less than 50 m i l l i c

ne i n t h e c e l l .

t h e atmosphere without s a f e t y limits .24

r e l e a s e rate permitted by t h e MSRE

The p r o b a b i l i t y of

review of t h e operation thermal s t r e s s e s .

u l t i n g from corrosion seems remote.

between t h e air shroud s p e c i f i e d as a f'uI.1 penetr

e freeze valve does not i n d i c a t e any excessive

x-rays and other inspect

This w a s released t o

eports

ere found upon reviewing t h e construction It w a s noted, however, t h a t t h e weld

s a l t piping a t t h e freeze valve w a s not eld.

This l e d t o t h e suspicion that t h e

eak may be a crack t h a t started a t t h i s point and w a s propagated by s t r e s s

cling.

Determinatio

attempted during t h

ocation and nature of examinations.

More i

on t h e preliminary evaluation of t h e leak i s given i n Reference 25.

;

5.7

Oper a t i o n

The operation of t h e f u e l system w a s s a t i s f a c t o r y .

D i f f i c u l t i e s en-

countered with t h e components are described i n the following sections.

The

rate of t r a n s f e r of salt t o the overflow tank w a s higher than expected. Details on t h i s and t h e loop void f r a c t i o n and xenon poisoning are given i n Reference 18.

5.8 5.8.1

Fuel Pump and Overflow Tank

Description The f u e l salt c i r c u l a t i n g pump w a s a centrifugal sump-type pump w i t h

an overhung impeller, developed at ORNL expressly f o r c i r c u l a t i n g molten fluoride salts of t h e type used at t h e MSRE:.26

(See Fig. 5-3 and 5-4.)

At t h e normal operating speed of 1189 rpm, it had an output of about 1250 gpm at a 48.5-ft

salt head.

About 50 gpm of t h e pump output was circulated

i n t e r n a l l y t o t h e pump bowl via a spray r i n g t o promote t h e release of entrained o r dissolved gaseous f i s s i o n products.

The gas space i n

t h e pump bowl w a s purged w i t h helium t o sweep these t o t h e offgas disposal system.

The helium w a s introduced j u s t below t h e s h a f t s e a l i n the bearing

housing.

Most of t h e gas flowed downward through t h e labyrinth between t h e

shaft and the s h i e l d block t o prevent radioactive gas and salt mist from

reaching t h e seal.

The remainder flowed upward t o prevent any o i l which

leaked through t h e seal Prom g e t t i n g i n t o t h e pump bowl.

O i l f o r lubri-

cating t h e bearings and cooling t h e s h i e l d plug w a s recirculated by an external pumping system.

H e l i u m bubbler type instruments w e r e used t o

measure t h e l i q u i d l e v e l as a means of determining t h e inventory of salt i n t h e f u e l system.

Small capsules were lowered i n t o the bowl t o take

samples for analysis o r t o add f u e l salt. The maximum height of t h e l i q u i d i n t h e pump bowl w a s l i m i t e d by the top

of an overflow l i n e (5-1/2 i n . above t h e center l i n e of t h e volute) which connected t o a 5.5-ft3 overflow tank located beneath t h e pump.

H e l i u m bub-

b l e r type instruments were a l s o used f o r measuring t h e l i q u i d l e v e l i n t h e overflow tank.

Since the overflow l i n e extended t o t h e bottom of t h e over-

f l o w tank, closing. a valve i n t h e overflow tank vent l i n e forced t h e salt back t o t h e f u e l pump.

Fig. 5.3

ORNL- LR. m0-56043-0

MSRE Fuel Pump

I-

a I v)

5.8.2

Early Operation The f u e l pump had been loop-tested with molten salt f o r 100 h r s at

1200'F before it w a s i n s t a l l e d a t t h e MSRE i n October, 1964.

A f t e r instal-

l a t i o n no s i g n i f i c a n t difficulties were encountered during helium circulat i o n while t h e f u e l system was being purged, during e a r l y flush salt operat i o n which s t a r t e d on January 12, 1965, o r during the c r i t i c a l i t y experiments. A continuous, but very slow, accumulation of salt i n t h e overflow tank w a s

observed throughout t h e early'operation.

5.8.3

(See 5.8.9.)

Examination of Fuel Pump afier C r i t i c a l i t y ExperimentP7 llhe fuel-pump rotary element was removed f o r inspection i n September 1965

at t h e end of Run 3.

The pump had been used f o r c i r c u l a t i n g helium f o r 1410

hours and it -had been f i l l e d with. salt at .the MSRE f o r 2120 hours, 1895 hours of which t h e s e t w a s being t i o n and appeared rea*

The pump was generally f o r Pull-power operation.

t o be used

The only dimensional change w a s a 0.006-in.

gro.cith

of t h e pump tank bore diameter where t h e upper O-ring mated with t h e pump tanlr. The most s i g n i f i c a n t discovery w a s evidence of a small o i l leak through the gasketed Joint at t h e catch basin f o r t h e lower o i l seal. This o i l had

run down t h e surface of t h e s h i e l d plug, where it had become coked by t h e higher temperatures near t h e bottom.

Some of t h e o i l had reached t h e upper h i e l d plug and had become coked i n t h e aked p a s t t h e O-ring during high-

il w a s observed below t h e r i n g af'ter the decontamination c e l l , but it w a s

t h e open l i n e s during the t r a n s f e r t o J

ions, considerabl c u l t y w a s enn offgas system, due largely t o decomleaked i n t o t h e pump l e , t h e gasketed Joint' w a s seal-welded ver, t h e spar6 never had t o be i n s t a l l e d

at t h e MSRE.

A l a y e r of flush salt about 3/8-inch deep containing about 40 i n . 3

w a s trapped and frozen on top of t h e labyrinth flange.

Apparently t h e

salt had drained through t h e l/&inch diameter holes i n t h e flange u n t i l

surface tension e f f e c t s balanced the hydrostatic head.

This l a y e r can be

seen i n Fig. 5-5, w h i c h i s a photograph taken during the inspection.

This

photograph a l s o shows the contrast between t h e surfaces exposed t o t h e

s a t , which'were b r i g h t but not corroded, and those above the salt, which were d i s t i n c t l y darker.

There w a s a coating of f i n e salt mist p a r t i c l e s

on the lower face of t h e s h i e l d plug and i n an "0"-ring

groove around t h e

s h i e l d plug t h e r e were s m a l l amounts of f l u s h and f u e l salt t h a t must have been transported as mist. The pump was r e i n s t a l l e d using remote maintenance techniques s o t h a t

these techniques could be evaluated.

Four universal j o i n t s on t h e flange

b o l t s t h a t had been found broken during the disassembly were repaired p r i o r t o t h e r e i n s t a l l a t i o n of t h e r o t a r y element.

These failures r e s u l t e d from

excessive b o l t torque t h a t had been used e a r l i e r t o obtain an i n i t i a l seal

on t h e flange.

( I t turned out t h a t the jack screws had not been f u l l y

backed o f f before t h e flange b o l t s were tightened.)

5.8.4

Ftmp and Pipe Support Problem

A problem r e l a t e d t o t h e o v e r a l l fuel-pump i n s t a l l a t i o n became evident during the p o s t - c r i t i c a l i t y shutdown i n t h e f a l l of 1965.

The f u e l pump

could move i n t h e horizontal plane, b u t w a s fixed against v e r t i c a l movement. The heat exchanger could move horizontally; t h e heat exchanger support frame

w a s fixed against v e r t i c a l movement at t h e north end, but w a s mounted on spring supports at t h e south end.

The south end of t h e heat exchanger w a s

coupled t o t h e fuel-pump bowl by a s h o r t length of 5-in. Sched.-bO pipe. The heat exchanger w a s supported from below, and t h e pump bawl w a s supported from above.

The connecting piping was supposed t o move upward at t h e heat

exchanger and downward at t h e pump when t h e system w a s heated. Because of the physical arrangement of t h e piping and equipment, s t r e s s ranges i n the piping due t o thermal cycling were calculated t o reach a

maximum of 20,000 p s i , which is acceptable.

Uncertainties e x i s t e d i n

c r i t i c a l p a r t s of t h e heat exchanger, however, p a r t i c u l a r l y i n t h e nozzle

100

where estimates were as high as 125,000 p s i during cycling from150 and The end of t h e exchanger toward t h e pump bowl w a s therefore

130O0F.

mounted on springs.

This should have reduced s t r e s s e s i n t h e nozzle t o t h e

range of 20,000 t o 70,000 p s i . C a r e f u l observations during a heating cycle t o 1200째F, showed t h a t t h e

equipment did not move as expected, however.

Because of t h e complex equip-

ment configuration and t h e inevitable uncertainty of t h e calculations, it

w a s decided t o make s t r a i n gage measurements with t h e equipment cold and moving t h e piping by mechanical means f o r measured distances with measurable loads.

The highest stresses w e r e found t o be i n t h e f i l l e t where t h e

nozzle w a s welded t o the heat exchanger head when a spring force of 2000 pounds w a s exerted t o raise t h e end of t h e heat exchanger 3/16 i n .

The

measured s t r e s s i n t h e f i l l e t w a s 13,000 p s i and w a s a f a c t o r of about

5 greater than t h e stress i n the nearby piping. Dye-check o f t h e nozzle t o t h e head of t h e heat exchanger showed no indication of cracks. The conclusion fromthese tests w a s that t h e mounting arrangement w a s

adequate t o allow t h e system t o go through more than t h e 50 thermal cycles required without a fatigue failure.

The system w a s then put t o use f o r

power operation.28

5.8.5

Effect of Bubbler Flow Rates on Indicated Fuel-Pump Level The fuel-pump l e v e l w a s determined by bubbling helium through t h e salt

and measuring t h e d i f f e r e n t i a l pressure between t h i s l i n e a&

a reference

l i n e w h i c h connected t o t h e gas space of t h e pump (see Fig. 5-6).

The end

of one of t h e bubbler dip tubes (596) w a s 1.874 i n . lower than t h e other

(593). A common reference l i n e (592) w a s used f o r t h e two bubblers.

The

l e v e l readout instrument had a full s c a l e (0-100$) range of 10 inches of The centerline of t h e volute w a s at 35%. Compensation w a s provided

salt.

i n t h e instrumentation f o r changes of density between flush and f u e l salt and f o r differences i n t h e lengths of t h e b d b l e r tubes. Since t h e d/p c e l l s used to indicate l e v e l were located outside t h e reactor c e l l , t h e r e w a s some pressure drop i n t h e l i n e s between these and t h e pump. lines.

The amount of pressure drop w a s dependent upon flows through t h e

T e s t s were made e a r l y i n 1965 t o e s t a b l i s h t h e relationship between

these flow rates and t h e indicated levels.

101

ORNL-DWG 70-5193

Fig. 5.6

Schematic Representation of Fuel Pump Bowl Level and Density Indicators

102

Data f o r t h e upper probe are presented i n Fig. 5-7. t h e lower probe were similar.

The curves f o r

It can be seen t h a t t h e indicated l e v e l de-

creased w i t h increasing pump-bowl reference l e g flow and t h a t t h e indicated l e v e l increased w i t h increased f l o w i n t o t h e dip tubes. l e v e l w a s not changed during t h e tests.

The actual s a l t

Normal operating conditions were

s e t at 25-psig forepressure on all three of t h e bubbler f l o w elements,

5 psig i n t h e fuel-pump cover gas. 5.8.6

Fuel-Pump Level Changes and Limitations Based on t h e recommendations of t h e pump development group at ORNL, the

l e v e l alarm and interlocks were o r i g i n a l l y s e t as shown i n Table 5-1. The differences i n t h e l e v e l a t which the f u e l pump could be s t a r t e d

(64%) and

t h e l e v e l at which t h e pump would-automatically s t o p (55%) w a s

necessary because t h e indicated l e v e l decreased 10 t o 12% immediately upon s t a r t i n g t h e pump.

This w a s due t o f i l l i n g t h e spray r i n g and fountain flow

chamber. The narrow differences between the high and low alarm and control setpoints caused considerable operational d i f f i c u l t i e s .

Prior t o s t a r t i n g the

pump after a fill, t h e average loop temperature could not be accurately determined.

Since t h e f'uel-pump l e v e l changed about 12.4% p e r 100째F change

i n loop temperature, the s e l e c t e d fill point w a s not a l w a y s s a t i s f a c t o r y f o r operation.

I n which case t h e freeze valve had t o be thawed and t h e system

l e v e l adjusted.

During operation t h e reactor o u t l e t temperature w a s usually

held constant when t h e power w a s changed.

Therefore, t h e bulk average tem-

perature changed w i t h power changes and t h i s caused changes i n t h e fuel-pump level.

I n addition t o t h i s , experiments were run which required operation

at d i f f e r e n t reactor o u t l e t temperatures. During load and rod s c r a m t h e system cooled rapidly. Sometimes it w a s necessary t o reheat t h e system before r e s t a r t i n g t h e pump.

The operating l e v e l s were f u r t h e r restri'cted

when it appeared t h a t t h e offgas plugged more rapidly when t h e salt l e v e l w a s above 60 t o 65% and gas entrainment i n t h e f u e l loop seemed t o increase below about 50%. The problem w a s f u r t h e r compounded by t h e changing pumpbowl l e v e l due t o salt t r a n s f e r t o t h e overflow tank.18

103

c, ORNL-DWG 73-598

72 71 70 69

68 67 66 65 Fl-593 (pig) NOTES:

FI-592 and FI-593 flow rates were proportional t o the difference between the pressure upstream o f the flow restrictors and the pressure in the fuel pump. The fuel pump pressure was held constant a t 5 psig during these tests. LR-593C was read from the bottom o f the inked space which was about 1.5% i n width.

Fig. 5.7

Effect of Bubbler Flow on Fuel Pump Level Indicators

. .

104

Table 5-1.

Level

+78%

' 475%

+64%

Original FP Level Interlocks

Action and Reason f o r Interlock Stops f i l l , gives a temperature setback and rod reverse t o prevent o v e r f i l l i n g of t h e Fuel pump (overflow point i s about 90%). Annunciation hnnp cannot be s t a r t e d b e l o w t h i s l e v e l .

This i s t o

prevent cavitation.

. +55%

Annunciation

Pump w i l l s t o p below t h i s level.

+53%

prevent cavitation.

Again, t h i s is t o

105

Tests were therefore performed whereby t h e interlocks were bypassed and t h e pump w a s started and operated at lower l e v e l s .

Since t h e r e appeared t o be no cavitation o r adverse e f f e c t s on t h e pump, t h e interlocks given i n Table 5-1 were changed t o 78%, 75%, 55%, 40%, and 38% l a t e i n 1965. Normal l e v e l during operation w a s s t i l l maintained between 50% and 60% due t o above considerations.

However, recovery after a load and rod scram w a s

much easier. 5.8.7

Coolant Air t o Fuel P m p

I n t h e design of t h e MSRE, it was calculated t h a t t h e upper portion of t h e fuel-pump tank would be subject t o s u b s t a n t i a l heating from f i s s i o n products i n t h e gas space above t h e salt. pump tank would be limited

Since t h e useful l i f e of t h e

thermal-stress considerations at t h e junction

of t h e volute support cylinder w i t h t h e spherical t o p of t h e tank, close control was at first maintained over t h e temperatures i n t h i s region. ,

Design s t u d i e s had indicated t h a t t h e maximum l i f e t i m e would result if the junction temperature were kept about 100째F below t h e temperature on

t h e tank surface 6-in. out from t h e junction.

Component cooling "air"

(95%N 2 ) was provided t o maintain t h i s temperature d i s t r i b u t i o n .

A secondary

consideration i n controlling t h e temperatures w a s a desire t o keep as much of t h e pump tank as PO

l e above t h e liquidus temperature of t h e

salt. I n operating t h e r e a c t o r , it would be i d e a l i f a fixed flow rate of

air over t h e pump

ovide a s a t i s f a c t o r y temperature d i s t r i b u t i o n

f o r all conditions

gn calculations indicated that t h i s condition

could be met with an ai

of 200 c f ~ n . However, ~ ~ temperature measure-

ments on t h e pump-test

d during t h e i n i t i a l heatup of t h e MSRE indi-

cated t h a t only about

ould be required, and t h a t the air would have

t o be turned o f f when t h e

To minimize t h e t e

air f l o w during power

tank w a s empty. e f f e c t s when t h e cooling air w a s turned on,

as t o have been set at t h e m i n i m t h a t

ould give t h e desired temperatures.

It w a s found t h a t an air flow of 30 cfIn

gave a s a t i s f a c t o r y temperature d i s t r i b u t i o n at a l l power l e v e l s .

I n order

t o achieve and control t h i s r e l a t i v e l y low flow rate, a new valve, having

106

a lower Cv had t o be s u b s t i t u t e d f o r t h e o r i g i n a l valve.

L+

Figure 5-8

shows t h e temperatures i n the two regions of i n t e r e s t as a function .of reactor power l e v e l with t h a t air flow.

The v a r i a t i o n s i n t h e individual

temperatures were caused by v a r i a t i o n s i n t h e pump-bowl.leve1 salt tempera-

ture and air flow.

Both the individual temperatures and t h e temperature

differences increased l i n e a r l y w i t h power, as expected.

It w a s f e l t during

e a r l y MSRE operation t h a t , although temperature differences would exceed 100째F at f u l l power, t h e r e a c t o r could be so operated w i t h t h e 30-cfm air Mow without s i g n i f i c a n t l y reducing the l i f e of the pump tank.

When t h e r e a c t o r w a s s t a r t e d up f o r R u n 8 i n Septeniber 1966, an unexplained'shif't downward i n these temperatures w a s noted.

Later t h e cooling

air t o t h e pump tank w a s turned o f f during t h e attempts t o m e l t out t h e salt plug i n t h e 522 l i n e , and although the temperatures on t h e pump-tank surface were higher than w i t h t h e cooling air, t h e temperature gradient w a s less. Since t h e temperature d i s t r i b u t i o n w a s as good o r better than w i t h t h e air cooling, t h e use of air cooling w a s discontinued.

5.8.8

S a l t Transfer t o the Overflow Tank E a r l y operation of t h e r e a c t o r showed t h a t by some unexplained mecha-

n i s m , salt gradually accumulated i n t h e f u e l pump overflow tank even when t h e salt l e v e l i n t h e pump bowl w a s w e l l below t h e overflow point.

The

t r a n s f e r rate depended on salt l e v e l , and t h e t r a n s f e r ceased when t h e l e v e l w a s about 3 i n . below the overflow point.

This s i t u a t i o n e x i s t e d

u n t i l about A p r i l 1966, when transfer began t o be observed at lower salt levels.

The rate appeared t o increase gradually as t i m e went on u n t i l

it leveled o f f i n June and J u l y a t 0.57 l b of salt per hour, independent of salt l e v e l as far down as 4.7 i n . below t h e overflow point.

The change

occurred at t h e time of t h e stepwise increase i n power, but no mechanism connecting t h e two has been i d e n t i f i e d .

This rate of t r a n s f e r continued

through t h e 235U operation and required emptying t h e overflow tank about t h r e e times per week. P r i o r t o operation w i t h 233U f u e l salt, t h e flush salt which had been processed t o remove t h e 235U was c i r c u l a t e d f o r 42 hours.

During t h i s and

t h e first 16 hours of f u e l c a r r i e r salt c i r c u l a t i o n , t h e t r a n s f e r rate and

â&#x20AC;&#x2DC;C

ORNL-DWG 66-4 4444

800

Fig.

5.8

Variation Power.

4 POWER

(Mw)

of Fuel-Pump Tank Temperatures with Cpoling-air flow, 30 cfm.

Reactor

108

loop void fraction appeared t o be unchanged from t h e previous periods. hours after t h e start of a =-hour

Two

exposure of a beryllium rod i n t h e pump

bowl, t h e bubble fraction i n t h e system increased from t h e normal 0.1 vol

% and t h e rate of t r a n s f e r t o t h e overflow from about 0.4 lb/hr t o greater than 4 lb/hr.

t o about 0.6 vol

tank increased

During t h e remainder of t h e 233U operation t h e t r a n s f e r rate was high

(up t o 72 l b / h r ) and variable.

The results of t h e investigation of t h e

overflow rate, the changing brrbble fraction, and subsequent power perturbations are given i n Referen6e 18.

Burps of t h e Overflow Tank

5.8.9

During early operation of t h e MSRE, when t h e overflow rate was very low and t h e need t o push salt from t h e overflow tank back t o t h e fuel-pump bowl

w a s infrequent, t h e practice w a s t o empty the overflow tank completely.

The

sudden pressurization of t h e f i e 1 pump at t h e end of t h e burp gave false l e v e l indications and stopped t h e pump.

Also when power operation w a s

started, it w a s noted that gaseous f i s s i o n products were being pushed out

the o i l seal l i n e (524) by t h e sudden pressurization.

Therefore, t h e

procedures w e r e changed such t h a t t h e overflow tank w a s not completely emptied of salt during operation.

In February 1969, t h e main offgas l i n e plugged t o t h e point where it presented a 4-psi pressure drop t o t h e normal 3.2 l i t e r / m offgas flow. During a burp of t h e overflow tank, t h i s plug b l e w out w i t h t h e r e a c t o r at fUl power r e s u l t i n g i n a complete burp of t h e overflow tank.

On at least

t h r e e other occasions when t h e overflow tank w a s being emptied, t h e offgas plug b l e w out causing more salt t o be t r a n s f e r r e d than planned. 5.8.10 -,

Variable-Speed Drive f o r t h e Fuel PUQ

P r i o r t o February 1969 t h e f u e l pump w a s a l w a y s operated with the

normal 60-HZ power supply at

- 1189 rpm.

Then, i n order t o investigate the

e f f e c t of fuel circulation rate on system b e h a k o r (bubble ingestion, xenon stripping, t r a n s f e r t o overflow tank, b l i p s ) , l e a variable-speed motorgenerator set w a s brought t o the,:.MSRE t o supply power t o t h e f u e l pump during experiments.

As described i n Chap. 16 ,â&#x20AC;&#x2DC;considerable e f f o r t w a s

109

expended i n modifying and repairing t h e M-G set t o obtain s a t i s f a c t o r y reliability.

The pump i t s e l f , however, operated with no d i f f i c u l t y f o r

s u b s t a n t i a l periods at speeds between 50% and 105% of normal.

5.8.11

Flooding of t h e Pump Bowl w i t h Flush S a l t

A t t h e end of Run 7 i n July 1966, t h e f u e l loop w a s f i l l e d with f l u s h

salt t o r i n s e out residual pockets of f u e l salt and thus reduce t h e radiation l e v e l s f o r t h e scheduled work i n t h e reactor c e l l .

As t h e salt

w a s overflowing i n t o t h e overflow tank t o r i n s e it out, t h e l e v e l i n t h e f l u s h tank w a s lowered too far which exposed t h e bottom of t h e dip tube.

A l a r g e bubble of helium gas at a pressur,e of about 30 psig entered t h e

As t h e bubble rose, i t s volume increased due t o t h e

bottom of t h e loop,.

decreased pressure and t h e fuel-pump l e v e l increased f a s t e r than it could overflow t o the overflow tank.

S a l t flooded t h e reference bubbler, t h e

annulus around t h e s h a f t , t h e sampler tube, and t h e main offgas l i n e . Several factors contrib

vented f i l l i n g t o t h e over flushing-of t h e overflow tank.

d t o t h e accident.

These had been bypassed t o allow

line.

The flush-salt inventory w a s rn-ginal

normal operating temperatures (1200'F). . low (about 114OOF) which mad

Interlocks normally preat

The flush-tank temperatures were

t h e salt more dense and thus there w a s an

i n s u f f i c i e n t volume. 5.8.12

PluRRing of t h e Offgas Line at t h e Fuel Pump

Intermittent problems were encountered w i t h plugging of t h e 1/2-in. point ,lust beyond where it ~ched-40 main offgas l i n e a t

I n t h e e a r l y years of operation, t h i s e i t h e r melted @;ht t o power, o r it w a s periodically i t s e l f f r e e when t h e re shut down, through t h e use of a mechanical, reamed out whe

t h e fuel-pump bawl.

f l e x i b l e r o t a t i n g "snake"

J u l y 1969, a specially b u i l t heater assembly

consisting of two 1000

ds w a s i n s t a l l e d remotely k m d t h e

522 offgas l i n e between t h

e pump bowl ahd t h e gas cooling "shroud" back-blowing with helium, w a s suc-

of t h e fuel pump. cessful i n c l e

This

are given i n Chapter 8.

t a i l s on t h e offgas plugging problems

110

5.8.13

Conclusions and Recommendations

The fuel pump w a s used t o c i r c u l a t e salt f o r 21,788 hours at temperat u r e s near 1200째F with no perceptible change i n performance and no failure. Some plugging of t h e offgas l i n e w a s encountered. sidered i n future designs.

This should be con-

Perhaps dual l i n e s could be used w i t h heaters t o

m e l t out plugs i f they develop.

Leakage of o i l (1t o 2 cc/day) i n t o t h e

pump bowl contributed t o t h e offgas plugging problem.

This p o s s i b i l i t y w a s

eliminated i n t h e replacement rotary element by seal-welding a gasketed j o i n t , but because the plugging problem w a s manageable, the spare element

w a s never i n s t a l l e d . The t r a n s f e r of salt t o t h e overflow tank w a s an operating nuisance i n that it had t o be periodically t r a n s f e r r e d back t o t h e pump.

The Mark-I1

pump (which has operated f o r over 13,000 hours i n a t e s t loop) has more height i n t h e pump b o w l , eliminating t h e need f o r an overflow tank.

5.9

Primary He at Exchanger

The primary heat exchanger w a s a horizontal s h e l l and U-tube type.

During e a r l y power operation t h e heat t r a n s f e r c a p a b i l i t y w a s found t o be considerably lower than expected.

A reevaluation of t h e physical properties

showed t h a t t h e t h e m conductivity of both t h e f u e l and coolant salts w a s s u f f i c i e n t l y below t h e value used i n design t o account f o r t h e overestimate * I

of the o v e r a l l c o e f f i c i e n t of heat t r a n s f e r .

Table 5-2 shows a comparison

of t h e physical property data used i n t h e o r i g i n a l design t o t h e current data.

The heat t r a n s f e r coefficients calculated by t h e conventional design

procedures using these two sets of data are a l s o shown. The heat t r a n s f e r d i d not change during subsequent operation of t h e

reactor and no other d i f f i c u l t i e s occurred.

The measured o v e r a l l heat trans-

f e r coefficients ranged from 646 t o 675 w i t h an average of 656 Btu/(hr-ft2-OF) f o r 8 measurements.

These measurements were made on t h e basis of nominal

full power at 8.0 Mw and a coolant salt flow of 850 gpm. If t h e a c t u a l coolant f l o w rate were 770 gpm, which would be t h e flow consistent with a power l e v e l of 7.34 Mw, t h e measured o v e r a l l c o e f f i c i e n t would be 594 as compared t o a calculated value of 599 Btu/(hr-ft2-'F). covered i n d e t a i l i n Referenees 30 and 31.

The performance i s

111

Table 5-2.

Physical Properties of Fuel ark Coolant S a l t s

Used i n MSRE Heat Exchanger Design and Evaluation

Original Fuel

Coolant

Current Fuel

Coolant

Thermal Conductivity, Btu/(hr-ft-OF)

2.75

3.5

o .a32

Viscosity , l b / ( f't-hr )

20 .o

18.7

Density, l b / f t 3

17.9 154.3

0 0659 23.6

120.0

141.2

123.1

Specific Heat , Btu/(lb-OF)

0.46

0.57

0.4735

Film Coefficient , Btu/(hr-ft2-OF)

3523

5643

1497

0.577 1989

Overall Coefficient, Btu/(hr-ft2-OF)

1186

618

112

5.10 5.10.1

Reactor Vessel and Reactor Access Nozzle

General Description

Reactor Vessel and Core -The 8-ft-high

r e a c t o r v e s s e l w a s a 5-fi-diameter by

cylinder which contained a 55-in.-diameter

graphite core.

cutaway drawing of t h e r e a c t o r v e s s e l and core i s shown i n Figure 5-9.

The vessel design pressure'and temperature were 50 p s i g and 1300'F with an allowable stress of 2750 p s i . The f u e l salt entered t h e flow d i s t r i b u t o r where it was directed downw a r d around t h e circumference of t h e vessel.

It flowed i n a s p i r a l path

through a 1-in. annulus between t h e v e s s e l w a l l and t h e core can f o r cooling purposes.

Anti-swirl vanes i n t h e lower head of t h e v e s s e l straightened t h e

f l o w path before i t entered t h e graphite moderator core.

Flow passages,

formed by grooves'in t h e s i d e s of t h e graphite moderator bars, d i r e c t e d t h e laminar salt flow t o t h e upper head plenum.

The salt flow then l e f t t h e re-

a c t o r vessel through t h e s i d e o u t l e t of t h e r e a c t o r access nozzle which i s described i n t h e following section. During operation t h e f u e l salt w a s h e l d i n the c i r c u l a t i n g system by

a freeze valve (FV-103) attached t o t h e lower head of t h e r e a c t o r vessel. The drain l i n e and freeze valve w a s a 1-1/2 i n . Sch.-hO pipe which w a s f l a t t e n e d f o r about 2 i n . t o give a flow cross section about 1/2-in. w i d e . Cooling air i n a shroud surrounding the f l a t t e n e d ' s e c t i o n maintained a frozen

salt plug f o r a "closed" valve.

The 1-1/2 i n . pipe extended 2-3/4 i n . i n t o

t h e lower head of t h e v e s s e l and w a s covered with a hood f o r protection against sediments c o l l e c t i n g on t o p of t h e frozen plug making it inoperative. The 1-1/2 i n . hooded drain thus would remove a l l but a small puddle of salt

from t h e lower head even i f there had been heavy sedimentation.

To provide

f o r drainage of t h e remaining puddle, a 1/2-in. tube was mounted through t h e w a l l of t h e portion of t h e drain protruding i n s i d e t h e v e s s e l and extended downward through t h e freeze valve below.

This, i n e f f e c t , formed

p a r a l l e l freeze valves operated by t h e same controls. The r e a c t o r core consisted of 617 f u l l and f r a c t i o n a l s i z e graphite

elements 2 x 2 i n . cross s e c t i o n and about 67 i n . long.

S a l t flow channels

were formed by machined half channels on each of four faces of each element.

113

ORNL-LR-GWG 61097RiA

GRAPHITE SAMPLE ACCESS PORT

FLEXIELE CONDUIT TO CONTROL ROD DRIVES

COOLING AIR LINES

Fig. 5 . 9

E R E Reactor Vessel

114

When not buoyed by being immersed i n salt, t h e v e r t i c a l graphite elements

were supported by a l a t t i c e of graphite blocks which i n t u r n a r e supported by a Hastelloy-N g r i d fastened t o t h e bottom of t h e core can.

The score can

w a s thus f r e e t o expand downward r e l a t i v e t o i t s t o p attachment t o t h e reactor vessel while t h e graphite was free t o expand upward r e l a t i v e t o i t s support at t h e lower end of t h e core can.

The graphite moderator elements

were restrained from f l o a t i n g out of the core by a Hastelloy-N rod through holes i n t h e dowel section at t h e bottom of each graphite element.

addition a wire passing through t h e upper graphite elements prevented t h e upper portion of a broken element from f l o a t i n g out of t h e core. To prevent possible overheating i n an otherwise stagnant region, a

small portion-of s d t enteripg t h e reactor w a s diverted i n t o t h e region j u s t above t h e core can support flange i n t h e annulus between t h e vessel and t h e core can.

Vessel w a l l temperatures i n t h i s region Etnd a l s o t h e lower

head were monitored during operation f o r possible deviations. A t t h e center of t h e core were located sample specimens, t h r e e control

rod thimbles, and f i v e removable graphite elements. w e r e located i n one corner of a 4-in. occupying t h e o*her three corners.

The 'sample specimens

square with the t h r e e control rods

The five.removable+elements then occupied

t h e remaining spaces between these four-positions.

The controls rods are

described elsewhere i n t h i s report (Chapter 19). The sample specimen assemblies mentioned above were made of var.ious com-

binations and arrangements of graphite and m e t a l .

They w e r e exposed t o much

t h e same salt velocity, temperature, and nuclear f l u x as t h e core matrix.

These are described in general elsewhere i n t h i s section.

The reactor vessel w a s supported-from t h e 'top removable cover of t h e thermal s h i e l d by twelve hanger-rodcasseniblies

These hanger-rod assemblies

were pinned t o lugs welded t o t h e reactor vessel just above t h e flow distributor.

The-support ,arrangement w a s such t h a t t h e reactor vessel could be

considered t o be anchored at t h e support lugs.

The portion of t h e vessel

below t h e lugs w a s free t o expand downward w i t h no r e s t r a i n t . React0.r Access nozzle (RAN) -Attached vessel w a s a 40-in.

long lO-in.-diam

nozzle.

t o t h e upper head of t h e reactor The nozzle had a 5-in.-diam

s i d e o u t l e t f o r t h e leaving salt located about 10 i n . above t h e r e a c t o r

115

L.,

v e s s e l upper head.

The remaining extension of t h e lO-in.-diam

nozzle pro-

vided a pocket of trapped gas during t h e f i l l i n g of t h e reactor.

However,

most of t h e volume of t h i s extension w a s occupied by t h e nozzle plug which

i s a removable support f o r t h e t h r e e control rod thimbles, t h e 2-1/2 i n . graphite sample access pipe, and f o r t h e discharge screen. . See Fig. 5-10.. Conventional Leak-detected metal oval-ring- j o i n t flanges were used on both sample access â&#x20AC;&#x2122;openings t o provide t h e necessary containment of t h e primary system.

The radial clearance between t h e removable nozzle plug and t h e

nozzle w a s 1/8-in. ered annulus.

at t h e top and 1/4-in. a t t h e bottom t o provide a tap-

It was intended t h a t salt be frozen i n t h i s tapered annulus

providing a salt s e a l t o prevent molten s a l t from contacting t h e metal seali n g surfaces. possible.

However, maintaining a frozen salt s e a l w a s found t o be not

This is discussed i n more d e t a i l l a t e r .

Cooling a i r w a s pro-

vided on both i n s i d e of t h e plug and outside of t h e 10-in.lnozzle but only t o the i n s i d e of t h e plug of t h e 2-1/2 i n . graphite sample access.

Heaters

were provided t o thaw any frozen salt remaining i n t h e 10-in. annulus after

a salt drain.

For thawing t h e 2-1/2 in. annulus t h e cooling tube w a s re-

moved from t h e plug as p a r t of t h e sampling procedure and temporarily replaced w i t h a metal sheathed heater. Some twenty thermocouples (not including spares) were i n s t a l l e d a t various locations on t h e FWl t o monitor temperatures of t h e nozzle, both plugs and.contro1 rod thimbles.

A thermocouple w e l l attached t o t h e

graphite-sample access plug extended i n t o t h e flowing s a l t stream t o i n d i c a t e r e a c t o r o u t l e t tempe Strainers w e r graphite chips lar lower end of t h e 10-in. n region of t h e r e a c t

i t e sample assembly ve gr,aphite elements were

r e a c t o r o u t l e t t o prevent passage of s t r a i n e r basket w a s attached t o t h e and extended downward i n t o t h e upper e t h r e e control rod thimbles and t h e ugh t h e s t r a i n e r basket.

Since t h e

ed as were t h e remaining moderator ele-

ments, a cross-shaped extension of t h e basket assembly projected beneath t h e basket t o provide a hold-down.

See Fig. 5-11.

The core sample specimens were removed and replaced onlyâ&#x20AC;&#x2122; when t h e re-

a c t o r w a s drained of s a l t and p a r t i a l l y cooled.

The specimens were removed

116 i ORNL-DWG 68-5507 /COOLING

COOLING AIR

AIR

(Gin.

2 '/,-in.

Fig. 5.10

Reactor Access Nozzle Showing 10- and 2-1/2 i n . Annuli

117

ORNL-DWG 63-6475

CONTROL ROD THIMBLE-

Fig. 5.11

Reactor Fuel Outlet Strainer

118

through t h e smaller of t h e two access nozzles described earlier.

A stain-

less s t e e l standpipe was l e f t attached t o t h e small graphite-sample access nozzle via a bellows.

The upper end of t h e standpipe w a s bolted t o a l i n e r

s e t i n t o t h e lower concrete roof plug.

During normal operation t h e l i n e r

opening w a s closed with a concrete plug.

When samples were t o be removed,

t h i s plug w a s replaced with a work s h i e l d which contained openings f o r tools, l i g h t s , etc.

All j o i n t s were moderately leak-tight and t h e stand-

pipe was provided with a nitrogen purge and offgas connections.

A Roots

blower located i n t h e service tunnel provided a l i g h t l y negative standpipe pressure t o assure inward leakage. Core samples were taken by removing t h e graphite sample access flange, withdrawing t h e sample i n t o t h e standpipe, placing t h e sample i n a s p e c i a l c a r r i e r , i n s e r t i n g a replacement sample specimen i n t o t h e core, and replacing t h e access flange. 5.10.2

H e a t Treatment of Reactor Vessel

After t h e reactor vessel w a s i n s t a l l e d , t e s t s of t h e p a r t i c u l a r heat of Hastelloy-N used i n the vessel showed that t h e closure weld beeween t h e top head and t h e flow d i s t r i b u t o r r i n g could have poor mechanical proper-

t i e s i n t h e as-welded condition.

Therefore i n t h e f a l l of 1965, t h e reac-

t o r v e s s e l w a s heat-treated i n place, using i n s t a l l e d heaters, f o r 90 hours at 1400'F.

5.10.3

Reactor Access Nozzle Freeze T e s t s and Effect of Circulating Bubbles

!be main purpose f o r a frozen salt s e a l i n t h e annuli of t h e 10-in. and 2-1/2-in.

dim.

RAN plugs w a s t o prevent contact of t h e s e a l i n g surfaces

with salt.

The freeze flanges used as piping disconnects have a similar

fhnction.

The main difference being t h a t t h e freeze flanges incorporated

a radial freeze j o i n t while the RAN employed a longitudinal freeze j o i n t . During development on t h e Engineering T e s t Loop freeze j o i n t , it w a s noted t h a t a frozen salt seal could not be maintained r e l i a b l y i f i n cont a c t with molten salt.

It was intended t o operate t h e MSRE RAN freeze

j o i n t with a r i n g of frozen salt above t h e normal molten-salt l e v e l .

This

would be accomplished by pressurizing t h e fuel system above normal operati n g pressure soon after f i l l i n g t h e r e a c t o r with salt.

Cooling a i r on t h e

119

RAN woula freeze a r i n g of salt i n t h e annuli before t h e pressure w a s low-

ered t o i t s normal value, leaving a gas void between t h e frozen and l i q u i d

salt.

The gas trapped i n t h e RAN annuli w a s thus t h e p r i n c i p a l barrier

between t h e molten s a l t and t h e containment s e a l i n g flange.

The r i n g of

frozen s a l t would then serve only as a backup protection against sudden pressure increases forcing salt up i n t o t h e annuli. The first attempts t o form t h i s backup s e a l i n t h i s manner on t h e MSRE RAN w e r e unsuccessful due

n s u f f i c i e n t cooling a i r although t h e availa-

b l e flow met t h e requirements of t h e design

calculation^.^^

The flow w a s

permanently increased from about 3 cf'm t o t h e range of 1 5 t o 20 cf'm by modifying t h e control valve t r i m .

Even with t h e increased a i r flow, t h e r e

w a s s u f f i c i e n t movement of salt On t h e annuli due t o t h e turbuient flow below t o prevent forming a good frozen-salt seal.

However, gas trapped

above t h e molten salt kept t h e l i q u i d l e v e l ,well below t h e flange.

The

l i q u i d gas i n t e r f a c e changed i n height due t o any change i n volume of t h e gas pockets.

During e a r l y operation, the i n t e r f a c e w a s at least a foot

below t h e flange and t h e flange temperature w a s about 200째F.

bi-l

During l a t e r

operations, t h e r e were periods when t h e salt l e v e l w a s higher and t h e temperature of t h e flange approached 300'F. There were two mechanisms which caused gas t o be t r a n s f e r r e d t o and from t h i s pocket.

Gas was t r a n s f e r r e d from t h e pocket as a s o l u t e and

adde'd by entrapment of

l e s from t h e salt stream below.

The equilibrium d i f f e r e

e two t r a n s f e r rates determined t h e

salt l e v e l i n t h e annuli

emperature readings of t h e w a l l .

Any change i n reactor between these t w o t r a n s i n t h e annuli.

It was

be increased by any of t

lowering t h e system pres culating salt temperat

conditions which disturbed t h e balance ould thus change t h e height of salt he salt level i n t h e RAN annuli would g changes i n operating conditions:

e r gas i n t h e f u e l pump), increasing c i r g t h e amount of c i r c u l a t i n g bubbles

asing pump spee es of a l l t h r e e

s e causes a r e shown i n Figs. 5-12 and 5-13.

Please r e f e r t o Fig. 5-10 for location of t h e thermocouples used i n these

120

a ORNL-DWG 73-599

EACTOR OUTLET 2-

lz00

---

----------y

TE R-7B

I 400-

3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 DATE BEGINNING FEBRUARY 3,1968

Fig. 5.12

Effect of Reactor Outlet Temperature and Fuel Pump CoverGas Pressure on Reactor-Access Nozzle Salt Level (as indicated by wall temperatures)

121

ORNL-DWG 73-600

1400

1200

lo00

800

600 .

400 0

TIME, MARCH 11,1969 (htd

Fig

Effect of Fuel Pump Speed (i.e. circulating bubbles) on Reactor-Access Nozzle Salt Level (as indicated by wall temperatures)

122 i--

The e f f e c t of temperature and pressure can be seen i n Fig. 5-12.

t h e beginning of t h e observed time period, t h e salt l e v e l i n t h e RAN was r e l a t i v e l y low with t h e pump pressure higher than i t s normal 5 psig.

After

a pressure drop on February 6, t h e salt l e v e l overcame t h e immediate e f f e c t s of pressure and began t o r i s e as indicated by t h e R-43 thermocouple.

Sev-

eral days l a t e r t h e reactor o u t l e t temperature was changed i n two small steps of 30'

and 15' y e t note t h e very l a r g e e f f e c t of almost 6 0 0 ' ~ on t h e

R-7 thermocouple due t o t h e salt l e v e l increase. Later by r a i s i n g t h e pump pressure From 3 t o 5 p s i g t h e salt l e v e l i s again lowered i n t h e RAN annulus. The e f f e c t of a d i r e c t change i n t h e amount of c i r c u l a t i n g bubbles can be seen i n Fig. 5-13.

A t t h e beginning of t h i s period t h e fuel-pump speed

w a s lower than normal and it can be seen t h a t t h e RAN salt l e v e l was very high.

Other indications such as t h e r e a c t i v i t y balance l e d t o t h e conclu-

sion there-we-re no bubbles c i r c u l a t i n g a t t h i s condition.

When t h e fuel-

pump speed was increased, with no other change i n operating temperature o r pressure, t h e RAN salt l e v e l dropped very rapidiy.

(Note t h e abscissa i n

Fig. 5-13 i s now hours instead of days.) Circulating bubbles were involved i n more dramatic e f f e c t s than temperature changes i n t h e RAN, of course.

Some more important variables were

involved such as t h e e f f e c t s of c i r c u l a t i n g voids and xenon poisoning on t h e r e a c t i v i t y balance.

The e f f e c t s seen i n t h e RAI? temperatures only

helped explain the cause of some other events. One phenomenom noted i n which it is thought t h e RAN trapped gas had a d i r e c t r e l a t i o n s h i p i s t h a t of parer perturbations i n t h e MSRE i n January, 1969. There was an unusually l a r g e amount of c i r c u l a t i n g bubbles at t h i s time as v e r i f i e d by several indications including low RAN temperatures.

It i s thought some gas, clinging t o t h e core, w a s suddently swept from t h e core causing a momentary increase i n r e a c t i v i t y .

The t r i g g e r i n g mechanism

f o r r e l e a s e of gas clinging i n t h e core was presumably a very small perturbation i n f l o w o r pressure.

An occasional r e l e a s e of a b u r s t of gas from

t h e RAN which was suddenly compressed i n t h e pump could have been t h e cause of such perturbations.

There were other indications t o connect t h e RAN

trapped gas with t h e power perturbations. 47

123

5.10.4

Core Specimen I n s t a l l a t i o n and Removal

Throughout t h e operation of t h e MSRE with s a l t i n t h e primary loop t h e r e w a s a sample a r r a y of one kind o r another i n t h e core.

The a r r a y s

t h a t were exposed between September, 1965 and June, 1969 w e r e of t h e des i g n shown i n Fig. 5-14.

The a r r a y t h a t w a s i n t h e core from t h e t i m e of

construction u n t i l August, 1965 contained similar amounts of graphite and

INOR-8 ( t o have t h e same nuclear r e a c t i v i t y e f f e c t ) b u t d i f f e r e d i n i n t e r n a l configuration.

I n 1969, during t h e l a s t f i v e months of operation, a

d i f f e r e n t array, designed t o study t h e e f f e c t s of s a l t v e l o c i t y on f i s s i o n product deposition, w a s exposed i n t h e core.

A core specimen assembly of t h e type shown i n Fig. 5-14 consisted of t h r e e separable s t r i n g e r s (designated R, L, and S).

Whenever an assembly

w a s removed from t h e core, i t w a s taken t o a hot c e l l , t h e s t r i n g e r s were

taken out of t h e basket, and a new assembly w a s prepared, usually including one o r two of t h e previously exposed s t r i n g e r s .

was reused, sometimes not.

Sometimes t h e o l d basket

The h i s t o r y of exposure of INOR-8 specimens i n

t h i s kind of a r r a y is outlined i n Fig. 5-15.

The numbers i n d i c a t e t h e

h e a t s of INOR-8 from which t h e rods i n each s t r i n g e r were made. The following s e c t i o n describes t h e experience with i n s t a l l a t i o n and removal of core specimen assemblies, with emphasis on t h e procedures, t o o l s , and systems involved.

Descriptions of t h e materials t h a t were exposed and

t h e r e s u l t s of t h e i r post-operation examinations are given i n d e t a i l i n 1

references 32-36. Pre-Power Array

ecimen a r r a y t h a t w a s i

t h e core during the

prenuclear t e s t i n g

nuclear s t a r t u p experiments d i f f e r e d i n t e r n a l l y from

t h e later s u r v e i l 1

assemblies, but e x t e r n a l l y i t was similar and i t s

i n s t a l l a t i o n and removal emp

ed t h e equipment and procedures proposed f o r

l a t e r use.

l a t i o n w a s i n 1964, before s a l t had been cir-

The o r i g i n a l i n s

culated, and, although done remotely, w a s c l o s e l y observe$ t o determine needs f o r modifications i n t h e t o o l s and procedures. moved i n August, 1965 by the remote procedure. s l i g h t l y radioactive from the'''%

The assembly w a s re-

(At that t i m e the salt was

s t a r t u p experiments.)

Only minor d i f f i -

c u l t i e s were encountered w i t h some of t h e t o o l s and f i x t u r e s .

Fig. 5.14

MSRE Surveillance F a c i l i t y Inside Reactor Vessel

STilNGbR

5085 5065

STRINGER 6’

I.‘

I%,,

STRMER

S’

n I’

CORE TEMPERATURE’

1 I II

1lll

SALT tN CORE RUN NO.

Fig.

5.15

Outline

of MSRE Core Surveillaixe

Program

II,

I I

Ill

t3 WI __’

126 While t h e core access w a s open f o r t h e sample removal, a v i s u a l inspection w a s made of t h e core with a 7/8-in.-diam.

The inspection

scope.

revealed t h a t pieces were broken from t h e h o r i z o n t a l graphite b a r at t h e bottom of t h e core t h a t w a s supposed t o support t h e sample assembly. broken pieces were recovered, using a vacuum cleaner.

The

To circumvent t h i s

damage, t h e new sample assembly t o be i n s t a l l e d w a s modified t o be suspended from above, using a s p e c i a l f i t t i n g i n s t a l l e d i n t h e s t r a i n e r bask e t above t h e core.

This f i t t i n g (the "basket lock" i n Fig. 5-14) had

spring f i n g e r s t h a t locked i n t o the s t r a i n e r screen when i t was pushed i n t o place.

The lower end of t h e sample basket w a s extended t o reach t h e

hole i n t h e lower horizontal graphite bar t o provide lateral support.

Array 1. I n s t a l l a t i o n of t h e basket lock and t h e f i r s t standard arary i n the core i n September, 1965 was uneventful.

The array was removed

i n July, 1966 a f t e r about 1087 equivalent full-power hours (EFPH) of operation.

When t h e a r r a y w a s examined i n t h e hot c e l l , portions of t h e s t r i n g -

ers were found t o have been damaged. Some obstruction'(thought t o be s a l t i n t h e annulus) had been encountered i n l i f t i n g t h e assembly through t h e access nozzle.

The damage w a s not caused by handling, however, but by con-

t r a c t i o n during cooldown.

When t h e core was drained some s a l t w a s trapped

between t h e ends of graphite specimens where i t f r o z e and i n t e r f e r e d with t h e d i f f e r e n t i a l contraction of t h e p a r a l l e l graphite and metal columns during cooldown.

Array 2.

As a r e s u l t t h e graphite columns buckled.

Because of t h e damage, none of t h e t h r e e s t r i n g e r s from t h e

f i r s t array were included i n t h e second. t o prevent trapping of

The new s t r i n g e r s were modified

s a l t , but otherwise t h e a r r a y w a s l i k e t h e f i r s t .

The second a r r a y w a s i n s t a l l e d on Sept. 16, 1966, near t h e end of t h e 2month shutdown t o replace t h e main blowers. The second a r r a y w a s removed i n May, 1967.

By t h i s t i m e t h e r e a c t o r

had operated 4510 EFPH and t h e samples were removed sooner a f t e r t h e end of power operation, so t h e s a l t w a s more highly radioactive than before, r e s u l t i n g i n troublesome contamination on t h e t o o l s and i n t h e standpipe. This slowed t h e operation somewhat and i t w a s necessary t o i n s t a l l a charcoal f i l t e r i n t h e standpipe vent l i n e t o l i m i t iodine release t o t h e stack. D i f f i c u l t y was encountered i n obtaining a s a t i s f a c t o r y seal on t h e r e a c t o r access flange, which required replacement of four b o l t s .

Hot-cell examina-

t i o n of t h e a r r a y showed t h a t the basket lock had pulled out of t h e s t r a i n e r

127 screen and was s t u c k on t h e basket.

A new basket lock was t h e r e f o r e fabri-

cated and w a s i n s t a l l e d i n t h e screen. Array 3.

The t h i r d a r r a y consisted of one new s t r i n g e r and two

s t r i n g e r s previously exposed i n Array 2, so i t w a s highly radioactive a t t h e t i m e of i n s t a l l a t i o n .

No p a r t i c u l a r d i f f i c u l t y was encountered i n handling t h e a r r a y , however, s i n c e t h e equipment and procedure were designed f o r t h i s condition..

A s a t i s f a c t o r y seal on t h e core access flange

was n o t obtained on t h e f i r s t t r y , and inspection showed t h e O-ring gasket w a s dented and scratched. The b o l t holes w e r e cleaned and retapped and a new gold-plated gasket w a s i n s t a l l e d . An acceptably t i g h t j o i n t w a s then obtained. Array 3 w a s removed on A p r i l 2, 1968 a f t e r t h e conclusion of U'" operation (9005 EFPH).

Some inconvenience r e s u l t e d when t h e closure on

t h e t r a n s p o r t containment l i n e r refused t o operate properly, b u t t h e r& moval and t r a n s f e r were s a f e l y accomplished without t h e l i n e r .

While t h e

array was out of the core t

lange w a s sealed with a rubber-gasketed

blind flange t o avoid poss

damage t o t h e permanent s e a l i n g surfaces..

Array 4.

One of the o l d s t r i n g e r s and two new s t r i n g e r s made up

Array 4, which was i n s t a l l e d uneventfully on A p r i l 1 8 , 1968. was removed i n June 1969, a t 11,555

EFPH.

This a r r a y

The removal was delayed f o r

several days due t o problems with t h e removable h e a t e r t h a t w a s required

the salt from the tapered annulus between t h e removable plug and

128

5.10.5

Radiation Heating

Radiation-produced h e a t i n t h e r e a c t o r vessel walls caused t h e outer surfaces of t h e v e s s e l t o be h o t t e r than t h e adjacent s a l t by an amount t h a t w a s proportional t o t h e r e a c t o r power. would cause even higher surface temperzitures.

Any deposition of s o l i d s There were two locations

i n t h e r e a c t o r v e s s e l where s o l i d s could tend t o accumulate.

These loca-

t i o n s w e r e t h e lower head and t h e lugs located j u s t above t h e i n l e t volute and which supported t h e core matrix.

The differences between t h e r e a c t o r

v e s s e l temperatures and t h e s a l t i n l e t temperature w e r e c a r e f u l l y monitored during power operation so that this condition would be detected had i t occurred.

There w a s no i n d i c a t i o n of sedimentation during the operating l i f e of the reactor.

The temperature d i f f e r e n c e d i d increase w i t h use of zssU

f u e l ; t h i s w a s expected due t o t h e higher neutron leakage associated with t h i s fuel. 5.10.6

Discussion and Recommendations

The r e a c t o r vessel and core s a t i s f a c t o r i l y performed a l l functions f o r which it was designed.

Originally t h e operating l i f e of t h e f u e l sys-

t e m w a s t o be determined by t h e number of thermal cycles on t h e f r e e z e

flanges.

However, t h e design l i f e of t h e f r e e z e flanges w a s extended on

t h e b a s i s of development tests.

The stress-rupture l i f e of t h e r e a c t o r

vessel w a s then reviewed t o determine i f t h e operation of t h e MSRE would be limited by t h e p o s s i b i l i t y of stress-rupture cracking.

The reevalua-

t i o n indicated t h e r e a c t o r could be operated a t least another year longer than i t s o r i g i n a l l y predicted 20,000 hours stress-rupture l i f e .

Sl3,SS

The r e a c t o r access nozzle a l s o adequately performed a l l of i t s func-

tions.

It w a s obvious, however, t h a t some amount of bubbles c i r c u l a t i n g

with t h e s a l t w a s necessary t o replenish t h e gas inventory of t h e RAN ann u l i and keep the s a l t l e v e l down t o the d e s i r a b l e l e v e l .

The RAN design

w a s thus inadequate from t h e standpoint of providing a frozen s a l t seal such as was t h e case with the freeze flanges.

The i n a b i l i t y t o f r e e z e a

129

s a l t seal was t h e r e s u l t of a l a r g e r than a n t i c i p a t e d s a l t flow i n t h e

RAN annuli which increased t h e h e a t load beyond t h e c a p a b i l i t y of t h e cooling air-provided.

Additional thermocouples i n t h e area of t h e RAN

would have allowed b e t t e r d e f i n i t i o n of t h e s a l t l e v e l .

If an access flange such as t h e RAN i s t o b e used again i n a molten s a l t system, t h e r e are t h r e e p o s s i b l e methods of obtaining i t s main function

- t h a t of

preventing contact of s a l t with s e a l i n g surfaces:

(1) Provide an e x t e r n a l source of gas t o t h e annulus. determination would a l s o be necessary.

A salt level

Provide an i n t e r n a l source of gas i n t h e form of c i r c u l a t i n g

(2)

bubbles which would constantly replenish t h e gas inventory i n t h e annulus.

A trapped gas pocket a t a higher pressure than t h e pump cover pressure w i l l not remain i n d e f i n i t e l y .

(3) Reduce t h e access of flowing s a l t t o t h e annulus by b a f f l e s , l a b y r i n t h o r such, so t h a t a frozen s a l t seal can be maintained. If bubbles are t o be purposely i n j e c t e d i n t o t h e f u e l s a l t f o r t h e

removal of xenon and krypton, then i t e m (2) appears t o be t h e b e s t solution.

The j o i n t could be designed t o remain e s s e n t i a l l y f u l l of gas a t

all times.

5.11

5.11.1

Fuel and Flush S a l t Drain Tanks

Description

Two f u e l s a l t d r tank c e l l f o r t h e s a f

(FD-1 and FD-2) w e r e i n s t a l l e d i n t h e d r a i n of t h e f u e l s a l t during shutdown periods.

A t h i r d tank (FFT), a l s o 1

s t o r i n g t h e f l u s h s a l t which was used f o r cleaning up t h e f u e l c i r c u l a t i n g system before and a f t e r m a

130

A f u e l drain tank i s shown i n Figure 5-16.

The INOR-8tank was 50 i n .

i n diameter by 86 in. high, and had a volume of 80 f t 3 , s u f f i c i e n t t o hold i n non-critical geometry a l l t h e salt from t h e f u e l c i r c u l a t i n g system. The tank was provided with a cooling system designed t o remove 100 kW of f i s s i o n product decay heat.

The cooling w a s accomplished by b o i l i n g water

i n 32 double contained bayonet ttibes and thinibles i n each of t h e tanks. The f u e l f l u s h tank w a s similar t o t h e fuel drain tanks, except t h a t

it had no cooling system.

Since t h e f l u s h salt d i d not contain f i s s i l e

material, t h e only decay heating was from t h e s m a l l quantity of f i s s i o n products t h a t it removed f r o m t h e f u e l system during t h e flushing operations. The weight of salt i n each of t h e t h r e e tanks was indicated by forced

balance pneumatic weigh c e l l s .

The weights were recorded on s t r i p charts

but could be read more accurately from i n s t a l l e d mercury manometers.

Two

conductivity type l e v e l probes (one near t h e bottom of t h e tank and t h e other near t h e t o p elevation of t h e s a l t ) were provided f o r use as reference points -5.11.2

Calibration of t h e Drain Tank Weigh C e l l s Using Lead Weights

During e a r l y t e s t i n g i n t h e f a l l of 1964 t h e weigh c e l l s were c a l i brated by loading t h e tank supporting rings with l e a d b i l l e t s , tanks were a t ambient temperature during t h i s calibrakion.

The drain

D r i f t s i n indi-

cated weight of up t o 39 l b s were noted when no changes were being made. Repairs of a i r leaks i n t h e instruments caused s h i f t s of up t o 65 l b s .

The

weigh c e l l s were f u r t h e r c a l i b r a t e d during t h e addition of f l u s h and 23%J

'fie1 salt,

5.11.3

(See below.)

Addition of Flush S a l t and..Further,Calibration .of Weigh Cells

Later i n 1964, t h i r t y - s i x batches of f l u s h s a l t . ( ' ~ 2 5 0l b s per batch)

were added from cans i n a portable furnace d i r e c t l y i n t o drain tank FD-2 through a heated l i n e and flanged d i p tube attached a t t h e inspection flange on t h e tank.

Each batch o f salt added w a s weighed on accurately

c a l i b r a t e d s c a l e s and t h e t a r e weight of t h e container was checked t o obt a i n t h e net weight of salt added. near 1200'F as possible.

A l l temperatures were maintained as

During additions some ' d i f f i c u l t y was encountered

with salt freezing i n sections of t h e addition l i n e .

On one occasion t h e

131

ORNL-LR-DWC 61719

INSPECTION, SAMPLER, AND

LEVEL PROBE ACCESS

FUEL SALT SYSTEM FILL AND DRAIN LtNE

Fuel S a l t Drain Tank

132

drain-tank vent l i n e plugged.

To l o c a t e and remove t h e plug, t h e vent l i n e

tpJ

w a s heated with a torch, s t a r t i n g a t t h e tank and working down t h e pipe u n t i l gas could be blown through t h e l i n e i n t o t h e drain tank,

The plug

blew loose while heating t h e l i n e several f e e t f r o m t h e tank.

The plug location, t h e r e l a t i v e l o w temperature required t o remove t h e plug and some o i l found i n t h e nozzle t o t h e inspection flange on t h e tank l e d t o t h e conclusion t h a t t h e r e s t r i c t i o n w a s caused by an o i l residue. During t h e addition of t h e first two batches of s a l t , weigh c e l l read-

ings were taken every f i v e minutes t o determine when t h e probe l i g h t came on.

Then t h i s salt w a s t r a n s f e r r e d through t h e t r a n s f e r l i n e s t o FFT and

FD-1 and back t o FD-2 through t h e r e a c t o r f i l l l i n e s .

This operation l e f t

salt i n all freeze valves which were then frozen t o i s o l a t e t h e t a n k s ' f o r t h e first time. While t h e salt was i n t h e FFT, two more cans of f l u s h s a l t were added t o FD-2 t o recheck t h e location of t h e lower probe.

I n t h e two checks t h e

probe l i g h t came on when 469.6 and 492 l b s had been added t o t h e tank. These figures corresponded t o weigh c e l l readings of 419 and 316 l b s .

Af-

t e r a l l of t h e f l u s h salt (9230 l b s ) had been added t o FD-2, it was transf e r r e d between t h e t h r e e drain tanks.

By using t h e weigh c e l l s on both

tanks involved during each t r a n s f e r , t h e weigh cells on FD-1 and FFT were calibrated.

The indicated weight at t h e upper and lower probes on FD-1

and FD-2 w e r e a l s o noted.

The lower probe of FTT was not functioning (See

24.15). "he weight between t h e probes w a s 7353 f 176 l b s f o r FD-1 and 7545 f 191 l b s f o r FD-2. From t h e above data and subsequent operation, it w a s concluded t h a t t h e weigh c e l l s are inadequate f o r precise inventory work.

I n general,

t h e longer t h e time required f o r a given operation, t h e l a r g e r t h e e r r o r introduced.

Transfers of small q u a n t i t i e s of salt over s h o r t time i n t e r -

v a l s appeared t o give f a i r l y good data.

Using t h e tank weight a t t h e lower

probe l i g h t as a fixed reference w a s u s e f u l i n preventing t h e t r a n s f e r of too much salt t o t h e f u e l system and as an indication of t h e amount of salt remaining i n t h e tank after a f i l l . c e l l s were at fault.

The implication i s not t h a t t h e weigh

Pipe loading on t h e tanks caused by temperature

changes of t h e tanks and adjoining pipes probably caused t h e v a r i a t i o n s . Errors as high as 200 t o 300 l b s have b e e n p o t e d during t r a n s f e r s .

133

5.11.4

U-235 Fuel S a l t Addition

The c a r r i e r s a l t and l a r g e r additions of 235U enriching s a l t were added by t h e same general method used f o r adding t h e f l u s h s a l t . Details of these additions and t h e 235U c r i t i c a l i t y experiment a r e given i n Ref. 40.

5.11.5

S a l t Transfers

S a l t i s t r a n s f e r r e d between drain tanks by pressurizing t h e supply tank and venting t h e receiver tank. through t h e 1/2-in.

Originally t r a n s f e r s were made only

transfer lines (lO7-llO).

I n t h e event of a drain t o

both drain tanks followed immediately by a r e f i l l of t h e c i r c u l a t i n g loop, considerable time w a s spent freezing t h e f i l l freeze valves (105 and 1061, thawing t h e t r a n s f e r freeze valves (108 and log), refreezing them a f t e r t h e t r a n s f e r and then rethawing t h e f i l l freeze valves (105 and 106).

After

c a r e f u l consideration, procedures were modified t o permit jumpering i n t e r locks and t r a n s f e r r i n g through t h e f i l l l i n e s .

No d i f f i c u l t i e s were en-

countered, however, a l l of salt could not be t r a n s f e r r e d t h i s way since t h e f i l l l i n e s did not go completely t o t h e bottom of t h e tanks.

5.11.6

A f t e r Heat Removal

On 2/17/64 and 2/24/64, t e s t s were made of t h e heat-removal capacity of t h e steam domes.

Flush salt w a s put i n t o FD-2 and t h e s a l t w a s heated

t o %l20O0F. With 40 gallons of water i n t h e feedwater tank (FWT-2), a supply valve (LCV-807) w a s opened t o admit t h i s water t o a l l 32 of t h e bayonet cooling tubes.

The water w a s refluxed f o r approximately two hours

with cooling tower water f l 40 gal/min,

o t h e drain-tank condenser kept constant a t

The i n l e t and e x i t temperatures of t h e cooling tower water

were monitored, as were t h e drain tank temperatures and feedwater tank levels.

A t equilibrium conditions, FD-2 temperatures dropped l i n e a r l y a t

a r a t e of 86'F/hr.

The heat-removal r a t e calculated from condenser cooling

water flows and temperatures

s 139 kW/hr.

The heat-removal rate calcu-

l a t e d from t h e temperature-decay r a t e of FD-2 and t h e heat capacity of FD-2 w a s 132 kW/hr.

A second sitnilar.tset was wade by opening t h e other supply valve (ESV-807) and keeping LCV-807 closed,

The flow through t h i s valve w a s

134 marginal, the l e v e l i n t h e FWT decreased slowly and t h e heat-removal r a t e

was only 110 kW/hr based on a water heat balance and 98 kW/hr from t h e salt temperature decay rate. These data showed t h a t the bayonet cooling tubes were more than adequate t o remove the design decay heat rate of 100 kW/hr ( f o r 10-MW operation).

Since t h e r e a c t o r was run a t only approximately 8 MW, t h e margin

of s a f e t y was enhanced. Because of leaks through t h e seats of t h e water supply.valves, t h e r e a c t o r w a s operated some of t h e t i m e w i t h t h e feedwater tanks empty and procedures were w r i t t e n t o manually valve water t o t h e s e tanks i f necessary t o remove af'terheat,

When t h e salt was divided about equally between

t h e two drain tanks, as would occur on an emergency drain, no heat removal

w a s necessary.

Turning o f f t h e drain tank heaters w a s s u f f i c i e n t t o prevent For instance, on November 2, 1970, t h e

excessive temperature increases.

r e a c t o r was drained while operating a t full power. t o FD-1 and 5130 lbs drained t o -2.

4380 l b s of salt drained

With t h e drain tank h e a t e r s turned

o f f , FD-1 temperature increased from 1060 t o 1155'F i n about 6-1/2 hours and then s t e e d decreasing.

FD-2 temperatures increased from 1075 t o

ll95OF i n about'9-1/2 hours and then s t a r t e d decreasing.

5 11.7

Tanks Plugging of HeLiLmLJ;ines.-at..the.~op..of,the,~e;l..~~ain a

The helium supply, Tent and equalizer l i n e s e n t e r t h e t o p of t h e f u e l

drain tanks via a common 1/2-in.

Schedule-40 pipe.

on FD-2 became almost t o t a l l y plugged.

I n June 1969, t h i s l i n e

The plug first appeared when t h e

main 4.2 l i t e r / m i n H e offgas flow w a s routed p a s t t h e drain tanks due t o plugging i n t h e regular offgas routes.

The plug w a s a t least p a r t i a l l y

cleared by first heating t h e tank t o Ql275'F overnight and then backblowing with helium a t 42 psig.

This technique was not successful when FD-1 l i n e

plugged i n A u g u s t 1969.

Due t o t h i s plug, f u e l drained p r e f e r e n t i a l l y t o

FD-2 during a planned drain on November 2, 1969,

During t h e l a s t drain on

December 12, 1969, t h e f u e l w a s made t o d r a i n more equally by.,deliberately closing t h e vent from FD-2 f o r p a r t of t h e time during t h e draining period. The result w a s t h a t t h e total sal$ weights i n FD-1 and FD-2 were within 5% of each other af'ter t h e drain had been completed.

135

5.11.8

Loading 233U i n t o FD-2

After t h e f u e l salt had been processed t o remove t h e 235U, t h e remaining c a r r i e r salt was t r a n s f e r r e d from t h e f u e l storage tank t o t h e f u e l drain tanks.

The method of adding 233U enrichment salt w a s d i f f e r e n t than

t h a t used f o r adding t h e f l u s h salt ana 235U f u e l salt.

The loading equip-

ment as shown i n Fig. 5-17 was attached t o t h e access flange of FD-2. containing

n ,

Cans

7 kg of uranium were lowered i n t o FD-2 which contained h a l f of

the molten c a r r i e r salt.

When t h e salt had melted out of t h e can, it w a s

withdrawn i n t o t h e shield.

Mixing w a s accomplished by t r a n s f e r r i n g t h e

s a l t between t h e two drain tanks.

The 233U addition and c r i t i c a l i t y ex-

periment i s described i n d e t a i l i n Ref. 41.

5.11.9

Heatup and Cooldown Rates

These a r e described under heaters i n Section 17. Discussion and Recommendations

5.11.10

The f u e l and f l u s h s a l t drain tanks have functioned very s a t i s f a c torily.

They w e r e very sluggish thermally and therefore easy t o control

during operation.

The a f t e r h e a t removal system functioned s a t i s f a c t o r i l y

when needed. The weigh c e l l s were not completely s a t i s f a c t o r y f o r inventory purposes.

Piping s t r e s s e s should be eliminated o r other type instruments

provided.

As with other offgas l i n e s , t h e p o s s i b i l i t y of plugging should be considered.

Perhaps two l i n e s could be used with heaters attached t o melt out

any plug which might develo

136

ORNL-DWG 68-967

-TURF

CARRIER

XHAUST BLOWER

FDT ACCESS FLANGE

Fig. 5.17

Arrangement for Adding U'"

Enriching S a l t t o Fuel Drain Tank

137

COOLANT SYSTEM

C. H. Gabbard

J. K. Franzreb

M. Richardson

6.1

Description

The coolant c i r c u l a t i n g system consisted o f a centrifugal-type coolant pump and an air-cooled r a d i a t o r which were connected t o t h e f u e l heat exchanger by 5-inch INOR-8 piping. t o llOO'F

at full power.

The normal operating temperature w a s 1000

The coolant s a l t w a s c i r c u l a t e d at 850 gpm.

i n g shutdowns t h e salt was drained t o t h e coolant drain tank. had low points on each s i d e of t h e radiator. and freeze valves were required.

Dur-

The piping

Therefore, two drain l i n e s

The c i r c u l a t i n g systemwas f i l l e d by

pressurizing t h e coolant drain tank with helium and venting t h e coolant

6.2

Purging Moisture and Oxygen from t h e System

Before salt was charged i n t o t h e coolant drain tank, oxygen and mois-

ture were purged from t h e coolant c i r c u l a t i n g system and t h e drain tank. This was accomplished by first pressurizing t h e system t o assure t h a t it

was leak-tight , then evacuating,. and purging with helium during t h e heatup. Details of t h e heatup a r e given i n Section 17. The helium w a s introduced at t h e coolant pump and coolant drain tank and evacuated o r vented through t h e offgas system. i n t h e loop.

The coolant pump was operated t o provide c i r c u l a t i o n

Purging continued u n t i l t h e system w a s at 1200'F and essenti-

a l l y no more moisture w a s bein

released.

Later analysis of coolant salt

samples indicated t h a t t h e purging had been very effective.

. A f t e r t h e coolant salt 11 days.

been added t o t h e coolant drain tank (see

A t e s t of t h e drain time indicated t h a t complete draining could

be accomplished i n approximately 12 minutes a f t e r t h e freeze valves thawed. During t h e following f i l l t h

i r c u l a t i n g system w a s calibrated.

The drain

tank was pressurized i n increments and t h e drain tank weight and t h e d i f -

Lid

f e r e n t i a l pressure between t h e drain tank and t h e coolant pump were re-' corded.

The system c a l i b r a t i o n curve i s shown i n Fig. 6-1.

138

ORNL-DWG 73-601

-x

25 0

U Y

z a z

I-

azo

w m w

8 K

f c

5 W

II: 0

800

1600

2400

3200

4000

4800

POUNDS OF SALT IN COOLANT LOOP

Fig. 6 . 1

Coolant Circulating System Calibration

5600

139

6.4

Operation

Other than t h e d i f f i c u l t i e s with t h e r a d i a t o r described l a t e r , t h e coolant system operated s a t i s f a c t o r i l y . No salt leaks were detected a t any time.

6.5 6.5.1

Coolant S a l t Circulating Pump

Description The coolant pump used t o c i r c u l a t e t h e LiF-BeF2 coolant s a l t was s i m i -

l a r t o t h e f u e l pump.

It did not have a spray r i n g since f i s s i o n product

gas removal was not required. stalled.

No overflow tank D r cooling shroud was in-

A t t h e operating speed of 1750 rpm, it had an output of 850 gpm

a t a 78-ft salt head,

A small helium purge w a s introduced j u s t below t h e

shaft s e a l i n t h e bearing housing.

Most of t h e gas flowed downward through

t h e labyrinth between t h e s h a f t and t h e s h i e l d block t o prevent s a l t mist from reaching t h e seal. The remainder flowed upward t o prevent any o i l which leaked through t h e s e a l from g e t t i n g i n t o t h e pump bowl. O i l f o r l u b r i c a t i n g t h e bearings and cooling t h e s h i e l d plug w a s r e c i r c u l a t e d by an e x t e r n a l pumping system. Helium bubblers and a float-type l e v e l i n s t r u ment were used t o measure t h e l i q u i d l e v e l . not i n s t a l l e d on t h e f u e l pump.)

( A float-type instrument w a s

The coolant pump w a s located outside t h e

r e a c t o r c e l l and could be d i r e c t l y maintained s h o r t l y a f t e r shutdown. 6.5.2

Operation

The coolant pump ran very w e l l throughout t h e operation of t h e reactor. ow range of operating l e v e l s caused operaA s with t h e f u e l pump, t h e tional difficulties.

Prio

loop temperature could not pump l e v e l changed about ? l e c t e d f i l l point w a s not

s t a r t i n g t h e pump a f t e r a fill, t h e average ccurately determined.

Since t h e coolant-

l O O * F change i n loop temperature, t h e se-

s s a t i s f a c t o r y f o r operation.

I n which case

t h e freeze valves had t o be thawed and t h e system l e v e l adjusted. m a l operating l e v e l w a s about

The nor-

55% at full power ( t h e l e v e l instrument had

a span of 10 inches f o r 100% and c e n t e r l i n e of t h e volute was a t 35%).

140

When t h e pump w a s first s t a r t e d after a f i l l , t h e l e v e l would drop about

9% due t o a gas pocket i n t h e loop and about 4% due t o t h e pump fountain flow

During load and rod scrams, t h e system cooled rapidly which decreased the l e v e l below t h e interlocks required f o r s t a r t i n g t h e pump. c i r c u l a t i o n t h e r e w a s a p o s s i b i l i t y of freezing t h e radiator.

Without

An emergency

start switch w a s i n s t a l l e d which enabled t h e operator t o bypass all i n t e r locks and s t a r t t h e pump i f freezing o f t h e r a d i a t o r appeared eminent. Based on t h e o i l found i n t h e coolant pump offgas l i n e s , t h e r e was a small o i l leak i n t o t h e pump bowl.

The offgas l i n e a t t h e pump bowl be-

came plugged occasionally and i n Noveniber 1968, a clamshell heater was installed.

This w a s used once and c l e a r e d t h e l i n e s u f f i c i e n t l y t o permit

operation u n t i l shutdown December 12, 1969. The f l o a t l e v e l indicator was not used f o r operation. 6.5.3

Conclusions and Recommendations The coolant pump w a s used t o c i r c u l a t e salt f o r 26,026 hours at tem-

peratures between 1000째 and 1200째F. Some plugging of t h e offgas l i n e w a s encountered. sidered i n future design.

This should be con-

Leakage of o i l i n t o t h e pump bowl could be e l i m i -

nated by seal-welding a gasketed j o i n t as described f o r t h e fuel pump (Section 6.1).

6.6 6.6.1

Radiator

Description The nuclear power generated by t h e r e a c t o r was f i n a l l y dissipated by

a coolant s a l t - t o - a i r radiator.

S a l t flowed through t h e 120 r a d i a t o r tubes

at 850 gpm where it w a s cooled from about 1075 t o 1015'F a t f u l l power. Approximately 200,000 cfm of air from t h e two stack fans flowed perpendicular t o t h e tubes and out t h e coolant stack. In passing through t h e r a d i a t o r , t h e a i r was heated about 100째F above t h e ambient i n l e t tempera-

ture.

The heat removal was adjusted t o match t h e nuclear power produced

by r a i s i n g o r lowering t h e two r a d i a t o r doors, a d j u s t i n g t h e bypass damper, o r s t a r t i n g and stopping t h e main blowers.

141

The r a d i a t o r tubes were 0.75 i n . OD x 0.072 i n . w a l l x 30 f t long, zee-shaped t o f i t i n t o t h e insulated r a d i a t o r enclosure. enclosure supported t h e tubes and t h e e l e c t r i c heaters.

The body of t h e Two hundred

twenty-five kW of heat were provided t o prevent freezing of s a l t i n t h e

tubes i n case of a sudden l o s s of nuclear power ( s e e Fig. 6.2). "he upstream and downstream faces of t h e r a d i a t o r enclosure were equipped with doors t h a t could move up and down i n a "U" shaped track.

t h e doors moved downward i n t o t h e f u l l y closed position, camming devices operated by t h e weight of t h e door forced them against t h e seals located on t h e face of t h e radiator.

The doors were r a i s e d and lowered by means

of cables driven by a s i n g l e motor. independent operation of e i t h e r door. an e l e c t r i c a l power outage, t

Magnetic clutches and brakes permitted I n case of a load scram request o r

brakes and clutches were deenergized and

the doors f e l l closed at a r a t e which w a s limited by t h e i n e r t i a of flywheels attached t o t h e ends of t h e drive s h a f t s through overrunning clutches (s7 f t / s e c ) .

6.6.2

Early Tests and D i f f i c u l t i e s . Preliminary heatup and checkout of t h e r a d i a t o r began i n October 1964.

The main d i f f i c u l t i e s were due t o i n s u f f i c i e n t i n s u l a t i o n and warping of

As a r e s u l t , t h e doors jammed and would not operate properly and t h e r e was t o o much heat loss t o reach operating temperatures. The tem-

the doors.

perature d i s t r i b u t i o n was poor and overheating of thermocouple and e l e c t r i c a l leads w a s encountered. After unsuccessful at

t s t o correct these d i f f i c u l t i e s , t h e doors

were temporarily sealed t o allow completion of t h e c r i t i c a l i t y t e s t s while new doors were being designed and b u i l t . During t h e l a s t h a l f of 1965, extensive changes, t e s t s , and remodifications were made on t h e r a d i a t o r assembly. provided more i n s u l a t i o n f

g t h e heat source and additional expansion

j o i n t s t o prevent warpage.

e "soft" s e a l gaskets were removed from t h e

doors and relocated on t h e

i a t o r enclosure.

3/4 i n . t o 1-1/2 in. wide f o r better sealing. i n s t a l l e d on t h e doors.

New doors were i n s t a l l e d which

They were increased from A modified "hard" seal w a s

The door guide channels were repositioned t o

142

Fig. 6.2

Radiator Coil and Enclosure

143

provide more clearance.

Four cams were i n s t a l l e d on each door t o provide

more p o s i t i v e seals when they were closed. Sheet metal door hoods were i n s t a l l e d above t h e r a d i a t o r t o reduce t h e

air leakage i n t o t h e penthouse.

The synchro i n d i c a t i n g device on t h e drive

s h a f t had not proved r e l i a b l e since it w a s possible f o r t h e door t o jam and t h e shaft t o continue t o run i n t h e downward direction.

End r o l l e r s were

positioned t o prevent s i d e motion from jamming t h e doors.

Additional upper

l i m i t switches and mechanical stops were added t o prevent l i f t i n g t h e doors

i n t h e event of a l i m i t switch f a i l u r e .

L i f t i n g motor overload switches

were i n s t a l l e d i n conjunction with t h e stops. The new doors weighed about 2700 pounds each which required t h e add i t i o n of two 20,000-pound capacity, 4-in. door.

s t r o k e shock absorbers on each

Although it w a s recognized t h a t with t h e a d d i t i o n a l weight, operation

of t h e e x i s t i n g brakes and clutches w a s marginal, no changes were made i n t h e l i f t i n g mechanism.

Adjustment of t h e door-lifting cables s o that t h e

doors hung straight d i d much t o ease t h e i r operation i n t h e s l i d e s . The above modifications improved t h e heat d i s t r i b u t i o n of t h e r a d i a t o r system.

Two major d i f f i c u l t i e s were then encountered; one w a s t h e exces-

s i v e temperature rise i n t h e pent house j u s t above t h e r a d i a t o r , t h e other

was maintaining a negative pressure i n t h e pent house with respect t o t h e high-bay pressure.

This w a s needed f o r beryllium containment.

E f f o r t s t o correct t h e s e problems included relocating t h e h e a t e r l e a d junction boxes, p u l l i n g t h e thermoplastic w i r e (194'F)

from t h e junction

boxes t o t h e h e a t e r leads coming from t h e r a d i a t o r , spreading out t h e beaded heater leads t o improve t h e cooling a i r c i r c u l a t i o n through t h e wiring,

and, modifying t h e exhaust

t i n g t o d i r e c t t h e sweep of cool a i r across

t h e w i r e troughs ( ~ 5 0 0cfm)

These e f f o r t s were successful. i n reducing

t h e temperature t o about 1

u s t above t h e i n s u l a t i o n on t o p of t h e

radiator.

he door hoods and r a d i a t o r enclosure w a s

Heat leakage aro

reduced by welding up cracks, adding i n s u l a t i o n , and i n s t a l l i n g boots around t h e door-lifting cab overduct w a s i n s t a l l e d bet

the i n l e t hood.

The pent

openings i n t h e door hoods.

A 1-f't2 cross-

t h e door hoods t o r e l i e v e

e pressure with respect t o

aided b y , t h e addition of a duct from t h e pent house t o t h e i n l e t s i d e of

144

t h e south annulus blower.

This reduced t h e pent house pressure t o -0.14

t o -0.28 inches of water without t h e main blowers.

Vith both blowers i n

operation, t h e pressure became s l i g h t l y positive.

6.6.3

Subsequent Operation and Modifications The systems were heated and t h e approach t o f u l l power was s t a r t e d

e a r l y i n 1966.

The r a d i a t o r functioned s a t i s f a c t o r i l y , however, due t o

heat l o s s e s through t h e door s e a l s , . i t was d i f f i c u l t t o heat t h e empty r a d i a t o r t o an acceptable temperature d i s t r i b u t i o n p r i o r t o a fill. The heat t r a n s f e r c a p a b i l i t y w a s found t o be about 20% less than calculated.

The cause f o r t h i s discrepancy has not been d e f i n i t e l y established

but may be due t o t h e l a r g e surface-to-air temperature difference which existed.

This is described i n detail i n References 20 and 36.

Coolant salt was frozen i n t h e r a d i a t o r tubes on two occasions.

The

f i r s t occurred on May 19, 1966 when a building power failure caused a load

scram and l o s s of a l l electrical. equipment.

Two tubes of t h e r a d i a t o r w e r e

p a r t i a l l y frozen as indicated by t h e temperatures reading about 75OF low. These temperatures recovered a f e w minutes a f t e r t h e coolant flow w a s reestablished i n d i c a t i n g t h a t t h e plugs had thawed. The second incident occurred on June 27, 1966 when again t h e e l e c t r i c a l system failed, t h i s t i m e causing a coolant salt drain because of low r a d i a t o r o u t l e t temperature.

The coolant salt weigh c e l l s indicated t h a t

about 200 lbs of salt were held up i n t h e coolant system.

Since it w a s

probable t h a t t h e salt w a s frozen i n the r a d i a t o r which had cooled below t h e freezing point, t h e r a d i a t o r w a s reheated and t h e system refilled.

When

t h e pump s t a r t e d , t h e temperature scanners indicated t h a t 5 tubes were frozen and not c i r c u l a t i n g .

These tubes thawed after s e v e r a l minutes of salt

c i r c u l a t i o n without damage t o t h e tubes.

To reduce the p o s s i b i l i t y of t h i s happening again, t h e control system w a s revised so t h a t a rod scram would a l s o cause a load scram.

The low-

temperature s e t p o i n t for a load scram was increased from 900째F, which i s only 6 0 O ~above t h e liquidus temperature, t o 990째F.

Other revisions were

made so t h a t t h e coolant pump would not be turned o f f unless absolutely necessary.

A r a d i a t o r door scram test w a s conducted from f u l l power, and

145

these revisions were adequate t o prevent freezing of t h e r a d i a t o r as long as t h e coolant pump remained running.

The average temperature of the f u e l

and coolant leveled out at about 1120'F. On J u l y 17, 1966, a sudden l o s s i n power indicated t h a t a main blower had been l o s t .

Main blower No. 1 had f a i l e d catastrophically and pieces

of t h e aluminum r o t o r hub and blading had gone i n t o t h e r a d i a t o r duct.

Af-

t e r t h e r a d i a t o r had cooled, examination revealed t h a t pieces of aluminum had passed through t h e r a d i a t o r tube bundle, several dented tubes were

found, and s e v e r a l pieces of aluminum were stuck t o tubes.

Metallographic

examination determined there were no punctures i n t h e tubing and t h e radia t o r w a s safe t o operate a f t e r each tube w a s thoroughly cleaned.

The exact

cause of t h e blower f a i l u r e was never s p e c i f i c a l l y determined but examinat i o n of numerous "old" cracks i n t h e blades and hub would indicate t h a t these were t h e probable point of origin.

Flying debris caused some damage

t o t h e h e a t e r l e a d wires, wire t r a y s , e t c . It w a s apparent from inspection of t h e "hard" s e a l i n g surface, which

w a s mounted on t h e hot s i d e of t h e r a d i a t o r door, t h a t a modification would be required.

The continuous s t r i p metal s e a l was badly buckled and frac-

t u r e d due t o exposure t o the r a d i a t o r heat.

The roughness of t h i s surface

damaged t h e "soft" s e a l (two widths of 3/4-in. Johns-Mansville Thermocore

#C-197 square asbestos packing s t r i p s mounted on t h e face of t h e r a d i a t o r ) s o as t o render t h i s s e a l ineffective. A short evaluation program of s i x types of "hardrr s e a l s r e s u l t e d i n

t h e s e l e c t i o n of a seal mad ments spaced 1/32-in.

apart

of 2-1/4-in.

long overlapping s t e e l seg-

ch w a s i n s t a l l e d on t h e new doors.

The t e s t

further indicated t h a t t h e "T" b a r , t o which t h e segments were plug-welded,

bowed as much as 3/16.in. on the 3-ft-long t e s t section. t h e new doors were provi

on t h e cold s i d e of t h e

The "T" bars i n

expansion s l o t s plus a tension bar mounted ch w a s used f o r s t r a i g h t e n i n g t h e doors

a f t e r heating. Performance of t h e r a d i a t o r enclosure and door was without incident

after .the above modificati operating period.

Heat l o

were completed f o r t h e balance of t h e E R E and mechanical operation were adequate.

Programmed maintenance on t h e l i f t i n g mechanism w a s performed when t h e opportunity presented i t s e l f .

Miscellaneous s m a l l items, such as replacement

146 of t h e "soft" i n s u l a t i o n along t h e t o p of t h e o u t l e t door, replacement of

broken ceramic i n s u l a t o r s on t h e h e a t e r leads within t h e enclosufe, r e p a i r ,

of grounded heaters, e t c . , w a s t h e extent of the maintenance required.

The

door on t h e inlet s i d e of the r a d i a t o r r e t a i n e d i t s shape and sealed reasonably w e l l , t h e s e a l along t h e t o p and s i d e s of t h e o u t l e t door r e t a i n e d i t s

seal except f o r replacement o f burned-out t a p e along t h e top.

The bottom

of t h e o u t l e t door continued t o d i s t o r t and, even after being straightened sevkral t i m e s , made a poor seal as a result of t h e door bowing away from t h e soft s e a l on t h e r a d i a t o r face.

As long as t h e s i d e and top seals were

i n t a c t , t h e r e w a s no chimney e f f e c t and t h e heat l o s s e s were within usable limits.

6.6.4

Automatic Load Control "he heat removal rate at t h e MSRE was dependent upon t h e positions of

the radiator doors and bypass damper and on t h e operation, of t h e main blow-

ers.

These could be manually manipulated as desired.

Instrumentation was

a l s o provided t o automatically sequence t h e operation of these components so t h a t t h e operator could use one switch f o r increasing o r decreasing heat removal rate.

An intermediate r a d i a t o r door s e t t i n g w i t h one blower

i n operation and 1 MW of heat removed was t o be used as t h e s t a r t i n g point. Since it w a s found t h a t t h e heat removal with t h e doors a t t h e i r minimum s e t t i n g of 12-in. w a s about 1.9 MW, t h e automatic system did not function properly.

The r e s t of t h e sequencing functioned properly during a t e s t .

However, i n v i e w of t h e above and since manually adjusting t h e h e a t load

w a s not a disadvantage, t h e automatic load control w a s never used i n t h e operation of the reactor.

6.6.5

Comments and Recommendations Assessment of t h e r a d i a t o r containment, doors, and door-lifting mech-

anism i s complicated by t h e many changes, l a r g e and small, which were required before acceptable performance w a s achieved. The r a d i a t o r enclosure d i f f i c u l t i e s can be a t t r i b u t e d p r i n c i p a l l y t o d e t a i l s of i n s t a l l a t i o n such as gross air leakage i n t o and around t h e enclosure as a result of cracks, unsealed openings, i n s u f f i c i e n t i n s u l a t i o n , lack of a s u f f i c i e n t number of expansion j o i n t s , e t c .

The p r i n c i p a l radi-

a t o r enclosure support "H" beam members were not v e r t i c a l and t h e r e f o r e

147

the adjustable "U" shapped door guides located within these members had t o be remachined t o prevent the doors from binding. The l i f t i n g mechanism was unduly complex.

A s u b s t a n t i a l increase i n

both r e l i a b i l i t y and control would have been achieved by providing each door with i t s own motor-gear reducer u n i t with a b u i l t - i n brake on t h e moA s m a l l , fast-acting brake i n t h e high-speed end of t h e

t o r r o t o r shaft.

drive would have reduced the number of electromechanical devices and reduced o r eliminated t h e problem of coordinating time constants between clutches, brakes, and motor.

The a b i l i t y t o operate t h e doors indepen-

dently w a s probably unnecessary. The u n r e a l i s t i c close tolerances between t h e doors and t h e radiator enclosure and d i s t o r t i o n s produced by thermal gradients, primarily i n t h e doors, required a massive e f f o r t t o ' c o r r e c t i n order t o provide t h e required seal.

These factors should be given careful consideration i n any

future design.

6.7

b 6.7.1

Performance of Main Blowers, MB-1 and MB-3

Description The two main blowers are a p a r t of t h e heat r e j e c t i o n equipment where

the heat generated by t h e reactor i s dissipated t o t h e atmosphere.

The

blowers were manufactured by t h e Joy Manufacturing Co. and a r e t h e "Axivane" type Model AR-600-3600-1225.

The blowers a r e direct-connected t o 250-hp,

1750-rpm motors by a short drive s h a f t with a f l e x i b l e shim coupling on each end. lthough t h e blowers r e e s s e n t i a l l y muse blowers were each o r i g i n a l l y 114,000 cfm of f r e e a i r de 100,000 cfm from each b l operating range of t

en o r i g i n a l l y purchased f o r another program, eginning of t h e MSRE program.

The

d at 82,500 cfm a t 15 inches of water o r The MSRE r a d i a t o r design requirement of

9 inches of water w a s within t h e normal , Had new blowers been proc

MSRE, t h i s same type of equipment would have been considered a s a t i s -

factory choice.

148

6.7.2

Preoperational Testing

preoperational t e s t i n g of t h e blowers consisted of about 25 hours

of t o t a l operation.

When t h e blowers were first delivered and i n s t a l l e d ,

they were operated a short t i m e ( l e s s than 10 hours) t o provide t h e elect r i c a l load f o r t e s t i n g t h e d i e s e l generators.

The blowers were a l s o oper-

ated f o r a s h o r t period of t i m e p r i o r t o power operation of t h e MSRE t o check t h e r a d i a t o r tubes f o r self-excited vibrations, t o permit a calibrat i o n of t h e s t a c k a i r flow instrumentation, and t o determine if control c i r c u i t r y w a s necessary t o avoid c e r t a i n abnormal operating conditions. During t h e abnormal operation t e s t i n g , the blade p i t c h was found t o be

less than t h e s p e c i f i e d 20 degrees. t e s t program was repeated.

The p i t c h w a s s e t c o r r e c t l y and t h e

The t e s t s indicated t h a t t h e blowers would

operate s a t i s f a c t o r i l y a t a l l t h e normal conditions, but t h a t a "surge" condition would occur i f t h e bypass damper were p a r t i a l l y closed when t h e r a d i a t o r doors were a l s o closed.

The manufacturer had s t a t e d during dis-

cussions of t h e t e s t program that t h e "surge" condition could be expected but t h a t no damage would occur o t h e r than overloading t h e d r i v e motors. The blowers were shut down immediately after t h e "surge" condition w a s en-

countered and administrative control was u s e d t o prevent f u r t h e r operation under surge conditions.

A second p i t c h change w a s made t o 2 2 4 2 degrees

immediately after t h e first full-power operation.

This 22-1/2 degree p i t c h

gave t h e maximumblower performance within t h e power r a t i n g of t h e drive motors.

6.7.3

Operating History The aerodynamic performance of t h e blowers w a s s a t i s f a c t o r y through-

out t h e operating l i f e of t h e reactor.

However, t h e r e were s e v e r a l me-

chanical failures which are l i s t e d i n Table 6.1 and which a r e discussed below.

This table a l s o shows the e f f e c t of t n e f a i l u r e on t*dperation

of the reactor. MB-1 Coupling Failure

- After' about

550 hours of operation, t h e f l e x i -

b l e coupling on t h e motor end o f t h e f l o a t i n g drive shaft f a i l e d during operation of MB-1.

"he motor end o f t h e s h a f t w a s then supported only by

t h e coupling guard, and t h e shaft whip t h a t occurred as t h e blower r o t o r

le 6.1

Date

Summary of Main Blower Failures

Failure

Effect on Reactor Operation

6/14/66

Coupling Failure MB-1

Zero-Power Operation for 14 hours

7/17/66

Hub and Blade Failure MB-1

Premature Shutdown from Run No. 7 Partial Power Operation During Run No. 8 Zero-Power Operation for 3 days Partial Power Operation for 6 days Maintenance Performed During Scheduled Zero-Power Operation Completed During Scheduled Maintenance Period

9/17/69

Bearing and Mount Replacement MB-1

Zero-Power Operation for 2 days

9/19/69

Faulty Bearing Replacement MB-1

Partial Power Operation 3 days Zero-Power Operation for 3 days

* CI \Fi

150

coasted down destroyed t h e guard and a l s o t h e coupling on t h e end of t h e s h a f t near t h e blower.

-W

The s h a f t came t o r e s t embedded i n t h e sheet-metal

nose with t h e blower end of t h e coupling near, but completely severed from,

i t s mating flange on t h e rotor.

The motor end of t h e s h a f t extended radi-

a l l y across t h e blower i n l e t at about t h e 4 o'clock position. Scratches and dents on t h e blades indicated t h a t some of t h e debris from t h e f a i l e d couplings had gone through t h e blower i n t o t h e r a d i a t o r However, t h e damage, except t o t h e couplings, appeared t o be super-

duct.

f i c i a l and of l i t t l e consequence.

N o a d d i t i o n a l inspection of t h e blower

was made. The primary cause o f t h e f a i l u r e , we believe, w a s a f a t i g u e f a i l u r e of t h e f l e x i b l e shims.

The exact cause o f t h e shim f a i l u r e i s unknown.

Some

home-made washers were found i n t h e d r b r i s t h a t were not rounded as they should have been.

High s t r e s s e s at t h e corners of these washers could have

s t a r t e d t h e f a i l u r e , o r t h e coupling could have been running with excessive misalignment.

The blower r o t o r and drive s h a f t assembly had been r o t a t e d

by hand when t h e f i n a l blade p i t c h change w a s made about 216 operating hours before t h e f a i l u r e .

There may have been some cracked shims i n t h e

coupling at t h a t time, but t h e shims were s t i l l supporting t h e s h a f t . The couplings were r e b u i l t with new b o l t s , washers, and shims.

The

s h a f t was realigned, and t h e blower w a s returned t o service.

The couplings

on MB-3 were also disassembled and new shims were i n s t a l l e d .

The o r i g i n a l

shims i n these couplings were not i n bad condition. MB-1 Hub and Blade Failure.-After

about 650 hours of a d d i t i o n a l op-

e r a t i o n , t h e r e w a s a second failure of -1.

ure of t h e blading and t h e hub of MB-1.

This w a s a catastrophic fail-

The o u t e r periphery of t h e r o t o r

hub disintegrated and all of t h e blading was destroyed.

Most of t h e frag-

ments were contained i n t h e blower casing, but numerous pieces of t h e c a s t aluminum-alloy hub and blades entered t h e r a d i a t o r duct and some of these a c t u a l l y passed through but did not seriously damage t h e r a d i a t o r tube bundle.

Figure 6 . 3 shows t h e f a i l e d hub and t y p i c a l pieces of t h e blading.

An inspection of t h e broken pieces of MB-1 revealed numerous "old" cracks i n t h e blades and i n t h e h& as evidenced by darkened o r d i r t y areas on t h e fractured surfaces.

One blade i n p a r t i c u l a r had f a i l e d along a

h*'

151

Fig. 6.3

Outer P e r i p h e r y of F a i l e d MB-1 Hub and Q p i c a l Blade F r a c t u r e s

152

large "old" crack.

The hub had contained s h o r t , 1-1/2 t o 2-inch, circum-

f e r e n t i a l cracks at t h e base of 8 of t h e 16 blade sockets and t h e f a i l u r e generally followed these cracks.

Figure 6.4 i s a piece of t h e hub showing

t h e darkened areas t h a t indicated previous cracking a t t h e base of t h e blade sockets.

Similar cracking w a s found i n t h e hub of t h e spare blower

t h a t had been i n storage and a continuous crack extending about 35%of t h e circumference as well as some of the s h o r t e r cracks were found i n t h e MB-3 hub.

No cracking was found i n t h e blading of MB-3 o r t h e spare blower.

The exact cause of t h e MB-1 failure i s uncertain, but a blade f a i l u r e at

one of t h e e x i s t i n g cracks seems t o be t h e most probable first event.

All three of t h e blowers were r e b u i l t by t h e manufacturer with new redesigned hubs and l i g h t e r magnesium-alloy blading.

The hubs were rein-

forced with r a d i a l r i b s t o reduce t h e bending moments i n t h e areas where t h e cracking had been observed.

There have been no f u r t h e r cracking prob-

lems i n e i t h e r of t h e r e b u i l t units i n about 11,000 eration.

- 12,000 hours of

op-

A more complete discussion of t h e f a i l u r e s and of t h e corrective

action t h a t was taken is given i n References 42 and 43. Thrust Bearing Failures -There

were a t o t a l of s i x f a i l u r e s assoc-

i a t e d with t h e main t h r u s t bearing o r i t s mounting.

Main blower MB-3 was

taken out of service on March 6, 1967 f o r a bearing replacement a f t e r t h e vibrations had increased from a normal value of about .8 t o 2.5 mils.

The

bearing contained an excessive quantity of very o l d appearing grease, and t h e bearing surfaces were severely worn and p i t t e d .

This bearing w a s ap-

parently from t h e replacement blower t h a t had been i n storage, and t h e bearing apparently had not been cleaned and relubricated when t h e r e b u i l t r o t o r w a s assembled. failure.

Improper l u b r i c a t i o n was judged t o be t h e cause of

The bearing had operated about 1800 hours.

Main blower MB-3 was shut down a second time f o r a bearing replacement on January 16, 1968, because of an increase i n bearing vibration.

Although

t h e vibration and temperature had been normal, t h e MB-1 bearing w a s a l s o found t o be defective and was replaced a t t h e same time.

The MB-1 bearing

had one p i t t e d b a l l , but t h e MB-3 bearing w a s severely damaged and one b a l l had a c t u a l l y fractured.

Figure 6.5 i s a photograph of these two b a l l s .

Since improper lubrication did not seem t o be t h e cause of these two f a i l u r e s and since three bearings had f a i l e d i n r e l a t i v e l y short operating

Fig. 6 . 4

Fragment of MB-1 Hub Showing Areas of Previous Cracking

154

Fig. 6.5

B a l l Bearing Failures Taken from MSRE Main Blowers MB-1 and MB-3 i n January 1968

times, t h e expected l i f e of these bearings w a s calculated from an estimated s e t of load conditions.

The predicted l i f e w a s 5000 hours as compared t o

operating times of 6630 and 4800 hours respectively f o r MB-1 and MB-3.

a r e s u l t of t h i s evaluation, t h e o r i g i n a l radial-type b a l l bearings were replaced with an angular-contact type having i d e n t i c a l mounting dimensions. The angular contact bearing had a greater t h r u s t load capacity. The angular contact bearings were apparently capable of s a t i s f a c t o r y l i f e , but additional d i f f i c u l t y was experienced w i t h t h e self-aligning mount.

The MB-3 bearing was replaced during a scheduled maintenance period

because of a thumping noise t h a t was heard when t h e blower was turned by hand.

The bearing was found t o be i n good condition, but t h e wear p a t t e r n

indicated t h a t t h e outer race had been misaligned w i t h t h e shaft.

The mis-

alignment plus t h e high r a d i a l clearance i n t h i s type bearing caused t h e noise.

No f u r t h e r d i f f i c u l t y was experienced with MB-3.

Blower MB-1 was shut down on September 17, 1969 because of an increase i n vibration.

The vibration increase had been caused by excessive wear i n

the spherical surfaces of t h e self-aligning mount. mount a l s o required replacement of t h e bearing. I

The replacement of t h e

The new bearing failed by

overheating during i t s test run because of i n s u f f i c i e n t clearnaces i n t h e phenolic b a l l r e t a i n e r .

This bearing was then replaced w i t h a bearing of

t h e o r i g i n a l r a d i a l type t h a t was on hand. f o r t h e remaining operation of t h e reactor.

This bearing was s a t i s f a c t o r y

The defective bearing w a s re-

turned t o t h e vendor.

6.7.4

Surveillance of Blower Operation Following t h e hub and

f o r monitoring t h e o ups were mounted horizon motors and thermoco A vibration meter w i t

de f a i l u r e of MB-1, a surveillance program

t h e blowers w a s i n i t i a t e d .

Vibration pick-

each bearing of t h e blowers and drive s t a l l e d i n each of t h e blower bearings. put channels w a s provided f o r each blower,

and t h e vibration of

g w a s recorded i n peak-to-peak

once per s h i f t by t

personnel.

displacement

The thermocouples were moni-

tored by t h e computer and i n i t i a l l y an a l a r m would be given i f t h e temperat u r e exceeded a preselected l i m i t .

This was revised l a t e r t o alarm i f t h e

156 temperature rise above ambient exceeded a l i m i t .

This permitted much

t i g h t e r operating l i m i t s because t h e ambient fluctuated w i t h t h e outside

a i r temperature.

I n addition t o t h e displacement readings taken from t h e vibration meters, an output w a s a l s o available t o display t h e vibration t r a c e on an oscilloscope o r other high-speed recorder.

Oscilloscope photographs were

taken f o r reference on new bearings and when t h e vibration meters indicated an increase.

Figure 6.6 shows a comparison of vibration t r a c e s from a nor-

m a l bearing and f r o m t h e bearing containing t h e fractured b a l l .

A complete inspection of t h e blowers including a dye-penetrant examina-

t i o n of t h e hubs and blading was completed annually during shutdown periods.

6.7.5

Discussion and Conclusions Following t h e major rebuilding of t h e main blowers i n

1966, t h e oper-

a t i o n has been generally s a t i s f a c t o r y even though there have been s e v e r a l power reductions caused by bearing problems.

The t o t a l l o s t operating time

caused by t h e various bearing f a i l u r e s w a s 8 days of zero-power operation and 9 days of partial-power operation.

A b e t t e r t h r u s t bearing arrangement

could have been required. The surveillance program t o monitor t h e mechanical performance of t h e

blowers has been very e f f e c t i v e .

All t h e bearing problems except t h e MB-1

p i t t e d b a l l i n January 1968 have been found by an increase i n vibration. When t h e b a l l fractured i n t h e MB-3 bearing, a temperature increase had preceded t h e vibration increase by several days.

However, t h i s tempera-

t u r e increase had gone unnoticed because t h e bearing temperature was well below t h e l i m i t t h a t had been set.

The computer program w a s t h e r e f o r e re-

vised t o alarm on t h e temperature rise above ambient, and much t i g h t e r l i m i t s were s e l e c t e d based on t h e previous operating experience.

This

alarm system was e f f e c t i v e i n detecting t h e defective new bearing on September 19, 1969.

The operating experience with t h e vibration monitoring system has c l e a r l y shown t h a t t h e displacement meter alone i s not adequate t o monitor t h e condition of t h e bearings.

Some system of maintaining a h i s t o r y of

t h e frequency spectrum i s a l s o required.

We used photographs of t h e o s c i l -

loscope t r a c e t o document the normal operating vibrations, and a d d i t i o n a l

157

Normal Trace -Taken 12/19/67

Trace w i t h

ured Ball -Taken 1/16/68 lower MB-3 Vibration Traces

158

photographs were taken f o r comparison when an increased vibration w a s indicated by t h e displacement meters.

A modification t o t h e computer t o in-

crease i t s sampling r a t e would have permitted t h e frequency spectra t o be generated by t h e computer.

This would have provided a more convenient and

more precise spectrum, but some degree of background experience would s t i l l be required f o r proper i n t e r p r e t a t i o n o f t h e s p e c t r a i n e i t h e r case because

there can be several normal frequencies.

There were several instances of

abnormal vibrations when t h e bearings were not a t fault.

The f i n a l de-

cision t o replace a bearing w a s made i n each case by t h e sound and f e e l of t h e blower during a coastdown from operating speed.

6.7.6

Recommendations The main blowers were conventional commercial equipment which would

normally be expected t o operate w i t h a minimum of a t t e n t i o n o r maintenance. The instrumentation of all r o t a t i n g equipment i s probably impractical i n regard t o both cost and manpower.

However, c e r t a i n c l a s s e s of equipment

should be instrumented and monitored on a routine basis.

The equipment

should be selected on t h e basis of: 1. Size and cost

The a c c e s s i b i l i t y during operation

The e f f e c t of shutdown on o v e r a l l p l a n t operation

The p o t e n t i a l f o r self-destruction

I d e a l l y t h e monitoring should be computerized but t h i s i s not a c t u a l l y required. Each type and piece of equipment w i l l have i t s own operating character i s t i c s i n regard t o vibration o r temperature, and these c h a r a c t e r i s t i c s may vary with t h e i n s t a l l a t i o n o r operating conditions.

Preli&nary limits

may be obtained from the manufacturer o r from Reference 44, but t h e f i n a l operating limits should be selected from a background of normal operating experience. "he experiences w i t h t h e cracked hubs and w i t h the o r i g i n a l t h r u s t bearing design are indications t h a t t h e Q u a l i t y Assurance Program should extend i n t o t h e design phase as w e l l as t h e fabrication phase of major o r c r i t i c a l items of equipment.

159

6.8 6.8.1

Coolant Drain Tank

Description The coolant drain tank was similar t o but smaller than t h e f u e l f l u s h

tank.

It w a s 40 in. i n diameter by 78-in. high and had a volume of 50 f t 3 .

A s t h e r e was e s s e n t i a l l y no a f t e r h e a t i n t h e coolant salt, cooling was not

provided.

Two s a l t l i n e s (204 and 206) and t h e i r associated freeze valves

were used f o r f i l l i n g and draining t h e coolant c i r c u l a t i n g loop. The weight of salt was indicated by forced balance w e i g h c e l l s .

The

w e i g h t s were recorded on s t r i p charts but could be read more accurately

from i n s t a l l e d mercury manometers.

Two conductivity type .probes (one near

the bottom and t h e other n e a r - t h e top elevation of t h e s a l t ) were provided

f o r use as reference points. 6.8.2

Calibration of t h e W e i a Cells Using Lead..Wei&ts The weigh c e l l s were c a i b r a t e d by loading t h e drain tank w i t h l e a d

billets.

The drain tank was a t ambient temperature during t h e c a l i b r a t i o n .

There was good agreement between the manometer readings and t h e a c t u a l weight added.. D i f f i c u l t y was encountered with a i r and mercury leaks i n t h e readout system.

The weigh c e l l s were f u r t h e r c a l i b r a t e d during t h e

addition of coolant salt. 6.8.3 -

Addition of Coolant S a l t Twenty-two batches of coolant salt (5,756 l b s ) w e r e added f r o m cans i n

a portable furnace d i r e c t

t h e coolant drain tank through a heated l i n e

attached t o t h e t o p of t h

After t h e addition w a s complete, t h i s

l i n e w a s cut off and welde

, A good l i n e a r p l o t of t h e manometer

readings vs a c t u a l weight w a s obtained except f o r a 150-lb s t e p i n t h e curve which occurred whe repaired.

The r e l a t i o n s h i p

n f i t t i n g i n t h e w e i g h c e l l tubing w a s een t h e weigh c e l l readings and t h e loca-

t i o n s o f t h e probe lights The weighing system o

more s t a b l e than on t h e

at about 12OO0F, t h e ext period of a week, was k 22 l b .

oolant drain tank proved t o be considerably n tanks.

With 5756 l b of salt i n t h e tank indicated weights, taken over a

160

6.8.4

Heatup and Cooldown Rates These are described under heaters i n Section 17.

6.8.5

Discussion and Recommendations

No unusual d i f f i c u l t i e s were encountered w i t h t h e coolant drain tank. The weigh c e l l s functioned s a t i s f a c t o r i l y , however, t h e trouble w i t h t h e f u e l drain tank weigh system emphasizes t h e need f o r c a r e f u l consideration of piping s t r e s s e s o r t h e use of other type instruments.

161

COVER-GAS SYSTEM P. H. Harley

The MSRE cover-gas system consisted of a helium t r a i l e r f o r normal supply, two banks of t h r e e helium cylipders f o r emergency use, two p a r a l l e l t r e a t i n g s t a t i o n s for removing moisture and oxygen t o a nominal 1 ppm by volume, a t r e a t e d helium surge tank, and two headers, 35 p s i g and 250 psig, which supplied cover gas t o t h e various systems.

Each of t h e t r e a t i n g sta-

t i o n s consisted of a dryer (molecular sieve) , preheater, and an oxygenremoval u n i t (heated titanium sponge).

Moisture and oxygen analyzers could

be valved i n t o monitor the t r e a t e d o r untreated helium as desired.

7.1

I n i t i a l Testin&

-The components o f t h e t r e a t i n g s t a t i o n were developed and designed by t h e Reactor Division Development Section.

The performance of t h e sys-

tem w a s t e s t e d by measuring oxygen and moisture content of t h e cover gas upstream and downstream of t h e t r e a t i n g s t a t i o n t o insure proper operation.

Other t e s t i n g a t t h e MSRE consisted of leak-testing t h e system and

checking flow control and temperature controls.

Thermal expansion of t h e

O2 removal beds during t h e i n i t i a l heatup caused t h e 3-in. r i n g j o i n t flanged heads t o leak.

After retightening t h e flanges, t h e u n i t s were kept

hot t o prevent thermal cy L a t e tests indicated

which w a s i n s t a l l e d t o red susceptible t o mois

Fisher Model 67-H pressure regulator from t r a i l e r pressure t o 250 p s i g w a s rough t h e diaphram. These tests indie Model RBX-204-15 were b e t t e r from

162

t h i s standpoint.

The Victor model was less complicated and therefore w a s

i n s t a l l e d i n t h e MSRE. The low-pressure cover-gas header was i n i t i a l l y a 40-psi header which

was protected by a 48-psig rupture disc.

During i n i t i a l t e s t i n g , t h i s rup-

t u r e d i s c f a i l e d on several occasions without apparent cause.

The header

pressure w a s subsequently lowered t o 35 p s i g which w a s s u f f i c i e n t f o r all normal operating requirements and did a l l e v i a t e t h e problem.

The s a f e t y

l i m i t s 2 0 (Lo H e supply pressure) had t o be lowered from 30 t o 28 p s i g t o

provide a margin from the normal operating pressure. Four helium control valves failed i n t h e first month of operation. Some g a l l i n g of t h e close f i t t i n g t r i m (17-4 PH plug and S t e l l i t e s e a t )

was apparently caused by the extremely clean, non-lubricating d r y helium. Other valves i n t h e system, including the four replacements have operated

sat i s f a c t o r i l y

7.2

Normal Operation

Since p u t t i n g the cover-gas system i n operation i n t h e f a l l of 1964, sixteen t r a i l e r s of helium have been used.

About 21,800 f t 3 of t h e

29,000 f t 3 i n each t r a i l e r was used before returning t h e t r a i l e r t o be re-

filled.

A t r a i l e r l a s t e d about

4 months on t h e average o r an average usage

of approximately 180 cubic feet p e r day. operating w a s 260 cubic f e e t per day.

Normal usage when t h e r e a c t o r was

Larger amounts were used during pur-

ging and f i l l i n g procedures and very l i t t l e w a s used during maintenance periods. Although helium w a s t h e normal cover gas, during maintenance periods t h e r e a c t o r system w a s purged with argon.

Argon w a s a l s o used during a

s p e c i a l study of c i r c u l a t i n g gas i n t h e fuel. &ppm

Bottles of argon containing

moisture were placed on one of t h e emergency cover-gas headers and

then fed through t h e heating s t a t i o n i n t o t h e regular cover-gas distribut i o n system. The cover gas w a s changed from one t r e a t i n g s t a t i o n t o t h e other only once i n t h e five years of operation.

This w a s done i n February 1967 when

trouble w a s encountered i n t h e power supply control c i r c u i t t o t h e operat i n g unit.

Although t h e dryer which had been i n operation p r i o r t o t h e

163 change was regenerated, there has been no indication t h a t t h e capacity of any dryer o r 02-removal u n i t has been exceeded. Spectrographic analysis of t h e helium received averaged 2 ppm oxygen and. 4.6-ppm moisture. helium contained

A f t e r passing through t h e t r e a t i n g s t a t i o n , t h e

0.2-ppm 02 and from 0.5 t o 10-ppm mo&ture.

The vari-

a t i o n i n moisture analysis appeared t o follow t h e ambient temperature of t h e helium t r a i l e r and t r e a t i n g s t a t i o n .

During hot weather t h e analyses

w e r e high and during cold weather t h e r e s u l t s were low.

Fluctuations could

be seen from day t o night temperature changes of several ppm moisture.

This f l u c t u a t i o n of moisture analysis w a s decreased by b e t t e r heating of the sample s t a t i o n location i n t h e winter and b e t t e r v e n t i l a t i o n of t h e

area during t h e summer months While i n v e s t i g a t i n g t h e

i s t u r e analysis-problem, another dryer was

i n s t a l l e d i n t h e main helium header between.the nonnal t r e a t i n g s t a t i o n and t h e l i n e t o t h e analyzers.

Whether t h i s second drying s t e p helped i s doubt-

ful but .it d i d give assurance t h a t t h e helium w a s as dry as possible.

The e l e c t r o l y t i c c e l l i n t h e moisture analyzer caused considerable d i f f i c u l t y during e a r l y operation.

The c e l l was replaced four times be-

tween June and October 1965, once i n June 1966, and i n January 1967. c e l l replacements have been necessary since then.

of t h e c e l l s was improved.

Apparently t h e q u a l i t y

Only one e l e c t r o l y t i c c e l l has f a i l e d i n t h e

oxygen analyzer and t h a t was i n June 1965.

To check t h e f i n a l performance of t h e t r e a t i n g s t a t i o n i n October 1969, a cylinder of He containing moisture was connected t o t h e analyzers. The analyzers read 20-ppm moi through t h e t r e a t i n g s t a t There has only been i n i t i a l testing.

This w

e and 0.6-ppm 02.

A f t e r t h e gas w a s run

analyzed 0.9-ppm moisture and 0.2-ppm 02.

failure of the 48-psig rupture d i s c since t h e sed by improper i n s t a l l a t i o n .

Each time the

s i d e of t h i s rupture d i s

s t was run, pressure w a s put on t h e back eveht it f r o m breaking. I n August 1969,

when t h e backpressure wa

ased, t h e rupture d i s c s t a r t e d leaking.

During replacement, t h e

support was found t o be on t h e wrong s i d e

60-psig f u e l system pres

of t h e disc.

Repeated flexing, when backpressure had been applied, was-

t h e apparent cause of t h e f a i l u r e .

More importantly, . t h e f a u l t y i n s t a l l a -

t i o n had n u l l i f i e d t h e s a f e t y function of t h e disc.

164

7.3

Conclusions and Recommendations

"he cover-gas system performed q u i t e s a t i s f a c t o r i l y w i t h a minimum amount of maintenance o r operator time.

Since t h e moisture and oxygen i n

t h e helium as received was very l o w , it i s doubtful t h a t t h e t r e a t i n g s t a t i o n s were r e a l l y necessary.

By obtaining analysis of t h e cover gas before

usage and having an on-line analyzer, t h e t r e a t i n g s t a t i o n could probably be eliminated.

I n l a r g e r r e a c t o r s , t h e t r e a t i n g s t a t i o n might be needed

t o dry t h e e a r l y recycle gas.

The MSFE charcoal beds remained "wet" f o r

a long period during e a r l y operation. W e have had considerable f l u c t u a t i o n i n t h e indicated moisture.

t h e t r e a t i n g s t a t i o n and analyzers w e r e located i n an a r e a which had a con-

s t a n t temperature, t h e analyzers would be more consistent.

If no t r e a t i n g

s t a t i o n i s used, t h e helium supply should be kept a t a constant temperature.

Rising temperature drives moisture out o f t h e container w a l l s and

decreasing temperature causes t h e w a l l s t o absorb moisture.

(14

165 8.

OFFGAS SYSTEM

A. I. Krakoviak 8.1

Description *

The offgas systems were comprised of those l i n e s , filters, charcoal beds, and valves through which cover-gas flowed from t h e s a l t systems t o t h e environment.

The off-gas systems f o r t h e coolant s a l t and f o r t h e

f u e l s a l t were similar i n many respects, but t h e coolant off-gas system

was simpler because i t did not have t o d e a l with t h e i n t e n s e t y radioactive f i s s i o n products found i n t h e ' f u e l off-gas during pdwer operation. 8.1.1

Coolant Off-Ras System Figure 8.1 i s t h e flowsheet f o r t h e coolant off-gas system (simplified

by t h e omission of elements which are not p e r t i n e n t t o t h e presentation Gas entered t h e coolant pump bowl

of operating experience i n t h i s r e p o r t ) .

through bubbler l e v e l i n d i c a t o r s and as a purge t o t h e pump s h a f t annulus. The s h a f t purge flow was s p l i t , with some flowing downward i n t o t h e pump bowl and t h e rest upwards and out through l i n e 526.

The upflow was t o

prevent o i l fumes from coming down the annulus i n t o t h e pump bowl. ever, some o i l (about 0.5

- 1.0

How-

cc/day) did g e t i n t o t h e pump bowl by a

leakage path around t h e s h i e l d plug.

This o i l was thermally decomposed

i n t o vapors and soot t h a t l e f t with t h e off-gas through l i n e 528.

tias

ILOWS

m c o ana

OUL

or r;ne pump

DOWL

w e r e CunLLnuous.

uaa w a s

admitted a t a fixed rate and t h e pump bowl pressure'was controlled by t h r o t t l i n g PCV-528 i n

r gas flow ( l i n e s normally vented f t h e c i r c u l a t i n g loop, All aiaine i n th

ess steel pipe.

The va

t i o n on experience.

The

ystem.

The d r i v i n g f o r c e for t h e

r e s s u r e d i f f e r e n t i a l ticross PCV-528. o l a n t d r a i n tank

e by p r e s s u r i z i n pas svstem w a s 1/2-inch Sch-40 s t a i n -

ters, and t r a p s are described i n t h e secas located i n t h e coolant d r a i n c e l l ,

which could be s a f e l y entered whenever t h e r e a c t o r power was s h u t down.

166

Fig. 8.1

Simplified Flowsheet of MSRE Coolant Off-Gas System

167

8.1.2

Fuel Off-gas System As seen i n Fig. 8.2, t h e flowsheet f o r t h e f u e l s a l t off-gas was

b a s i c a l l y similar t o t h a t f o r t h e coolant ofâ&#x201A;Ź-gas, but w a s more complex. Three d r a i n tanks were'required i n t h e f u e l system (two f o r f u e l , one f o r f l u s h s a l t ) and an overflow tank w a s required f o r t h e f u e l pump bowl. major d i f f e r e n c e was t h a t a1

u e l system off-gas w a s s e n t through charcoal

beds f o r f i s s i o n product absorption before release t o t h e stack.

The gas

hold-up and cooler consisted of a 68-ft length of 4-inch Sch-40 s t a i n l e s s s t e e l pipe.

The remainde

off-gas piping with t h e exception of t h e

f l e x i b l e jumper l i n e near t h e p a r t i c l e t r a p and l i n

bowl and t h e l i n e i n t h e v i c i n i t y of 57 w a s 1/2-in.

Sch-40 s t a i n l e s s s t e e l pipe.

Therefore, a t the normal off-gas flow rate o

-1 liters per minute, t h e

f i s s i o n gases were a t l e a s t 45 minutes old o

rrival at the particle trap

and approximately 100 minutes old on a r r i v a l a t t h e main charcoal bed. An a u x i l i a r y charcoal bed w a s provided t o accommodate l a r g e venting

rates ( f o r s h o r t i n t e r v a l s ) ,such as were required during f i l l i n g and draining t h e r e a c t o r or during s a l t t r a n s f e r s .

'c,

A sampling system was a l s o

provided t o study t h e concentrations of f i s s i o n products and o i l decompos i t i o n produces i n t h e f u e l off-gas. The f u e l off-gas system was contained and shielded and t h e r e f o r e required remote maintenance t

iques f o r work on any portion upstream of

t h e charcoal beds. 8.2

Experience with Coolant S a l t Off-gas System

A summary of t h

encountered with the coolant off-gas e s c r i p t i o n of these i

n t r o l valve (PCV-528)

l a n t system, t h e o r i g i n a l pressure t t l e d t h e off-gas from t h e coolant pump ate pressure r e l i e f and w a s replaced

by a valve having l a r g e r

cient

(q =

0.02).

168

169

Table 8.1.

Summary of D i f f i c u l t i e s with t h e Coolant Off-gas System

Description

Date

I n s t a l l e d l a r g e r control valve (PCV-528) t o 0.02)

Jan. 1965

Added a 2511 f i l t e r ahead of t h e control valve

Jan. 1965

Replaced t h e 25p f i l t e r (1.4 sq in.)

Feb. 1965

I n s t a l l e d a l a r g e r 25p f i l t e r (50 sq in.)

June 1965 Jan. 1966

I U

(Cv 0.032

Jan. 1965

â&#x20AC;&#x2122;

Control valve plugged -cleaned a 1 v f i l t e r (50 sq in.) Replaced 1 p f i l t e r

Feb. 1966

Feb. 1967

June 1967

Sept 1967

Jan. 1968

Sept 1968

t h i s and i n s t a l l e d

Restriction a t coolant pump - c l e a r e d

by backblowing

Nov. 1968

Restriction a t coolant pump - c l e a r e d by heating with a torch. I n s t a l l e d heaters on t h e l i n e . Cleaned off gas l i n e s and replaced 1 p f i l t e r .

Feb. 1969

Restriction a t coolant pump with electrical heaters.

June 1969

Control valve, 1 p f i l t e r and check valve cleaned and back-diffusion preventer removed.

Sept 1969

Replaced 1p f i l t e r .

- cleared by heating

170

A s i n t e r e d metal f i l t e r (tubular c a r t r i d g e one-inch long by 1/2-inch

diameter) capable of stopping p a r t i c l e s g r e a t e r than 2 5 w ~as installed i n the off-gas l i n e t o p r o t e c t t h e pressure c o n t r o l valve.

It w a s necessary

t o replace t h i s f i l t e r (because of excessive pressure drop) a f t e r only 24 days of service with s a l t i n t h e coolant loop.

Inspection showed t h a t

t h e plugged f i l t e r was covered with amorphous carbon containing traces of t h e coolant s a l t and INOR-8.

A second i d e n t i c a l c a r t r i d g e plugged a f t e r

20 days of service and a t the completion of Run 1, (see Table 3-2 f o r dates

of runs) i t was replaced with one having 35 t i m e s t h e surface area.

Cool-

a n t s a l t was not c i r c u l a t e d again f o r 3-1/2 months and then f o r only 118 hr.

During c i r c u l a t i o n , pressure control again became erratic which indi-

cated an obstruction of e i t h e r t h e f i l t e r o r t h e valve.

Both w e r e removed

f o r inspection, and although there was no deposit on the f i l t e r , the valve

was p a r t i a l l y obstructed by a black, granular material. Rinsing w i t h acetone restored t h e o r i g i n a l flow characteristics of t h e valve, and it w a s r e i n s t a l l e d f o r t h e next run.

The f i l t e r w a s replaced with a similar one

but capable of removing p a r t i c l e s g r e a t e r than

1 u i n diameter.

The smaller

pore size w a s chosen because the black granular material was determined t o

be a mixture of glassy spheroids of coolant s a l t approximately,lU i n di-

ameter and carbon o r carbonaceous material.â&#x20AC;&#x2122; It i s n o t known whether these materials were c a r r i e d i n t o t h e l i n e continuously o r w e r e swept o u t of t h e pump bowl by sudden venting. The valve w a s r e l a t i v e l y trouble-free a f t e r t h e i n s t a l l a t i o n of t h e 1 u f i l t e r . The f i l t e r , however, experienced gradual plugging and w a s changed o r cleaned twice i n 1966, t h r e e times i n 1967, t h r e e t i m e s i n 1968, and twice i n 1969.

The f i l t e r s , when removed

from t h e system, showed no evidence of foreign matter o t h e r than a l i g h t

f i l m of o i l i n t h e f i l t e r housing and a l i g h t f i l m of o i l w a s t r a n s f e r r e d from t h e porous metal f i l t e r onto the p l a s t i c bag used i n removing t h e filter, Flow tests on the plugged f i l t e r s showed pressure drops of less than 0.1 p s i a t flow rates two and t h r e e t i m e s g r e a t e r than the normal MSRE rate ( e s s e n t i a l l y t h e same r e s u l t s were obtained on a test of t h e f u e l off-gas filter).

Plugging i n these f i l t e r s w a s a t t r i b u t e d t o t h e l i q u e f a c t i o n of

oil i n t h e pores of t h e f i l t e r .

The f a c t t h a t t h e pressure drop (at nor-

m a l MSRE flow rates) changed from

%lo p s i

before removal t o e s s e n t i a l l y

171 zero when t e s t e d a f t e r removal w a s a t t r i b u t e d t o t h e removal of some of t h e o i l from t h e pores onto t h e p l a s t i c bag and rubber gloves during removal and i n s e r t i o n of t h e f i l t e r i n t h e flow test r i g .

The f i l t e r s

were rinsed i n acetone before reuse. On s e v e r a l occasions when t h e f i l t e r became r e s t r i c t e d and could not

be replaced during power op I

the output from t h e pressure c o n t r o l l e r

w a s switched from PCV-528 t

536.

The coolant pump pressure w a s then

controlled with t h i s l a r g e by-pass valve (Cv = 3.5).

Although t h e valve

w a s operated a t e s s e n t i a l l y i t s f u l l y open o r f u l l y closed position, adequate pressure c o n t r o l w a

achieved because t h e check valve and back-

d i f f u s i o n preventer offered During t h e May 1967 shutdown

appreciable r e s i s t a n c e t o l a r g e flows. t h e off-gas piping downstream of HCV-536 w a s

changed t o j o i n l i n e 528 downstream of PCV-528 r a t h e r than l i n e 560.

This

-change routed t h e gas by t h e coolant system r a d i a t i o n monitor so t h a t t h e coolant off-gas could be monitored when i t became necessary t o use HCV-536 t o c o n t r o l t h e system pressure.

On one occasion l a t e i n r e a c t o r operations (September 1969) when t h e

h*,

f i l t e r becaue r e s t r i c t e d , t h e coolant system pressure w a s allowed t o reach

10 p s i g before remedial ac

w a s taken.

i n t h e system meandered ba

d f o r t h between 10 and <5 p s i g f o r approxi-

During t h i s test t h e pressure

mately one month before exceeding 10 p s i and thus requiring pressure cont r o l with t h e vent valve (HCV-536).

This pressure drop i s i n agreement

with t h e gas pressure required t o enlarge o r b u r s t a f i l m of o i l from a one-micron pore i f one

t the bubble radius i s t h e same as t h a t

of t h e pore.

t i o n f o r a s p h e r i c a l bubble i s

The e q u i l

where u = s u r f a c e t e n

r = bubble radius pore =

This c a l c u l a t i o n

l l e r than 111, once plug plugged a t pressure d i f f e r o s c i l l a t i o n experienc

bid

es/cm f o r o i l s i n t h e coolant 0.5 x 10-~ em. a t any pore i n t h e s i n t e r e d metal f i l t e r densed o i l vap

1s of less than 9.'3 p s ttributed t o a n

ould remain

e slow pressure rium condition

of condensation and evaporation of o i l a t the pores of t h e f i l t e r .

172 8.2.2

R e s t r i c t i o n a t t h e Pump B o w l I n September of 1968 approximately one month a f t e r t h e s t a r t of Run

15, i n addition t o a r e s t r i c t i o n a t t h e f i l t e r , a r e s t r i c t i o n developed i n the off-gas l i n e a t the e x i t of t h e coolant pump bowl.

This r e s t r i c t i o n

w a s blown clear once by closing t h e pump off-gas and equalizer valves and

allowing t h e pump pressure t o increase t o 15 p s i g before suddenly venting t h e pump t o t h e drain tank which w a s a t atmospheric pressure.

The r e a c t o r

w a s shut down i n November of 1968 when t h i s r e s t r i c t i o n reappeared and a similar r e s t r i c t i o n developed i n t h e f u e l off-gas system.

During the shutdown, l i n e 528 w a s c u t approximately 3 f e e t downstream

of the pump bowl and a clean-out t o o l which was i n s e r t e d toward t h e pump bowl came out covered with black, heavy grease.

A brown vapor (hydro-

carbons) was collected when t h i s s e c t i o n of l i n e w a s heated with a torch. i

The r e s t r i c t i o n w a s not cleared, however, u n t i l t h e junction of t h e offgas l i n e and t h e pump bowl w a s heated t o red heat.

There w a s p r a c t i c a l l y

no evolution of vapor from t h e l i n e a t t h a t t i m e , suggesting t h a t t h e res t r i c t i o n w a s primarily s a l t o r a combination of s a l t and hydrocarbons.

A h e a t e r w a s i n s t a l l e d on t h e l i n e a t t h e e x i t from t h e pump bowl; a l s o a t r a p which w a s packed with s t a i n l e s s - s t e e l mesh w a s i n s t a l l e d i n t h e l i n e approximately s i x f e e t from t h e pump bowl i n an e f f o r t t o condense o i l vapors a t t h i s l o c a t i o n and thus prevent o r minimize o i l transport downstream t o t h e f i l t e r s and valves.

Also during t h i s shutdown a l l coolant off-gas valves i n t h e coolant d r a i n c e l l were dismantled and cleaned.

The valves and interconnecting

l i n e s and f i l t e r housing were found t o have a l i g h t l a y e r of o i l on t h e inner surface, and approximately 10 t o 20 cc of o i l w a s cleaned from low pockets i n t h e interconnecting l i n e s . Approximately s i x weeks after resumption of operations, t h e l i n e a t t h e pump bowl again showed evidence of a r e s t r i c t i o n .

The h e a t e r w a s

energized and t h e r e s t r i c t i o n cleared when t h e temperature of t h e pipe under t h e heater (normally a t 485OC) increased t o 620째C. gas problems were encountered a t t h i s location.

N o f u r t h e r off-

However, about s i x months

later r e s t r i c t i o n s appeared i n t h e pressure c o n t r o l valve, check valve, and t h e back-diffusion preventer.

During t h e June 1969 shutdown,

tbe f i l -

ter, c o n t r o l valve, and nearby check valve were cleaned and r e i n s t a l l e d ;

i173

t h e back-diffusion preventer was removed completely.

Although a l l four

components were coated with o i l which impeded off-gas flow, i t is believed t h a t the back-diffusion p r r e s t r i c t i o n i n t h e s i n t e r e d metal f i l t e r i n

With t h e exception of September 1969, no f u r t h e

oolant off-gas problems w e r

f i n a l shutdown. I n summary, t h e r e s t r i c t i o n s i n 'the coolant off-gas l i n e s were primarily r e l a t e d t o o i l leakage i n t o the pump bowl.

I n retrospect, a f i l t e r The

of an intermediate pore s i z e (say 2-4u may have been a b e t t e r choice.

r e s t r i c t i o n a t t h e pump exit w a s probably more of a s a l t mist carryover phenomenon r a t h e r than hydrocarbon r e l a t e d .

The plugging a t t h e coolant

pump bowl e x i t , although occurring less frequently, seems t o be similar t o t h a t experienced a t t h e f u e l pump which w i l l be discussed later.

8.3 A summary of t h e f u e l off-gas experience is given i n Fig. 8.3.

The

s a l t c i r c u l a t i n g t i m e s and power operations are shown and t h e type d i f f i c u l t i e s encountered are tabulated.

These are discussed in the following

sections.

8.3.1

Experience with Fuel System Off-gas System P r i o r t o Power Operation The problems encountere

i t h t h e f u e l off-gas system during t h e pre-

c r i t i c a l and low-power opera i n the coolant system. 8.3.1.1

F i l t e r an

ifficulties.

After about two months of

f l u s h s a l t c i r c u l a t i d n during t h e p r e c r i t i c a l t e s t i n g , t h e f u e l system ng t h e shutdown a t t h e end of Run 1

pressure c o n t r o l bec t h e f i l t e r and valve peared clean but t h e v own out with gas and t

both f i l t e r and v of t h e next run, t

ed.

The f i l t e r (pore diameter of ' ~ 2 5 ~ )

as p a r t i a l l y plugged.

The obstruction w a s

w a s washed with acetofie, reassembled, and Approximately a week a f t e r t h e s t a r t came p a r t i a l l y plugged.

This t i m e i t

w a s replaced with er % (0.077 instead of 0.022). The smaller valve w a s then c u t open and a black deposit of amorphous carbon and glassy beads w a s found p a r t i a l l y covering the tapered stem. The acetone

bs U

175 I

r i n s e from t h e f i r s t cleaning (Run 1) and t h e deposit removed from t h e valve during Run 2 contained 1- t o 5-11 glassy beads which proved t o have the composition of f l u s h s a l t .

The s a l t beads i n t h e off-gas l i n e s were

probably frozen droplets of m i s t produced by t h e s t r i p p e r spray i n t h e f u e l pump.

J u s t as i n t h e c

a n t system, i t is not known whether t h e

m i s t was transported continuously o r w a s swept out by sudden venting. The performance of t h e f u e l off-gas system w a s s a t i s f a c t o r y f o r t h e remainder of Run 2 and t h e zero-power exper shutdown following 'Run 3, l a r g e r s i n t e r e d m

n 3).

f i l t e r s ( f i l t e r area

0.34 f t " ) capable of removing p a r t i c l e s gre i n s t a l l e d ahead of t h e pressure control valve

During t h e

111 i n diameter were

0 t h t h e f u e l and coolant

off -gas l i n e s . 8.3.1.2

Check valve malfunctions.

The check valves i n t h e off-gas

system are spring-loaded poppet-type valves and were designed f o r low pressure drops.

During p r e c r i t i c a l checkout of t h e system when a b i t of

debris caused one such valve t o remain open, a l l t h e other s i m i l a r valves

were removed f o r inspection, cleaning, and pressure drop checks before

reinstallation.

Those which did not open with a forward pressure of a t

least 5-in. HaO and reclose with a forward pressure d i f f e r e n t i a l of 'b2-in. H a 0 were replaced with new

The next indication o

alves meeting these criteria.

a check valve malfunction w a s i n January 1966

when a r e s t r i c t i o n developed i n t h e check valve downstream of HCV-533.

The

blockage occurred a f t e r t h e f u e l system pressure w a s vented f o r a s h o r t period of t i m e through t h i s valve t o the auxiliary charcoal bed. Although some foreign material, shown i n Fig. 8.4, was found

itl

t h e poppet of t h e

check valve, i t w a s judge

t s u f f i c i e n t t o s t o p t h e gas flow (the plug

may have been dislodged d

g disassembly).

The s o f t O-ring (which makes

t h e s e a l ) and t h e cone of t h e poppet were covered with an amber varnishl i k e material which w a s i d e n t i f i e d as a hydrocarbon.'

The check valve w a s

not replaced and administrative c o n t r o l of HCV-533 w a s used t o prevent a backflow of gas i n t o t h e fuel-pump gas space during venting of t h e d r a i n tanks.

I n April 1966, d i f f i c u l t y i n venting through t h e a u x i l i a r y charcoal bed w a s found t o be due t o a check valve poppet which had become lodged i n

the i n l e t l i n e t o the bed.

Apparently t h e poppet had vibrated loose from

176

F i g . 8.4

Check Valve CV-533

17.7

-u

a check valve a t one of t h e drain tank l i n e s and had been c a r r i e d down-

stream during venting

rations.

prevent reverse flow i n the

The inab'ility of t h e check valves t o l i n e s of a l l t h r e e d r a i n tanks implies

t h a t any and a l l of these valve poppets may have vibrated loose. 8.4.

D i f f i c u l t i e s with the Fuel Off-gas System During Power Ascension

The f u e l system off-gas problems became somewhat more complex during power ascension than those encountered e a r l i e r because of f i s s i o n fragment production i n t h e reactor and subsequent transport of t h e f i s s i o n gases and some s o l i d decay daughters i n t h e off-gas l i n e s .

Some of t h e l u b r i c a t i n g

o i l which leaked i n t o t h e pump bowl (0.5 t o 1.5 g/day)

, was

polymerized t o

a varnish-like substance by t h e high radiation f i e l d i n t h e off-gas l i n e s

and components.

This material, amorphous carbon from t h e cracking process,

and s o l i d decay daughters of the f i s s i o n gases were trapped i n t h e various components of t h e off-gas system such as f i l t e r s , check valves, c a p i l l a r i e s ,

valves, t r a p s , and presumably in t h e charcoal beds.

The d i f f i c u l t i e s ex-

perienced during t h i s period are reported by run numbers o r major shutdowns. 8.4.1

Run 4 The f i r s t indication o

power l e v e l was raised t o 0 t o t a l integrated because of a r e s t r i c t i o n s t a r t of t h i s pressure t r a

t h i s type of r e s t r i c t i o n occurred when t h e

MW on January 23, 1966 g f t e r only 3 MWhrs of

system pressure began t o rise slowly i c i n i t y of PCV-522.

Shortly a f t e r t h e

t h e response of t h e drain-tank pressures

a t a capilla'ry flow r e s t r i c t o r i n l i n e 521, t h e fuel-loo f u e l system was re

alizing line.

The excess pressure i n

t i n g gas through HCV-533 t o t h e auxili-

a t PCV-522 w a s cleared ts f u l l stroke.

These opera

n i n the line t o the coal bed, apparently a t H -1/2-hr

reactor w a s made sub

operatio maintained f o r about 24 h r , but t h e

f t e r t h e reactor power was raised t o

j 178

1 MW.

Again, t h e operation of t h i s valve through i t s f u l l s t r o k e w a s ef-

f e c t i v e i n r e l i e v i n g a t least p a r t of the r e s t r i c t i o n .

Also on t h i s oc-

casion, evidence of r e s t r i c t i o n s a t t h e charcoal-bed i n l e t s began t o appear. These r e s t r i c t i o n s were bypassed by p u t t i n g t h e two s p a r e charcoal-bed sect i o n s (2A and 2B) i n s e r v i c e and closing t h e i n l e t valves t o t h e two t h a t

were r e s t r i c t e d (1A and 1B).

Within about 6 h r t h e i n l e t s of beds 2A and

When t h e previously plugged beds were checked, a t least

2B a l s o plugged.

one w a s found t o be clear and off-gas flow w a s reestablished through bed 1B.

The combination of d i f f i c u l t i e s i n t h e off-gas system r e s u l t e d i n a r e a c t o r shutdown t o correct these conditions.

The r e s t r i c t i o n s i n t h e

equalizing l i n e (521) and t h e a u x i l i a r y vent l i n e (533) were relieved by removing t h e c a p i l l a r y (FE-521) and a check valve (CV-533, reported earlier). The t h r o t t l e valve and f i l t e r assembly, and a second valve t h a t had been t r i e d b r i e f l y , w e r e removed from l i n e 522.

A l a r g e , r e l a t i v e l y coarse ( 5 0 ~ )

f i l t e r and a l a r g e , open hand valve were i n s t a l l e d a t t h e PCV-522 location.-This arrangement eliminated t h e small, e a s i l y plugged passages associated and i t appeared that t h e f i l t e r would s t i l l remove any p a r t i -

with PCV-522,

cles l a r g e enough t o plug valve 522B, which w a s t o be used f o r system pres-

s u r e control.

) , t h e flow The o r i g i n a l valve, f i l t e r (pore diam. ~ ~ l p and

element were subseguently examined at the High Radiation Level Examination Laboratory (HRLEL)

Although nothing was done t o clear the charcoal beds o t h e r than realignment*of the operating mechanisms t o the valves, successive pressure drop measurements showed that the r e s t r i c t i o n s in the beds had decreased s i g n i f i c a n t l y and before the start of Run 5 , flows through t h e beds were back t o normal.

Flow measurements through t h e a u x i l i a r y bed showed t h a t

the bed had a l s o returned t o normal.

Examination of FE-521 -The flow r e s t r i c t o r consisted of a s h o r t length of 0.08-in.

I D tubing welded i n t o l i n e 521 a t one end and coiled t o f i t

i n s i d e a 3-in.

I D container.

i n t h e container.

The other end of t h e c a p i l l a r y w a s l e f t f r e e

When t h e container was cut open, t h e entrance region w a s

found t o be clean, and only a small deposit was found on t h e container w a l l near the end of t h e discharge region as shown i n Fig. 8.5.

These i s o t r o p i c

p a r t i c l e s were characterized as p a r t l y coalesced amber globules w i t h a

179

PHOTO R27589

Fig. 8.5

Flow Restrictor Eâ&#x20AC;&#x2122;E-521.

180 r e f r a c t i v e index of 1.520.

Spectrochemical analyses indicated microgram

quantities of lithium, beryllium, and zirconium.

The r e s t r i c t o r w a s not

completely plugged when examined i n the hot cell. Examination of PCV-522 -The s t e m of the control valve shown i n Fig. 8.6A, w a s covered with an amber oil-like coating and there w a s an accumul a t i o n of o i l i n the recess formed by the bellows-to-stem

adapter.

The

tapered f l a t (flow area) on the s t e m had small accumulations of s o l i d s ; the body was coated with a s i m i l a r material and had a semisolid mass of

material on the surface near the seat port as shown i n Fig. 8.6B. scan' of t h i s material indicated the presence of "'Sr, Examination of the F i l t e r Preceding PCV-522 diameter by 15-in.

14'Ba,

- Filter-522

A gamma

and '"'La.

was a 1-in.

long cylindrical type-316 s t a i n l e s s - s t e e l sintered metal

element enclosed by a 1-1/2-in.

schedule-40 pipe.

The pore diameter of t h i s

p a r t i c u l a r f i l t e r was " 1 ~ and the element thickness w a s 1/16 in. area w a s 0.34 f t ' and t h e flow was from t h e outside i n .

The f i l t e r

The f i l t e r assem

bly was removed*on February 8, 1966 and examined a t HRLEL on t h e following day.

The upper t h i r d of -the element was steel gray i n appearance; t h e

lower two-thirds had a f r o s t y white appearance. evidence of buildup o r cake.

A t no place w a s there any

A day later t h e f r o s t y white areas had be-

come darker and tended toward the steel gray appearance.

The index of re-

f r a c t i o n of some of the material scraped from t h e s i d e s of t h e element w a s , 1,524 and a gamma scan' of t h e sample indicated t h e presence of '"'Ba,

'"La,

103

Ru, and '"Cs. The f i l t e r w a s then reassembled f o r flow tests.

The data i n d i c a t e

t h a t f o r a given pressure drop t h e f i l t e r w a s passing only 5% of the corresponding flow of a clean f i l t e r .

However, extrapolation of the data t o

the normal MSRE off-gas flow indicates that even though 95% plugged, t h e contribution of the f i l t e r (0.075 p s i measured a t t h e hot c e l l ) t o t h e t o t a l pressure drop (>5 p s i during Run 4) w a s negligible.

Exposure t o atmosphere

could have opened some of the pores i n t h e f i l t e r ; however, a more l i k e l y explanation is t h a t handling and sampling opened enough pores i n the element t o produce these r e s u l t s ,

181 d

PHGTO R278l.2

Fig. 8.6A

PCV-522 Valve S t e m

Fig. 8.6B

PCV-522 Valve Body

182 8.4.2

Run 5 S a t i s f a c t o r y manual pressure control by t h r o t t l i n g V-522B w a s demon-

s t r a t e d during t h e preceding shutdown; however, a t t h e beginning of Run 5 a f t e r only 2-1/2 hours of power operation (1 MW), t h i s valve showed evidence of rapid plugging and required constant adjustment u n t i l i t w a s i n the fully-open position f i v e hours later.

During one such adjustment,

rapid plugging of t h e charcoal beds occurred.

A t t h e conclusion of s a l t

c i r c u l a t i o n (terminated by a space-cooler motor f a i l u r e ) , t h e r e s t r i c t i o n

a t the i n l e t t o charcoal bed 1 B could not be cleared; therefore, t h e i n l e t valve (HV-621) w a s removed f o r examination a t t h e HRLEL.

With t h e valve

out, t h e pressure drop across the bed w a s much lower but i t w a s s t i l l higher than f o r a normal u n r e s t r i c t e d bed.

An excess of helium (forward blow) w a s

forced through the bed and the pressure drop suddenly decreased t o normal. The cone of t h e i n l e t valve t o t h i s bed (shown i n Fig. 8.7) w a s shiny as though w e t with an o i l - l i k e material.

The s m a l l metallic chips near t h e

l a r g e end of t h e cone were a r e s u l t of t h e sawing operation.

There w a s

some white amorphous powder on t h e tapered s e c t i o n bf t h e valve t r i m and appeared t o be adhered f a i r l y strongly t o t h e m e t a l surface.

A similar

The material removed from t h e s t e m

material was found i n t h e valve body.

w a s described as an i s o t r o p i c , f a i n t l y colored material, varnishlike i n appearance with a pebbly surface.

was

132

The predominant isotope i n t h e material

The r e f r a c t i v e index of t h e varnishlike material w a s 1.526

Te.

COP

pared w i t h 1.50 f o r a d i s t i l l e d f r a c t i o n of t h e l u b r i c a t i n g o i l used i n t h e f u e l pump. 8.4.3

Ekp eriments and Alterations DurinR t h e March-1966 Shutdown

Off-gas samples taken while t h e reactor w a s shut down showed an increasing hydrocarbon content i n the gas as t h e r e a c t o r c e l l temperature w a s increased, lending support t o t h e hypothesis t h a t t h e r e w a s a r e s e r v o i r of hydrocarbons i n the holdup-volume portion of t h e off-gas l i n e . practicable t o clean t h e 68-ft-long,

4-in.-diam

It w a s not

pipe, so t h e off-gas l i n e

w a s disconnected a t t h e f u e l pump and i n t h e vent house; l a r g e q u a n t i t i e s of helium w e r e blown through t h e l i n e i n t h e forward and reverse d i r e c t i o n s a t v e l o c i t i e s up t o 20 t i m e s normal.*,â&#x20AC;&#x2122;

Very l i t t l e v i s i b l e material w a s

183

PHOTO R28500

F i g . 8.7

Charcoal Bed I n l e t Valve S t e m HV-621

collected on f i l t e r s a t t h e ends of the l i n e , but t h e r e w e r e f i s s i o n prod-

ucts, and t h e amount doubled when the cell w a s heated from 50 t o 8OOC. V i s u a l observation showed t h a t the head end of t h e holdup volume w a s clean

except f o r a barely perceptible d u s t l i k e film.

A thermocouple w a s attached

t o the holdup pipe near t h e head end f o r monitoring temperatures during power operation.

(When the power w a s subsequently r a i s e d , t h e temperature

indeed r o s e from c e l l a i r temperature of about 55°C a t zero power t o about 113OC a t 7.5 MW.

The temperature increase occurred with a t i m e constant of

about 30 min which w a s not inconsistent with buildup of gaseous f i s s i o n products i n the l i n e . )

As a r e s u l t of extensive laboratory tests on organic vapor t r a p s , 7 t h e f i l t e r - v a l v e assembly w a s replaced with a p a r t i c l e t r a p and an organic vapor t r a p i n series and upstream of the pressure c o n t r o l valve (PCV-522) whose flow c o e f f i c i e n t was increased t o 0.865.

The Mark I p a r t i c l e trap,’

shown i n Fig. 8 . 8 w a s designed t o remove p a r t i c u l a t e s and mist.

G a s from

t h e fuel-pump bowl entered a t t h e bottom of t h e u n i t through a c e n t r a l pipe, reversed flow in t h e Yorkmesh, and passed i n succession through two conc e n t r i c cylinders of porous m e t a l ( f e l t metal) and a bed of inorganic f i b e r s . The f i r s t f e l t metal f i l t e r w a s capable of stopping 96.7% of p a r t i c l e s

greater than 0 . 8 ~and the second -99.4% p a r t i c l e s g r e a t e r than 0 . 3 ~ . The organic vapor trap’ consisted of 13 f e e t of 1-in. sch.-40 s t a i n l e s s -

steel pipe, arranged i n three hairpin sections of approximately equal lengths. The vapor t r a p w a s loaded with 1092 gm of Pittsburgh PCB charcoal and instrumented w i t h thermocouples a t 5, 12, 51, 59, 105, and 113 inches, respectively, from the bed i n l e t .

As a r e s u l t of the experience with and examination of HV-621 and o t h e r off-gas components, the i n l e t valves t o the mafn and a u x i l i a r y charcoal beds and V-522A were replaced w i t h ones having l a r g e r flow c o e f f i c i e n t s .

8.4.4

R u n s 6 and 7 A t t h e s t a r t of Run 6, the pressure drops ( a t normal flow rate) across

the p a r t i c l e t r a p , organic vapor t r a p , and charcoal beds (sections IA and 1 B i n p a r a l l e l service) w a s ~ 0 . 0 5 , 0.7, and 1.6 p s i respectively.

The off-

gas system performance w a s s a t i s f a c t o r y during t h e f i r s t t e n days of R u n 6 w i t h the r e a c t o r operating a t 1 MW o r less.

However, when t h e power was

he,

ORNL-DWG -6-11444R

FINE METALLIC FILTER FIBERFRAX (LONG FIBER) 7

COARSE METALLIC F I L T E R 1

I-0WER WELD3

.......... .. ............. ~~~~

SAMPLE 97

&y-.

-SAMPLES 5B AND 6

SAMPLE 3 A STAINLESS STEEL MESH

Fig. 8.8

THERMOCOUPLE (TYP)

NICKEL BAFFLE ( 8 )

Line 522 Particle Trap (Mark I) and Location of Sampling Points During Subsequent Examination

186 increased t o 2.5 MW, t h e pressure drop across t h e charcoal beds a l s o increased.

Pressurization and equalization experiments established t h a t t h e

r e s t r i c t i o n s were a t t h e i n l e t s t o t h e beds, probably a t t h e packing of steel wool above t h e charcoal.

It was found t h a t t h e pressure drop could

be reduced, usually t o near t h e normal 1.5 p s i , by blowing helium a t 35 p s i g e i t h e r forward o r backward through t h e bed; back-blowing seemed t o be more e f f e c t i v e .

Back-blowing of t h e beds w a s done whenever t h e pressure

drop of two s e c t i o n s i n p a r a l l e l approached 3 psi.

Later, higher pressure

Plugging occurred

drops were t o l e r a t e d before back-blowing was i n i t i a t e d .

when t h e power w a s r a i s e d during t h e approach t o f u l l power with s e c t i o n 1 B plugging more o f t e n than t h e others.

Plugging became less frequent later,

but a t t h e end of Run 6, it w a s s t i l l necessary t o back-blow t h e beds about once a week. P a r t i c l e Trap Performance.

After approximately 1 0 days of operation

a t t h e 5-MU l e v e l , the pressure drop across the p a r t i c l e t r a p s t a r t e d to

increase and reached 6 p s i approximately a week later.

When t h e power was

increased t o 7.5 MW, t h e pump pressure increased t o 9 p s i i n less than 12 hours a t f u l l power.

Since t h e pressure drop across t h e organic vapor t r a p

w a s less than 0.5 p s i and t h a t across t h e charcoal bed w a s kept below 3 p s i by periodic back-blowing,

t h e v a r i a t i o n i n pump-bowl pressure w a s i n t e r -

preted as a corresponding r e s t r i c t i o n i n t h e p a r t i c l e t r a p .

About t h r e e

hours a f t e r an unscheduled power i n t e r r u p t i o n , t h e fuel-pump pressure decreased t o 6.5 psig and t h e following day i t decreased t o 4.3 psig.

How-

ever, a few hours a f t e r t h e r e a c t o r power w a s again r a i s e d t o 7.5 MW, t h e pump pressure increased t o 8 psig i n d i c a t i n g a r e s t r i c t i o n i n t h e p a r t i c l e trap.

A t t h i s time t h e helium purge t o t h e pump s h a f t was decreased from

2.4 t o 1.9 liters per minute i n an e f f o r t t o reduce t h e fuel-system pressure.

The r e a c t o r w a s then taken s u b c r i t i c a l f o r two days.

When t h e power

w a s again increased t o 5 MW, t h e pump pressure increased from 3 p s i g a t zero power t o 8 p s i g and eventually t o 12 p s i g a t f u l l power.

The pump

pressure gradually returned t o t h e 6 t o 8 psig range a f t e r approximately

a week a t f u l l power.

After Run 6 w a s terminated by a component cooling

pump f a i l u r e on May 28, 1966, t h e p a r t i c l e t r a p pressure drop aecreased t o

less than 1 p s i and it remained r e l a t i v e l y u n r e s t r i c t e d f o r a month a f t e r full-power operation was restored i n Run 7.

The pump pressure then gradu-

a l l y increased t o 8.7 p s i g before another power reduction

m-1 f a i l u r e )

187

occurred ending Run 7.

A few hours a f t e r t h e power reduction, t h e pump

pressure returned t o a normal value of 3 p s i g and remained a t t h i s pressure f o r approximately a

t h of zero-power operation before t h e r e s t r i c t i o n

gradually returned.

I n summary, pump pressure increases ( r e s t r i c t i o n s i n t h e p a r t i c l e

traps) were usually associated with periods of power increases.

Pressure

decreases were sometimes unexplained but usually were associated with power decreases o r with d e l i b e r a t e

tempts t o clear t h e t r a p such as reverse-

blowing with helium. Rerouting of Line 524.

H e l i u m , a t a rate of 2.4 2/min, w a s introduced

through l i n e 516 i n t o t h e fuel-pump s h a f t annulus j u s t below t h e lower s h a f t The l a r g e r p a r t of t h i s flow goes down t h e s h a f t i n t o t h e pump

o i l seal.

bowl thus preventing f i s s i o n gases and s a l t m i s t from entering t h i s radia t i o n s e n s i t i v e region of t h e pump.

The smaller portion of t h i s flow

( c O . 1 2/min) goes up t h e s h a f t t o prevent o i l vapors from migrating t o t h e

f u e l . s a l t ; i t a l s o a i d s i n transporting any o i l seal leakage t o t h e o i l catch tank.

The d r i v i n g f o r c e f o r t h e smaller flow ( l i n e 524) is t h e pres-

Sure d i f f e r e n c e between t h e f u e l pump and l i n e 522 downstream of t h e f u e l system pressure c o n t r o l valve (PW-522).

However, when PCV-522 was opened

f u l l y (because of r e s t r i c t i o n s i n t h e charcoal beds), l i n e 524 had no press u r e d i f f e r e n t i a l and consequently no gas flow; therefore, it w a s rerouted t o e n t e r t h e off-gas

V-557B.

l i n e a t V-560A downstream of t h e main charcoal bed and

V-557B w a s then t h r o t t l e d t o c o n t r o l system pressure.

On two oc-

casions i n Run 6, sudden increases i n pump pressure ( a t t h e completion of

salt recovery from t h e

tank) caused a gas flow reversal a t t h e

pump s h a f t and gaseous

was c a r r i e d i n t o l i n e 524.

bypassed t h e charcoal bed,

as flow had very l i t t l e decay t i m e .

consequence, gaseous act

iggered t h e closure of t h e r a d i a t i o n block

valve on t h e main off-ga

up krypton and xenon f o

days,- respectively.

t e d of 9 f t of 3-in.

2-1/2 and 30

sched.-10

stainless-

On a t least four

steel pipe loaded with

i t t s b u r g h PCB charcoal.

occasions i n t h e subsequen

s s i o n product a c t i v i t y was blown o r dif-

fused up t h e pump s h a f t annulus.

As a

Therefore i n June of 1966, a small charcoal

bed' w a s added t o l i n e 524't The bed

Since t h i s l i n e

Although t h e a c t i v i t y l e v e l i n t h e o i l

catch tank and l i n e 524 increased, t h e r e w a s e s s e n t i a l l y no a c t i v i t y re-

lease, thus proving t h e effectiveness of the added charcoal bed.

However,

188 t h e last two releases from t h e pump bowl caused t h e flow element i n l i n e

524 t o plug p a r t i a l l y and then completely. The plug w a s found t o be i n t h e s i n t e r e d s t a i n l e s s - s t e e l d i s c at t h e i n l e t t o t h e matrix-type flow element.

The element was replaced with a c a p i l l a r y tube.

8.4.5

Experiments and A l t e r a t i o n During t h e June-September Shutdown (1966)

A t t h e conclusion of Run 7, it was decided t o f l u s h o u t t h e r e s i d u a l

f u e l i n t h e overflow tank and t o check t h e indicated l e v e l a t which overflow occurred.

Because of an insufficiency of s a l t i n t h e d r a i n tank, pres-

s u r i z i n g gas from t h e d r a i n tank entered t h e r e a c t o r and flooded t h e pump bowl with s a l t , thus binding t h e pump s h a f t with frozen s a l t and pushing

s a l t i n t h e gas l i n e s a t t h e top of t h e pump bowl. The frozen s a l t w a s cleared from t h e reference bubbler l i n e with t h e

use of remotely-applied e x t e r n a l heaters; s a l t i n t h e sampler l i n e was melted by t h e same technique but i t ran down and r e f r o z e a t t h e j u n c t i o n with the pump bowl.

This obstruction and t h e frozen s a l t i n the annulus

around t h e pump s h a f t w e r e cleared when t h e pump bowl w a s r e f i l l e d and heated t o 650OC.

Although some s a l t entered t h e main off-gas l i n e as indi-

cated by a temporary rise i n temperature a t TE-522-2 (Fig. 8 . 9 ) , pressure drop measurements showed no s i g n i f i c a n t d i f f e r e n c e from t h e clean condition. This w a s a t t r i b u t e d t o t h e b l a s t of compressed helium from t h e d r a i n tank I

t h a t w a s released backward through t h e off-gas l i n e , before t h e s a l t had time t o f r e e z e completely, when t h e o v e r f i l l triggered an automatic drain.

Therefore t h e only work done on the main off-gas l i n e (522) during t h i s shutdown w a s t o replace t h e s h o r t , f l e x i b l e "jumper" s e c t i o n of t h e l i n e , where t h e convolutions would be expected t o hold some s a l t . The p a r t i c l e t r a p was a l s o removed from l i n e 522 and sent t o HRLEL f o r d e t a i l e d examination.

An i d e n t i c a l replacement (Mark I) w a s i n s t a l l e d t o

permit operation while t h e o r i g i n a l was examined and a new one designed and constructed.

Also during t h i s shutdown, a remotely replaceable h e a t e r w a s designed and i n s t a l l e d on t h e i n l e t s e c t i o n of t h e a u x i l i a r y charcoal bed (ACB).

8.4.6

Runs 8, 9, and 10 With t h e resumption af operations t h e off-gas flowed f r e e l y , with no

unusual pressure drop f o r 26 days of c i r c u l a t i n g helium, f l u s h s a l t , and

189

ORNL-OWG 67

4764

SS HOLDUP ME, 4 in. D I N

OVERFLOW TANK -c

TO PARTICLE TRAP AND CHARCOAL BEDS

Fig. 8.9

Off-Gas Piping Near Fuel Pump and Overflow Tank

190 f u e l s a l t a t low power.

Then, two days a f t e r power operation w a s resumed

a t 5.8 MW, a plug developed i n l i n e 522 somewhere between t h e pump bowl and the junction of the overflow tank vent with t h e 4-in.

holdup l i n e .

The

f i r s t indication was a decrease over a few hours from 107OC t o 7l0C a t TE522-2, as t h e plug caused t h e off-gas t o bypass through t h e overflow tank (OFT).

The presence of the plug was confirmed when HCV-523 w a s closed t o

build up pressure i n t h e OFT t o r e t u r n s a l t from t h e tank t o t h e pump bowl; pressure i n the pump bowl a l s o b u i l t up.

E f f o r t s t o remove t h e r e s t r i c t i o n

by applying a 10-psi d i f f e r e n t i a l e i t h e r forward o r backward were unsuc-

cess f u l

The bypassing of off-gas through t h e OFT d i d not hinder operations except f o r one s p e c i f i c job:

recovery of s a l t from t h e overflow tank.

S a l t slowly but continuously accumulated i n t h e tank during t h e e n t i r e operating l i f e of the MSRE, and it was therefore e s s e n t i a l t o r e t u r n s a l t t o t h e pump bowl two o r three t i m e s a week i n order t o maintain proper

levels.

With a plug i n t h e off-gas l i n e of t h e pump bowl, i t was necessary

t o g r e a t l y reduce helium flows i n t o t h e pump so t h a t t h e overflow tank pressure could be increased f a s t e r than t h a t i n t h e pump bowl t o make t h e salt transfer.

Through t h e remainder of Run 8 , s a l t w a s returned (burped)

from t h e OFT six t i m e s , and on a t least four of t h e s e occasions some f i s s i o n product a c t i v i t y w a s blown o r diffused up t h e pump s h a f t annulus i n t o t h e o i l c o l l e c t i o n space.

This w a s a consequence of t h e reduced helium purge

down the s h a f t annulus and t h e unavoidable, sudden pressurization of t h e pump bowl t h a t occurred a t t i m e s i n the procedure. After "Run 8 w a s terminated, s t e p s were taken t o clear t h e plug from the off-gas l i n e so that the normal s a l t recovery procedure could be used. Frozen s a l t w a s suspected as the cause of t h e plug, so t h e f u e l loop w a s flushed t o reduce r a d i a t i o n l e v e l s , t h e r e a c t o r c e l l was'opened, and specia l l y b u i l t electrical heaters were applied t o t h e l i n e between t h e pump bowl and the f i r s t flange.

Heating alone did not clear t h e plug, but when,

with t h e l i n e hot, 10 p s i w a s applied backward across t h e plug, i t blew through.

The pressure drop came down as more helium w a s blown through un-

t i l i t became indistinguishable from t h e normal drop i n a clean pipe.

I n Run 9, t h e power operation w a s begun 8 h r a f t e r f u e l c i r c u l a t i o n had commenced, but TE-522-2

came up t o only 66"C, i n d i c a t i n g t h a t t h e l i n e

.LJ

191

w a s again plugged.

While t o o l s and procedures were being devised, t h e re-

a c t o r w a s kept i n operation, but g r e a t care w a s taken t o avoid g e t t i n g f i s s i o n products o r s a l t spray lowering t h e power t o 10 kW

the-pump s h a f t annulus again.

This e n t a i l e d

4 h r before t h e overflow tank was t o b e emp-

t i e d , then stopping t h e pump 4 h r beforehand t o l e t t h e s a l t m i s t settle. During t h e s a l t t r a n s f e r , t h e f u e l pump was vented through t h e s a m p l e r and a u x i l i a r y charcoal bed.

After t h r e e cycles of t h i s , t h e r e a c t o r w a s drainel

and flushed again i n preparation f o r working on t h e off-gas l i n e . This t i m e heat w a s applied t o t h e s h o r t s e c t i o n of l i n e between t h e second flange and t h e top of t h e 4-in.

decay pipe.

When heating t o about

60OOC d i d not open t h e l i n e , t h e f l e x i b l e jumper w a s disconnected t o permit c l e a r i n g t h e obstruction mechanically. t h e 1/2-in.

I n t h e flange above t h e 4-in.

line,

bore was completely blocked, but t h e weight of a c h i s e l t o o l

broke through what appeared t o be only a t h i n c r u s t of s a l t .

Borescope

inspection showed t h a t t h e rest of t h e v e r t i c a l l i n e w a s p r a c t i c a l l y clean, and t h e r e w a s only a t h i n l a y e r of s a l t i n t h e bottom of t h e horizontal run of t h e 4-in.

(Ced

pipe.

Helium was blown thraugh t h e l i n e a t f i v e t i m e s t h e

normal flow, and t h e pressure drop indicated no r e s t r i c t i o n .

The flange

near t h e pump bowl contained a similar plug which w a s e a s i l y broken. 1/4-in.

f l e x i b l e t o o l w a s then i n s e r t e d a l l t h e way i n t o t h e pump bowl t o

prove t h a t a good-sized p

existed.

A new jumper l i n e was i n s t a l l e d

and operation was resumed. Because obtaining samples remotely wxthout spreading contamination would have been most d i

ses w e r e made of t h e material i n

t h e flanges.

had frozen i n t h e li

But i t ap

p l e t e l y blocking i t , d u r i n off-gas stream during s During t h e J u l y cleanout, pipe, b u t l e f t behind a t h i n

e r f i l l of J u l y 1966.

rs apparently melted t

t h e flanges w e r e mechan

d reassembled

off-gas system operated

r i n g Run 10.

8.4.7

*m-I P a r t i c l e

Material i n t h e

on then plugged t ge of s a l t i n t h e co

almost com-

s m a l l passages.

s a l t o u t of t h e

er flanges.

After

t h i s p a r t of t h e

Trap

The second p a r t i c l

ugh Runs 8, 9 , and 10.

This u n i t

behaved i n Runs 8 and 9 much as hadâ&#x20AC;&#x2122;the f i r s t t r a p ; t h e pressure drop oc-

casionally b u i l t up t o 5 t o 10 p s i , beginning two days a f t e r power operations

192 s t a r t e d i n Run 8.

Back-blowing with helium-was e f f e c t i v e i n reducing t h e

pressure drop t o 2 t o 4 p s i ; however, i n Run 10, a f t e r t h e f i r s t week of power operation, back-blowing became i n e f f e c t i v e .

Various tacticse were

used t o g e t f i s s i o n gas t o t h e p a r t i c l e t r a p w i t h as l i t t l e delay as possib l e , t o see i f increasing t h e f i s s i o n product heating i n t h e t r a p would drive off t h e material a t t h e r e s t r i c t i o n .

After t h i s proved t o be inef-

f e c t i v e and recognizing t h a t heating caused t h e central i n l e t tube t o expand f a r t h e r i n t o t h e Yorkmesh f i l t e r i n t h e t r a p , t h e opposite approach

was used.

Eight hours of delay w a s obtained by routing t h e gas through an

equalizing l i n e t o t h e empty d r a i n tank, through t h e tank and s a l t f i l l

lines t o t h e o t h e r tank where i t bubbled through several inches of s a l t heel, then out through t h e drain-tank vent l i n e t o t h e p a r t i c l e t r a p .

The

pressure drop across t h e t r a p w a s 16 p s i when t h e gas w a s f i r s t rerouted, but within a few hours it was below 2 p s i .

When t h e o r i g i n a l route w a s

again t r i e d , the r e s t r i c t i o n began t o b u i l d up almost immediately; therefore, t h e delayed route w a s used u n t i l t h e end of Run 10 which w a s terminated f o r in-cell r e p a i r s and a l s o because a newly constructed p a r t i c l e t r a p (Mark 11) based on the Hot C e l l examination of t h e f i r s t t r a p w a s

ready f o r i n s t a l l a t i o n . Examination of MK-I P a r t i c l e Trap.'

Flow tests on t h e f i r s t t r a p ( i n

service from April through July of 1966) indicated t h a t (1) t h e pressure drop a f t e r service w a s about 20 t i m e s the "clean" pressure drop.

Assuming

o r i f i c e type flow, t h i s would represent a reduction i n flow area of about 80%.

(2) The bulk of t h e pressure drop occurred in t h e Yorkmesh and Felt-

metal s e c t i o n s of t h e f i l t e r .

The Fiberfrax s e c t i o n w a s e s s e n t i a l l y clean.

The area of t h e Yorkmesh which had been immediately below t h e i n l e t pipe w a s covered with a blue-gray t o black m a t which had completely f i l l e d t h e space between t h e wires of t h e mesh (Fig. 8.10).

The shape of t h e m a t

corresponded t o the bottom of t h e i n l e t pipe, and i t is l i k e l y t h a t t h i s

material w a s t h e major r e s t r i c t i o n t o gas flow while i n service. Since t h e i n l e t pipe temperature probably increased s e v e r a l hundred degrees during power operation of t h e r e a c t o r , i t is believed t h a t t h i s r e s t r i c t i o n behaved similar t o a thermal valve.

This could account f o r t h e

unexpected increase i n pressure drop while a t power and t h e decrease i n pressure drop when t h e power was reduced o r when t h e off-gas w a s delayed

via t h e d r a i n tank.

193

PHOTO R32824

Fig. 8.10

Inlet to Yorkmesh Section MSRE Particle Trap Mark I

194 ~

A r a d i a t i o n survey made around the outside s u r f a c e of t h e lower s e c t i o n

of t h e t r a p gave readings ranging from 6900 t o 8000 R/hr.

The probe indi-

cated 11,400 R/hr when inserted i n t o the p o s i t i o n formerly occupied by t h e

i n l e t pipe.

The r a d i a t i o n l e v e l dropped off t o 4100 R/hr a t t h e bottom of

t h e trap. The bottom of the i n l e t pipe had a deposit of blackish material which corresponded t o t h a t i n t h e Yorkmesh.

The i n s i d e of t h e i n l e t pipe

a t t h e upper end of t h e bellows appeared clean and f r e e of deposit.

The

e x t e r i o r of t h e bellows had some of the light-yellow powdery material on a background of dark brown.

When t h e Yorkmesh was removed, i t was found t h a t t h e surface of t h e outer wires of the mesh bundle was covered with a t h i n l a y e r of ambercolored organic material.

Much of t h i s material evaporated from t h e h e a t

of t h e floodlamp used t o make the photographs.

As t h e bundle w a s unrolled,

t h e color of the f i l m on the wire changed from amber t o brown t o black near the center.

or 3.

The black material was thicker than the wire by a factor of 2

This material w a s b r i t t l e , as w a s the w i r e , and much of i t came

loose as t h e w i r e w a s flexed.

Samples of t h i s material, designated Nos.

5B and 6 (Fig. 8 . 8 ) w e r e taken f o r examination and chemical analysis. M e t allographic examination of a piece of t h e w i r e covered with t h e black mat e r i a l showed t h a t t h e wire w a s heavily carburized with a continuous network of carbide i n t h e grain boundaries.

There w a s no evidence of melting

of the w i r e ; however, t h e grain growth and o t h e r changes indicated opera t i n g temperatures of a t least 650OC.

The nonmetallic deposit observed on

t h e w i r e mesh w a s apparently of a carbonaceous nature and appeared t o have been deposited i n layers.

These "growth rings" were probably t h e r e s u l t

of off-gas temperature, r e a c t o r power, and gas flow-rate changes.

The perforated p l a t e of t h e coarse f i l t e r s e c t i o n and t h e lower flange of t h e f i l t e r assembly were covered with a s t r a t i f i e d s c a l e (view A-A, Fig. 8 . 8 ) .

The colors varied from a very l i g h t yellow t o orange.

stratum i n t h e lower flange area appeared gray, almost black.

One

A sample

(No. 9) was taken of t h e light-colored material, including some of t h e black material.

The perforated p l a t e of t h e f i n e f i l t e r s e c t i o n w a s covered

with a t h i n , dark-brown coating, which seemed t o be evenly d i s t r i b u t e d over t h e surface of the plate.

The inner surface of t h e outer w a l l of t h e t r a p

w a s covered with a light-amber coating, which w a s a l s o evenly d i s t r i b u t e d .

195

It is believed t h a t these coatings were deposited by condensation and sputt e r i n g of t h e o i l from t h e adjacent f i l t e r and t h a t the dark brown color i n d i c a t e s t h a t t h e porous metal screen had operated a t a much higher temperature than had the

t e r - w a l l (during operation t h e t r a p w a s immersed

a tank of water f o r of t h e upper flange ( v i appearance of organic r

g by n a t u r a l convection).

The lower surface

g. 8.8) contained a deposit which had t h e e deposit w a s amber colored, and t h e frac-

tured edge (Fig. 8.11) gav

impression that the material was b r i t t l e .

There i s as y e t no expl

or formation of t h i s deposit o r how i t came location, The r a d i a t i o eve1 on t h e out-

t o be formed i n t h i s pa s i d e of t h i s s e c t i o n of The material a t

r w a s 200 t o 360 R/hr. of t h e Fiberfrax s e c t i o n showed an o i l -

l i k e discoloration, b

evidence of any s i g n i f i c a n t accumula-

of material.

weights of t h e d i f f e r e n t l a y e r s with

t h e weights of t h e materia than 0.2 g.

g i n a l l y loaded indicated changes of less

An i n t e r e s t i n g observation r e l a t e s t o t h e very low radioac-

t i v i t y l e v e l of t h e Fiberfrax a t t h e entrance s e c t i o n which is separated

only'by t h e Feltmetal f i

an area containing material with a c t i v i t y

l e v e l s of thousands of

only d e t e c t a b l e a c t i v i t y (above t h e exami-

nation c e l l background It is p r

f r a x section.

r) was,at t h e discharge end of t h e Fibert t h i s a c t i v i t y r e s u l t e d from back-blowing

t h e t r a p or from pressure t r a n s i e n t s which could have c a r r i e d gaseous decay products from t h e vapor t r a p back upstream and i n t o t h e p a r t i c l e trap. Even so, t h e a c t i v i t y 1 nly 1.8 R/hr above background. Analytical Results. cations (Fig. 8.8) were

1 of four samples from t h r e e d i f f e r e n t lo= t o a v a r i e t y of a n a l y t i c a l tests. The

samples were i d e n t i f i e d Taken from

Sample No.

3A 9

5B 6

oarse s e c t i o n of porous ale on lower flange of t a t i n l e t t o Yorkmesh t a t i n l e t t o Yorkmesh

For s e c t i o n s of

metal filter porous metal s e c t i o n section section

s found t h a t about t h e same weight 10s

(0.2% of sample weight) r e s u l t e d from heating t o 6OOOC i n helium as from

dipping i n a trichlorethylene bath.

The material removed by heating w a s

196

F i g . 8.11

Upper End of Porous Metal S e c t i o n

- MSRE

P a r t i c l e Trap Mark I

.._ . -"

197

.-a*:<

'*'1

cold trapped and found t o "e e f f e c t i v e l y Lxontaminated; however, t h e trichlorethylene wash w a s contaminated with f i s s i o n products. Samples 5B and 9 were compared f o r low-temperature v o l a t i l e s ; a t 15OoC, No. 9 l o s t 5% and No. 5B l o s t none.

When r a i s e d t o 6OO0C, t h e weight l o s s e s

were 35% f o r sample 5B and 32% f o r sample 9.

Analysis of t h e carbon content

gave none f o r sample 9 and 9% f o r sample 5B.

This i n d i c a t e s t h a t sample 9

had not reached as high a temperature as had sample 5B. The mass spectrographic a n a l y s i s of sample 6 indicated t h a t t h e r e w a s a very high f r a c t i o n of f i s s i o n products.

w t % Ba, 15 w t % S r , and 0.2 w t % Y.

These are estimated t o be 20

I n t h e same a n a l y s i s t h e s a l t con-

s t i t u e n t s B e and Z r were estimated t o be 0.01 and 0.05 w t % respectively. I n addition t h e material i n samples 5B and 6 contained s m a l l quantities of C r , Fe, and N i , while sample 9 d i d not.

The r e l i a b i l i t y of t h e s e

values w a s compromised by d i f f i c u l t i e s caused by t h e presence of organics and small s e c t i o n s of w i r e i n t h e sample. The gamma-ray spectrographic work indicated t h e presence of t h e following isotopes:

"'cs,

Sr,

109

Ru o r "'Ru,

'lomAg,

Nb, and '"La.

A l l t h r e e samples were chemically analyzed f o r B e , and t h e l e v e l w a s

below t h e d e t e c t a b l e l i m i t of 0.1%. Attempts t o analyze f o r Z r were comp l i c a t e d by t h e presence of l a r g e q u a n t i t i e s of Sr. The f r a c t i o n of s o l u b l e hydrocarbons w a s determined using CS1, and t h e values were 5B, 60%; 6 , 73%; and 9, 80%. The e x t r a c t s o l u t i o n s from samples 5B and 6 were allowed t o evaporate, and a few milligrams of t h e residue w a s mounted between s a l t c r y s t a l s for i n f r a r e d analysis.

T h e sam-

p l e s were i d e n t i c a l and were c h a r a c t e r i s t i c of long-chain hydrocarbons.

There w a s no evidence of any f u n c t i o n a l groups o t h e r than those involving carbon and hydrogen, bonds.

Nor was t h e r e any evidence suggesting double o r t r i p l e

There w a s an i n d i c a t i o n of a possible mild cross-linkage.

It is

l i k e l y that t h e r e is more cross-linkage of t h e organic i n t h e gas stream than appeared i n t h e s e samples, and t h e low i n d i c a t i o n could be due t o t h e i n s o l u b i l i t y of t h e cross-1

d 'organic and t h e high operating temperatures

of t h e w i r e mesh, which would cause breakdown of t h e organic i n t o elemental carbon and v o l a t i l e s .

MK-I1 P a r t i c l e Trap

8.4.8

As a r e s u l t of t h e operating experience with and d e t a i l e d examination of t h e first p a r t i c l e t r a p , t h e MK-I1 trap,'

(shown i n Fig. 8.12) w a s de-

signed w i t h t h e following features:

1. The t r a p housing was increased from 4 t o 6-in. I D , r e s u l t i n g i n an increase i n cross-sectional area of 225% i n both t h e Yorkmesh and Fiberf r a x sections. 2.

The u n i t was i n e f f e c t turned upside down so t h a t t h e Yorkmesh

s e c t i o n is a t t h e top of t h e unit and t h e Fiberfrax s e c t i o n is a t t h e bottom.

This change permits heating of t h e Porkmesh s e c t i o n (using beta decay

heat and lowering t h e water l e v e l ) while s t i l l maintaining cooling on t h e other two sections.

The d i s p o s i t i o n of t h e Yorkmesh w a s modified t o provide increased

f r o n t a l area i n t h e d i r e c t i o n normal t o t h e flaw.

Since t h e f i r s t t r a p had shown very l i t t l e loading i n t h e f i n e

Feltmetal s e c t i o n , only t h e coarse Feltmetal w a s used. 5.

The t o t a l f i l t e r area was increased from 22 in' f o r t h e o r i g i n a l

t o 288 i n 2 f o r t h e new trap.

The depth of the F i b e r f r a was reduced by 50% 7. The pressure 'drop a t 15 !t/m was less than 1 i n . HaO and t h e t r a p efficiency w a s 99.9X f o r p a r t i c l e s g r e a t e r than 0 . 8 ~ . 6.

8.4.9

Organic Vapor Trap It was expected t h a t accumulation of organic material i n t h e charcoal

t r a p immediately downstream of t h e p a r t i c l e t r a p would result i n progressive poisoning along t h e length of the trap.

Such poisoning would s h i f t t h e lo-

cation of maximum f i s s i o n product adsorption and produce a s h i f t i n t h e temperature p r o f i l e of t h e trap.

Except f o r t h e upward s h i f t due t o in-

creased power l e v e l , t h e b a s i c shape of t h e temperature p r o f i l e d i d not change, i n d i c a t i n g t h a t no s i g n i f i c a n t poisoning by organics occurred. Since the pressure c o n t r o l valve (PCV-522) w a s operated i n the f u l l y opened position and i t appeared t h a t nothing would be l o s t by eliminating t h e charcoal t r a p , both were removed t o make room f o r two new MK-I1 p a r t i -

cle t r a p s which were i n s t a l l e d i n p a r a l l e l i n January 1967.

ORNL- m

/YORKMESH ,.-T.E.

[FELTMETAL

\FlElERFfW

NKXEL

M Fig.

8.12

Line 522 Particle

4 --

I INCHES

Trap Mark II,

3 I

Particle

Trap Subassembly

6?- 4766

/-T.’

BAFFLES

200 8.4.10

Charcoal Bed Performance

After t h e removal of t h e check valve poppet i n A p r i l 1966, venting through t h e a u x i l i a r y charcoal bed was uneventful through t h e approach t o power operation. July.

However, an i n t e r m i t t e n t r e s t r i c t i o n w a s noted e a r l y i n

During t h e f i l l and drain a f t e r t h e J u l y 17 shutdown, t h e restric-

t i o n became continuous and more severe and was d o t amenable t o back-blowing. A remotely placeable assembly of electric h e a t e r s w a s designed and placed

around t h e junction of t h e i n l e t pipe and steel wool s e c t i o n of t h e a u x i l i ary charcoal bed.

I n t h e b r i e f shutdown between Runs 8 and 9 t h e l e v e l i n

t h e p i t was lowered below t h e bed inlets and t h e h e a t e r w a s energized. Some improvement w a s observed when t h e h e a t e r temperature reached 385OC.

After t h e temperature was raised t o 670째C and then cooled, t h e pressure drop was downby a f a c t o r of 5, t o a s a t i s f a c t o r y l e v e l .

The a u x i l i a r y bed

operated r e l a t i v e l y trouble-free u n t i l i t became p a r t i a l l y r e s t r i c t e d i n March 1969 and again about two months later.

Back-blowing cleared the bed

i n March; a forward-blw i n May w a s adequate but not as e f f e c t i v e .

No fur-

t h e r operational problems with t h e a u x i l i a r y bed were reported. Sections lA and 1 B of the main charcoal bed had been used almost exclusively during earlier runs, and they w e r e on-line when Run 8 s t a r t e d . The pressure drop a t t h e i n l e t of these s e c t i o n s began t o increase a few hours a f t e r t h e power was r a i s e d and when t h e pressure drop reached 7 p s i , s e c t i o n s 2A and 2B w e r e a l s o brought on l i n e .

Since back-blowing had be-

come progressively less e f f e c t i v e and i n l i g h t of t h e success with h e a t e r s on t h e i n l e t of t h e a u x i l i a r y bed, an attempt w a s made t o clear t h e res t r i c t i o n i n s e c t i o n 1 B by heating t h e i n l e t with a torch (electric h e a t e r s t o f i t t h i s bed were not then available).

Although t h e bed temperature

reached 465OC, t h e r e was no improvement and a l l four s e c t i o n s of t h e main bed were operated i n p a r a l l e l f o r Run 9.

After t h i s run, electric h e a t e r s

were i n s t a l l e d on t h e i n l e t s of sections lA and 1B.

Heating t o 400*C f o r

8 hours brought t h e pressure drop back t o t h e normal range f o r clean beds.

201

8.5

Subsequent Operating Experience with Fuel Off-gas System

With t h e major p a r t of t h e MSRE operating t i m e y e t t o come, t h e off-gas problems during t h i s period while s t i l l present were somewhat ameliorated by more e f f i c i e n t p a r t i c l e t r a p s and methods of c l e a r i n g t h e s e r e s t r i c t i o n s . Some new r e s t r i c t i o n s appeared but b a s i c a l l y they were similar t o those already described.

Remote maintenance methods were used t o cope with some

of these.

8.5.1

P a r t i c l e Trap The MK-I1 t r a p (Unit No. 1) w a s i n service from January 1967 t o May

of 1969 with no noticeable r e s t r i c t i o n ; however, during t h e f i r s t week i n

A week later

May, t h e pressure drop across t h e t r a p increased t o 0.4 p s i .

when t h i s r e s t r i c t i o n increased t o 0.75 p s i , t h e reserve u n i t , which had been valved o u t up t o t h i s t i m e , w a s put i n p a r a l l e l service. t h e pressure drop t o 0.1 p s i o r less.

This lowered

Also temperatures i n t h e t r a p re-

f l e c t i n g fission-gas heating indicated t h a t nearly a l l of t h e gas was flowi n g through t h e f r e s h bed.

tifad

I n October 1969, pressure drop measurements on

t h e off-gas system indicated a r e s t r i c t i o n of 0.3 p s i across t r a p No. 2; however, cycling t h e i n l e t valve, V-522C, between i t s closed and open pos i t i o n s cleared t h e r e s t r i c t i o n ; t h i s indicated a r e s t r i c t i o n i n t h e valve r a t h e r than t h e trap. ~

Three days later t h e r e s t r i c t i o n reappeared and w a s

cleared by t h e same method.

f u r t h e r problems i n t h i s area were en-

countered.

8.5.2

Introduction of Oxypen i n t o Fuel Drain Tank No. 1 I n t h e event of a s a l t

n t o t h e r e a c t o r c e l l , provisions had

been made t o vent t h e reacto

t o t h e a u x i l i a r y charcoal bed through

a l i n e equipped with two b l o

es.

I n January 1967, t h e s e valves,

V-571 A and B, were inadvertently opened while t h e r e a c t o r c e l l w a s being pressure-tested p r i o r t o Run 11. Reactor c e l l a i r entered t h e a u x i l i a r y charcoal bed which a t t h e t r e a c t o r via FD-1.

The slow

being used t o vent t h e empty b u t h o t age,of c e l l a i r i n t o t h e ACB r e s u l t e d i n

an increase i n f u e l system p r

u r e and was i n c o r r e c t l y diagnosed as plug-

ging i n the a u x i l i a r y charcoa

To clear t h e bed, i t was pressurized

t o 50 p s i g with helium and released i n t o FD-1 (V-571A was discovered t o be

about 1 / 2 t u r n open during t h e back-blowing operation).

202

It w a s estimated t h a t a maximum of 1.25 g-moles of 02 could have been

back-blown i n t o FD-1 which contained approximately 2 cu f t of molten f u e l It i s believed t h a t t h e oxygen w a s later purged out of t h e d r a i n

salt."

tank without appreciable mixing with t h e f u e l system piping. Subsequently, l i n e 571 which joined t h e two systems w a s c u t and both ends were cap-welded. 8.5.3

Charcoal Beds The r e s t r i c t i o n s a t t h e i n l e t sections of the,main charcoal bed con-

tinued t o exhibit chronic plugging throughout t h e remainder of MSRE power operations. , 1

Sections lA and 1 B began t o plug slowly a f t e r t h e two weeks of

power operation i n Run 11. After 4 weeks of operation t h e pressure drop increased from 2.5 t o 5 p s i and reached 7 p s i two months a f t e r t h e s t a r t of t h e run.

When sections 2A and 2B w e r e put i n s e r v i c e , t h e pressure drop in-

creased from 2.6 t o 9 p s i i n only t e n days.

Sections 2A and 2B were cleared

by forward-blowing with helium and sections lA and lB were cleared with t h e previously i n s t a l l e d heaters.

These operations d i d not r e q u i r e a reduction

i n r e a c t o r power but only a temporary lowering of the w a t e r . l e v e l i n t h e charcoal bed p i t t o allow the heaters t o function.

Approximately one month

d.'

later t h i s operation had t o be repeated. Owing t o t h e success of t h e h e a t e r s on t h e a u x i l i a r y bed and s e c t i o n s

IA and 1 B of t h e main bed, similar h e a t e r s w e r e i n s t a l l e d a t t h e i n l e t s of sections 2A and 2B during t h e shutdown between Runs 11 and 12. Approximately one month a f t e r the start of Run 12, i t w a s again neces-

sary t o clear a l l four sections. During MSRE s t a r t u p , i n l e c t i o n s of *'Kr

indicated t h a t t h e charcoal

bed hold-up time was 5-1/2 days f o r krypton with two s e c t i o n s i n p a r a l l e l service; t h e equivalent xenon hold-up time w a s calculated t o be 88 days. Thermocouples are located a t t h e entrance regions of t h e l-l/F,3-, and 6-inch diameter sections of t h e charcoa1,beds.

These s e c t i o n s contain 4.75,

27.4 and 67.85 percent, respectively, of t h e t o t a l charcoal i n t h e beds. On t h e b a s i s of a 5.5 day hold-up

t i m e f o r krypton, t h e l a g i n temperature

response t o f i s s i o n gas adsorption between t h e l-l/Fand 3-inch diameter sections should be 6.3 hours (4.755% of 5.5 days) when two beds are i n serv i c e and 3.1 hours when one bed is i n service.

During full-power operation

203

on September 15, 1967, t h e temperature response i n t e r v a l f o r beds lA and 1 B w a s 11-1/2 hours.

During a similar s t a r t u p on September 21, 1967, t h e

response i n t e r v a l w a s 9-1/2 hours f o r bed lA and 11-1/2 hours f o r bed 1 B i n d i c a t i n g t h a t bed 1 B had developed more of a r e s t r i c t i o n than bed lA. During t h e s t a r t u p of November 25, 1967 with beds 2A and 2B i n service, t h e response i n t e r v a l w a s 8-1/2 hours f o r bed 2A and no response i n bed 2B inBeds 2A and 2B were then

d i c a t i n g an appreciable r e s t r i c t i o n i n bed 2B.

valved off and bed lA was put i n service; t h e temperature response i n t e r v a l

was 6-1/2 hours.

Bed lA remained i n service f o r t h e remainder of Run 14

(130 days) with t h e exception of a 10-day period i n t h e middle of February when two s e c t i o n s w e r e put

s e r v i c e t o achieve t h e low fuel-pump pressure

required f o r a r e a c t i v i t y test.

The maximum pressure drop across t h e s i n g l e

s e c t i o n during Run 14 was 4

psi.

a c t i v i t y w a s detected by t h

a d i a t i o n monitors downstream of t h e charcoal

No d i s c e r n i b l e increase i n e f f l u e n t

t r a p during t h e s i n g l e bed test. No plugging i n t h e bed was experienced during t h e p r e c r i t i c a l and low power (<25 kW) operation using assU f u e l (Runs 15 and 16).

However, during

t h e next runs, gradual plugging recurred a f t e r t h e f i r s t 12 days of power operation (5 MW).

The electric h e a t e r s , previously i n s t a l l e d on t h e i n l e t

s e c t i o n , were used t o cle

t h e beds, although two of t h e t h r e e h e a t e r ele-

ments i n s t a l l e d on s e c t i o

lA and 1 B and one element on 2B f a i l e d .

This

procedure along with a back-blow a t t h e end of t h e 8-hour heating cycle w a s required t o clear t h e bed

ximately t h r e e months later.

during t h e valving operat

volved, t h e stem o r stem extension on t h e

Unfortunately,

i n l e t valve t o s e c t i o n 2A broke with t h e valve i n t h e closed position. p a i r s were not attempt and t h e s u f f i c i e n c y of During portions o gas t o study i t s e f f e c unusual off-gas problems t e r p r e t a t i o n of f

Re-

f t h e high r a d i a t i o n level a t t h e valve

ree sections.

(19) argon w a s s u b s t i t u t e d as a cover f r a c t i o n and r e a c t i v i t y changes.

with argon were manifest o t h e r than

re drops through t h e

tem during t h e

several switchovers

e other.

crease i n t h e e f f l u e n t

d i c a t o r ( a t t r i b u t e d t o neutron acti-

vation of argon i n t h e r e a c t o r ) was noticed.

A b a r e l y p e r c e p t i b l e in-

During t h i s run, i t w a s neces-

s a r y t o clear t h e charcoal beds twice, using t h e i n l e t h e a t e r and a

204 forward-blow with helium a t 25 psig.

No f u r t h e r problems were encountered

with t h e bed during the last run. 8.5.4

Restriction a t t h e Pump B o w l The r e s t r i c t i o n i n the off-gas l i n e a t t h e pump bowl reappeared i n

November 1967, one year s i n c e t h e l i n e w a s last cleared with a f l e x i b l e

reamer.

The r e s t r i c t i o n became noticeable approximately 2-1/2 days a f t e r

t h e reactor power was reduced t o 10 kW t o r e p a i r t h e sampler-enricher. During t h i s period t h e operational and maintenance valves w e r e open, and

a 0.6-liter/min

helium purge was maintained down t h e sampler tube t o t h e

pump bowl t o prevent contamination of t h e sampler by f i s s i o n gases.

The

r e s t r i c t i o n w a s evidenced by an increase (0.5 psig) i n t h e fuel-pump bowl pressure when t h e overflow tank vent valve w a s closed during a routine return of s a l t from t h e overflow tank t o t h e pump bowl.

After t h e sampler

w a s repaired, t h e r e s t r i c t i o n w a s relieved by pressurizing t h e pump bowl t o 6.0 psig and suddenly venting t h e gas i n t o a drain tank which w a s a t 3 psig.

The pressure drop then appeared notmal ( ~ 0 . 1p s i ) , b u t a f t e r t h r e e more days of low-power operation, an abnormal pressure rise (0.2 p s i ) w a s again observed when s a l t was being returned from t h e overflow tank.

Repetition

of t h e mild blow-through t o t h e drain tank had l i t t l e o r no e f f e c t t h i s The l i n e w a s not completely blocked, however, and full-power opera-

time.

t i o n w a s resumed. A t f i r s t , temperatures on t h e overflow tank and t h e offgas l i n e (responding t o f i s s i o n product heating) c l e a r l y indicated t h a t

there was enough pressure drop a t t h e pump-bowl o u t l e t t o cause much of t h e Then,

off-gas t o bubble through t h e overflow tank and out i t s vent l i n e .

a f t e r 1 5 h r a t f u l l power, t h e bypass flow stopped, indicating t h a t t h e res t r i c t i o n had decreased s i g n i f i c a n t l y ; i t remained below t h e l i m i t s of det e c t i o n throughout t h e next nine weeks while t h e reactor w a s operating a t 7.2 MW o r 5 MW.

After two days a t 10 kW, t h e pressure drop again became

detectable and continued t o increase over t h e next s i x days a t very low power.

When power operation was resumed a t 5 MW, temperatures indicated

t h a t there was again bypass flow through t h e overflow tank.

The restric-

t i o n increased during two weeks of operation a t 5 MW, but t h e l i n e never became completely plugged.

Four days a f t e r t h e resumption of full-power

operation i n February 1968 while t h e overflow tank was being emptied, t h e

205

r e s t r i c t i o n p a r t i a l l y blew out, bringing t h e pressure drop again below t h e

l i m i t of detection.

The pressure drop remained j u s t a t o r below t h e l i m i t

u n t i l the March-1968 shutdown f o r f u e l processing. The behavior w a s q u i t e unlike the plugging t h a t occurred earlier i n t h e same s e c t i o n of l i n e as a r e s u l t of t h e o v e r f i l l with f l u s h s a l t , mainl y i n t h a t it never completely pluggeqand t h e r e w a s a suggestion of some e f f e c t of power level.

A f t e r t h e f u e l was drained, the f l e x i b l e s e c t i o n of off-gas l i n e w a s removed, and a f e l x i b l e t o o l was run through t h e 28-in. pump bowl.

of l i n e back t o t h e

The t o o l s , a 1/4-in. cable with a diamond d r i l l t i p , encountered

some r e s i s t a n c e a t f i r s t and came out with a considerable amount of s o l i d s on it.

Although t h e f l e x i b l e jumper and t h e l i n e downstream were not re-

s t r i c t e d , t h e jumper l i n e along with the f l e x i b l e reaming t o o l s were s e n t t o t h e High-Radiation-Level-Examination

Laboratory f o r cut-up and exami-

nation. The replacement jumper l i n e was equipped with four thermocouples t o provide more information during power operation.

Lp,

A slender basket contain-

ing devices of metal and graphite t o c o l l e c t material from t h e off-gas

stream was suspended i n t h e entrance of t h e 4-in. off-gas hold-up pipe. The r e s t r i c t i o n a t t h e pump bowl reappeared a f t e r approximately one month of operation with t h e n

f u e l i n September 1968.

The r e s t r i c t i o n

was blown,clear once during p

sure-drop measurements, but reappeared a

week later and could not be dislodged by back-blowing helium from t h e d r a i n drop across t h e r e s t r i c t i o n had intank t o t h e pump b creased t o 4 p s i b a n t systems were drained t o

also t o pix t h e f u e l befor The flange a t b l e cable assembly

ream out t h e off-gas li t h e r e s t r i c t i n g materia recovered on t h e f

was reconnected a

i n November 1968, when t h e f u e l and coolt t h e off-gas l i n e s of both systems and

of full-power operation.

was again disconnected, and a f l e x i l t e r and a vacuum pump, w a s used t o 1 pump and a l s o t o c o l l e c t a sample of ry small amount of blackish d u s t was t h e f l e x i b l e tubing.

The flange

f i l l e d ; however, after about t h r e e

weeks of operation, t h e r e s t r i c t i o n reappeared i n mid-January and gradually

increased t o a AP of 5 . 5 p s i .

Because t h e rapid recurrence of t h e r e s t r i c t i o n

206

a t the pump-bowl e x i t threatened t o become a serious nuisance, a compact

2-kW heater u n i t was designed f o r i n s t a l l a t i o n on t h i s l i n e a t t h e f i r s t opportunity; however, i n t h e last week i n February during a routine trans-

fer of f u e l from t h e overflow tank t o t h e f u e l pump, t h e r e s t r i c t i o n was blown clear, r e s u l t i n g i n a complete t r a n s f e r of f u e l from t h e overflow tank t o t h e pump bowl.

The pressure drop i n t h i s s e c t i o n of l i n e w a s then

below t h e l i m i t of detection (about 0.1 p s i ) . On f i v e other occasions i n t h e next 9 weeks t h e pressure drop became

noticeable, but each time i t blew out when t h e overflow tank w a s burped. On May 1 it reappeared and remained detectable through t h e June-1969 shut-

down.

The r e s t r i c t i o n caused most of t h e pump-bowl purge gas flow t o bub-

b l e through t h e overflow tank except when t h e overflow tank w a s being pressurized t o push accumulated s a l t back t o t h e pump bowl.

As a consequence,

on May 25, t h e off-gas l i n e from t h e overflow tank became almost completely plugged and i t became necessary t o turn off t h e overflow tank bubbler flow most of t h e time.

Thereafter a l l t h e pump purge gas w a s forced out through

t h e off-gas l i n e from t h e pump bowl.

This s i t u a t i o n caused some inconveni-

ence during t h e burping operation, but by reducing purge flow t o t h e pump bowl a t these times, s a l t could be transferred back t o t h e f u e l pump without exceeding 1 5 psig i n t h e pump bowl. During t h e June shutdown, t h e 2-kW heater t h a t had been b u i l t earlier

was i n s t a l l e d remotely on t h e pipe as near t o t h e pump as possible and connected t o s p a r e power and thermocouple leads i n t h e r e a c t o r cell.

The pump-

tank furnace heaters and t h e new heater were turned up t o bring t h e s e c t i o n of off-gas l i n e t o near 65OoC, then gas pressure w a s applied t o blow t h e r e s t r i c t i n g material back toward t h e pump bowl.

The heater was subsequent-

l y turned off but l e f t connected so t h a t i t could be used again without reopening t h e reactor cell. The set of specimens exposed i n t h e off-gas hold-up volume s i n c e J u l y 1968 w a s a l s o removed during t h e June shutdown. Although the r e s t r i c t i o n did not reappear during Run 19, t h e pressure modifier and recorder which was i n s t a l l e d on t h e fuel-pump pressure trans-

m i t t e r t o d e t e c t high frequency pressure noise i n t h e r e a c t o r core indicated t h a t a r e s t r i c t i o n a t t h e pump exit was again i n i t s formative stage.

Lid

207

The r e s t r i c t i o n did not become evident (by previous standards) u n t i l approximately two weeks after t h e f u e l f i l l of Run 20.

I n t h e t h r e e days

before f i n a l r e a c t o r shutdown, t h e r e s t r i c t i o n increased from 1.3 t o 2.2 p s i as measured with t h e procedure f o r s a l t recovery from t h e overflow tank. Because of t h e proximity of t h e r e a c t o r shutdown d a t e and t h e d e s i r a b i l i t y of not i n t e r r u p t i n g t h e experiment i n progress, t h e h e a t e r which had previously been i n s t a l l e d and successfully used t o clear a similar r e s t r i c t i o n ,

was not used. Examination of t h e 522 Jumper Line Removed March 1968.

An 8-mg Sam-

p l e of what appeared t o be soot was c o l l e c t e d from t h e upstream flange of t h e f l e x i b l e jumper l i n e ; a

-mg sample of a similar material w a s collected

from t h e downstream flange.

he 8-mg sample indicated about 80 R/hr and t h e

15-mg sample indicated about 200 R/hr a t 'tcontact."

Each of t h e flanges and

t h e convolutions of t h e f l e x i b l e hose appeared t o have remaining on them a t h i n dull-black film.

A one-inch s e c t i o n c u t from t h e upstream end of t h e

jumper l i n e indicated 150 R/hr; a similar s e c t i o n from t h e downstream end indicated 350 R/hr.

The reaming t o o l s used w e r e covered with blackish,

c l e a r i n g t h e r e s t r i c t i o n a t t h e pump bowl t y , granular material which was i d e n t i f i e d

by x-ray as f u e l s a l t p a r t i c l e s .

The r e l a t i v e l y high values of l i t h i u m

and beryllium found i n t h i s sample may i n d i c a t e pickup of some r e s i d u a l LiFa-BeFa f l u s h s a l t which had entered t h e off-gas l i n e as a r e s u l t of t h e pump o v e r f i l l i n 1966. Electron photomicrogr

of t h e upstream d u s t showed r e l a t i v e l y s o l i d

p a r t i c l e s of 1~surrounded

material of l i g h t e r and d i f f e r e n t s t r u c t u r e

which appeared t o be.amor p l e s showed 12 t o 16%

carbon.

Chemical analyses of t h e dust sam-

rbon, 28 t o 54% f u e l s a l t , and 4% s t r u c t u r a l metals.

Based on a c t i v i t y data, f i s

products could have amounted t o 2 t o 3%of

t h e sample weight.

t a b i l i t y of t h e remainder is a t t r i b u t a b l e

The una

t o t h e small amount of samp From t h e dust c o l l e c t e d from l i n e 522 onto a f i l t e r during t h e December1968 shutdown, t h e followin

a 3 3 ~ 1 4 0Ba,

'44Ce,

(2) t h e isotopes

and ''2 Sr,

onclusions"

were reached:

(1) t h e isotopes

ere transported only as s a l t constituents;

Y, and "'CS,

which have noble gas precursors, were

208

present i n s i g n i f i c a n t l y g r e a t e r proportions consistent with a mode of transport other than s a l t p a r t i c l e s ; (3) t h e isotopes '"Nb, "Mo, .10 6 Ru, and "'ye

w e r e present i n even g r e a t e r proportions, i n d i c a t i n g t h a t they

were transported more vigorously than f u e l s a l t .

A more complete discussion of these findings and t h e r e s u l t s of examination of t h e sample specimens removed from t h e 4-in. hold-up volume w i l l be covered i n a report t o be issued by Compere and Bohlmann e n t i t l e d "Fiss i o n Product Behavior i n t h e MSRE During

'Uranium Operation" (ORNL-TM-

2753). 8.5.5

Other Restrictions i n Line 522

On f i v e occasions during Runs 1 7 and 18, complete flow blockage occurred i n t h e main off-gas l i n e somewhere between t h e 4-in.-diam hold-up pipe and t h e junction of Line 522 and t h e auxiliary charcoal b e d ' l i n e , 533. Although t h e plug could have occurred a t t h e i n l e t p o r t of V-522A, i t is unlikely t h a t t h e plug formed i n t h e valve i t s e l f because manipulating' t h e valve stem t o i t s f u l l travel i n both d i r e c t i o n s d i d not break t h e plug loose.

I n each case t h e f u e l system pressure w a s vented f o r s h o r t periods

of time through t h e equalizer and d r a i n tank vent l i n e s t o t h e a u x i l i a r y charcoal bed u n t i l l i n e 522 was unplugged by back-flowing helium from l i n e

571, through l i n e s 561, 533, and 522 t o t h e hold-up volume. each plug became successively more d i f f i c u l t t o dislodge.

It seems t h a t To dislodge 'the

l a s t plug, it was necessary t o vent t h e f u e l system t o 2 p s i g , pressurize l i n e 533 t o 60 psig, and quickly open V-52%.

The l a s t e f f o r t apparently

restored t h e l i n e t o near its o r i g i n a l condition because plugging a t t h i s l o c a t i o n had not recurred a t the end of r e a c t o r operations.

Because of t h e

remoteness and i n a c c e s s i b l i t y of t h i s l i n e , no f u r t h e r e f f o r t was made t o l o c a t e t h e exact p o s i t i o n o r characterize t h e nature of t h e plug. 8.5.6

R e s t r i c t i o n i n t h e Overflow Tank Exit Line The f u e l pump was provided with an overflow pipe (shown i n Fig. 8.9)

and a catch tank (5.4 ft')

t o prevent s a l t from becoming,high enough t o al-

low it t o . e n t e r t h e gas and l u b r i c a t i n g passages of t h e pump as a r e s u l t of unusually high gas entrainment i n t h e c i r c u l a t i n g s a l t o r possibly as a r e s u l t of a high-temperature excursion while operating.

The overflow pipe

from t h e pump dips t o within less than 1/2-in. of t h e bottom and i n t o a

209

1/2-in.-deep

dimple pressed i n t h e lower tank head t o permit more complete

r e t r i e v a l of i t s contents.

During operations, s a l t spray and/or bubbles

slowly s p i l l e d i n t o t h e overflow tank (OFT);

t h e s a l t w a s p e r i o d i c a l l y re-

turned t o t h e pump bowl by closing t h e OFT gas discharge valve and allowing helium from t h e bubble-type level i n d i c a t o r s t o build up t h e required press u r e (%3.5 p s i > pump bowl pressure) t o f o r c e t h e s a l t back t o t h e pump bowl. When t h e OFT discharge is opened, t h e s a l t remaining i n t h e tank and t h a t remaining i n t h e pipe is s u f f i c i e n t t o provide a seal a t t h e bottom of t h e tank and thus prevent f i s s i o n product gases from entering t h e overflow tank. However, when t h e main off-gas l i n e , 522, became r e s t r i c t e d near t h e pump bowl, p r a c t i c a l l y a l l of t h e pump purge gas and gaseous f i s s i o n products

were bubbled through t h e s a l t h e e l i n t h e OFT and o u t through l i n e 523 t o j o i n t h e main off-gas l i n e downstream of TE-522-2.

From TE-522-2

d a t a and

pressure measurements during s a l t recovery operations, i t w a s estimated t h a t f i s s i o n product gases were f i r s t diverted through t h e OFT on October 15, 1966.

Since t h e fuel-pump ove

ill on J u l y 24, 1966, i t is estimated t h a t

most of t h e fuel-pump gases were forced through t h e overflow tank f o r ap-

proximately 28 days i n 1966, 6 days i n 1967, 76 days i n 1968, and 68 days i n 1969, before t h e overflow tank gas e x i t l i n e (523) became plugged on May 25.

Thereafter, a l l of t h e pump purge gas w a s forced o u t through t h e

r e s t r i c t i o n i n t h e off-gas l i n e a t t h e e x i t of t h e pump bowl.

This s i t u -

a t i o n caused some inconvenience during t h e s a l t recovery operation, but by reducing purge flow t o t h e pump a t these t i m e s , t h e operation could be done without exceeding 15 psig

During t h e shutdown i scanned with t h e remote g observed a t t h e air-operate the overflow tank.

he pump bowl.

e 1969, when t h e overflow tank vent l i n e was rometer, an unusually s t r o n g source w a s (HCV-523),

about 33 f t downstream from

Flanges were opened and pressure observations showed

that the restriction w

langed s e c t i o n containing both t h e air-

operated valve and a h

This s e c t i o n was removed t o a h o t c e l l

where polymerized hydrocarb i n g t h e hand valve i n l e t

operated valve but no hand v

nd f i s s i o n products were found t o be blockreplacement s e c t i o n containing an aire was i n s t a l l e d .

Although t h e r e s t r i c t i o n

a t t h e exit of t h e pump bowl reappeared during t h e l a s t week of MSRE operat i o n , t h e OFT discharge l i n e functioned normally.

210 8.5.7

R e s t r i c t i o n s a t Exits of Drain Tanks During t h e r e a c t o r d r a i n i n June of 1969, pressure measurements indi-

cated t h a t a r e s t r i c t i o n had developed i n t h e gas l i n e somewhere between Fuel Drain Tank No. 2 (-2) t h e e x i t l i n e (575).

and t h e junction of t h e i n l e t l i n e (574) and

The f i r s t half of t h e d r a i n appeared normal a t a d r a i n

rate of %2 cu f t of s a l t per minute a f t e r which i t tapered o f f t o 1~0.25cu f t per minute when t h e trapped gas i n t h e d r a i n tank balanced t h e s a l t i n

t h e f u e l system.

Two and a h a l f hours were required f o r t h e d r a i n as op-

posed t o a normal d r a i n time of less than 40 minutes when drained i n t o one tank.

The r e s t r i c t i o n w a s l a t e r cleared by heating t h e tank t o 68OOC and

applying a pressure d i f f e r e n t i a l of 60 p s i g across t h e plug via l i n e 561.

A similar but lesser r e s t r i c t i o n i n t h e gas l i n e a t t h e o t h e r f u e l d r a i n tank w a s p a r t i a l l y cleared during t h e f u e l system pressure test i n August 1969 by heating t h e tank t o 68OOC and flowing helium from FD-2 ( a t 50 psig) t o FD-1 (at 2 psig). R e s t r i c t i o n s i n these l i n e s were most probably deposited (but not t o t h e point of detection) during December of 1966 and January of 1967 when t h e r e a c t o r off-gas w a s routed t o -1,

through FV-106 and FV-105,

t h e s a l t h e e l i n FD-2, and out through l i n e 575 (Sect. 5.6.1).

through

Fission gas

decay heating, during t h i s t i m e , was appreciable i n t h e d r a i n tanks and required downward adjustments t o t h e d r a i n tank heaters.

The off-gas flow

through t h e d r a i n tanks was reversed after t h e first week i n an e f f o r t t o minimize plugging.

;

The off-gas flow w a s again routed through d r a i n tank No. 2 gas lines f o r a day when l i n e 522 plugged i n May of 1969. I n October of 1969 during a pressure release test when t h e fuel-pump gas was released i n t o FD-1,

it w a s discovered t h a t t h e gas l i n e was again

r e s t r i c t e d near t h e d r a i n tank.

Two attempts w e r e made t o clear t h i s l i n e

by back-blowing helium, f i r s t a t 30 psig and then a t 50 p s i g , from l i n e 561 through HCV-573 and i n t o FD-1.

Although only marginal improvement w a s ob-

tained, t h e gas flow through t h i s l i n e was judged adequate f o r any emergency d r a i n s i t u a t i o n . During t h e subsequent d r a i n i n November 1969, approximately 4300 l b of

s a l t drained i n t o FD-1 and 5100 l b s drained i n t o FD-2 i n d i c a t i n g t h a t a s l i g h t r e s t r i c t i o n existed i n t h e exit gas l i n e of FD-1.

During t h e

December drain, the exit gas valve from FD-2 was closed during part of the drain to compensate for the partial restriction in FD-1 gas line; however, overcompensation resulted and 5400 lbs drained into FD-1 and 4100 lbs drained into FD-2. The drain.times for both drains were less than 20 minutes which is normal for a drain when both tanks are used. 8.5.8

Restrictions at the Off-Ras Sampler A system to permit the analysis of the reactor off-gas stream was in-

stalled downstream of the particle trap as shown in Fig. 8.2. The sampler contained two thermal-conductivity cells, a copper oxide converter, and two molecular sieves, one operating at .liquid nitrogen temperature and the other at room temperature.8 Since the sampler was an integral part of primary containment during sampling operations, and since some components of the sampler did not meet the requirements of primary containment, solenoid block valves were installed in the inlet and outlet lines which connect the sampler to the reactor system. Two fail-closed valves in series were installed in each line and were instrumented to close on high sampler activity, high reactor cell pressure, and high fuel-pump pressure. Although located downstream of the particle trap, the inlet line to the off-gas sampler periodically developed a restriction in the vicinity of the safety block valves. The block valves have 3/32-in.-diam ports and the inlet piping is 0.083 in ID autoclave.tubing. The restriction was successfully cleared each time by back-blowing with helium. As encountered in other parts of the off-gas system, successively higher pressures were required to clear the restriction each time. During the latter part of uld not shut off tightly and were replaced Run 18, the,inlet block va ction of the faulty inlet valves did not rein July of 1969. Visual i g valves nor the nature of the restriction veal the.reason for the.le (the restriction had been blown clear before the valves were replaced). During the August startup, it was again necessary to blow out the restriction at the inlet to the sampler.

212 8.6

Discussion and Conclusions

The periodic plugging i n t h e coolant system off-gas f i l t e r and a l s o

i n t h e f u e l system f i l t e r during t h e p r e c r i t i c a l and low-power (<25 kW) operation can b e a t t r i b u t e d primarily t o t h e accumulation of o i l i n t h e 0.7 t o

diameter pores of t h e s i n t e r e d metal f i l t e r s (2 t o 411 pores might

have been a b e t t e r choice).

However, subsequent f i l t e r s i n t h e f u e l system

were constructed of f e l t metal (Huyck Nos. FM-225 and FM-204) which w a s not only e f f i c i e n t a t stopping s o l i d f i s s i o n products but a l s o w a s apparently immune t o plugging by hydrocarbons.

V i r t u a l l y a l l of t h e s o l i d decay daugh-

ters were stopped by t h e Yorkmesh and f i l t e r before reaching t h e Fiberfrax s e c t i o n of t h e p a r t i c l e trap. O i l vapor i n t h e off-gas stream condensed on various p a r t s of t h e off-

gas system components and apparently enhanced t h e adsorption o r trapping of p a r t i c u l a t e matter, p a r t i c u l a r l y on the Yorkmesh portion of the p a r t i c l e trap The f a c t t h a t such l a r g e amounts of s o l i d decay daughters were trapped i n t h e p a r t i c l e t r a p i n d i c a t e s t h a t most of these s o l i d s tend t o remain i n t h e gas phase u n t i l trapped onto a surface wetted by hydrocarbons.

It i s

not clear whether t h e s o l i d decay daughters agglomerate i n t h e hold-up vol-

ume and i f so whether they agglomerate with o t h e r s of t h e i r own species, with other species, with hydrocarbons, or possibly combine chemically t o form both v o l a t i l e and non-volatile compounds.

A l a r g e f r a c t i o n of t h e

noble gas f i s s i o n products decay i n t h e 4-inch-diameter hold-up volume and t h e i r decay daughters apparently p l a t e out o r are adsorbed onto t h e f i r s t flow channel r e s t r i c t i o n where t h e gas flow changes d i r e c t i o n and/or is c l o s e t o a r e l a t i v e l y cold surface such as valves and flow r e s t r i c t o r s .

The high hydrocarbon content (>60X) of t h e material c o l l e c t e d on t h e Yorkmesh i s inconsistent with t h e high temperature (%65OoC) i n t h i s region during power operation; hydrocarbons are vaporized and cracked a t this tem'perature.

One must assume, then, t h a t t h e hydrocarbons were collected

sometime a f t e r power operation had ceased and a f t e r t h e f i s s i o n product decay heat became negligible.

The growth r i n g s on t h e Yorkmesh were probably

caused by a l t e r n a t e periods of power and zero-power operation.

213

Since spectrographic an l y s i s d i d not show any of t h e a l k li m e t a l

cesium o r rubidium, on t h e Yorkmesh, they were apparently boiled o f f a l s o

by f i s s i o n product decay heating i n t h i s region.

I f t h i s be t h e case, ap-

preciable q u a n t i t i e s of these materials would be expected t o be c o l l e c t e d i n t h e charcoal beds.

On t h e o t h e r hand, l a r g e q u a n t i t i e s of barium and

strontium were reported on t h e Yorkmesh; t h i s would argue f o r e f f e c t i v e trapping of t h e i r precursors, cesium and ribidium.

However, t o be trapped

on Yorkmesh a t 65OoC, t h e a l k a l i metals would n e c e s s a r i l y have t o be chemic a l l y combined (as halides f o r instance) r a t h e r than i n t h e elemental form, otherwise they would b o i l off and be c a r r i e d downstream. Although none of t h e s o l i d decay daughters i n t h e 4-in.-diam

pipe got

p a s t t h e p a r t i c l e t r a p , another batch (though somewhat smaller number) of s o l i d decay daughters are born (and presumably remain i n t h e gas phase) i n t h e l a r g e r hold-up volume ahead of t h e main charcoal beds.

These f i s s i o n

products along with hydrocarbons are probably t h e source of t h e plugging experienced a t t h e entrance region of t h e beds. After examination of t h e MK-I p a r t i c l e t r a p , i t is understandable why t h e entrance region of t h e charcoal beds plugged periodically; t h e entrance pipe ( 1 / 4 4 n . sched-40) e n t e r s t h e bed normal t o t h e s e c t i o n where t h e s t a i n l e s s steel wool is packed, thus t h e cross-sectional flow area (trapping area) a t t h e entrance t o t h e s t a i n l e s s s t e e l wool i s only 0.1 sq. i n . Although a p a r t i c l e t r a p w a s not i n s t a l l e d between t h e second hold-up volumdand t h e charcoal bed, i t w a s considered and w a s d e f i n i t e l y needed.

Also the.entrance s e c t i o n of f u t u r e beds should be redesigned such t h a t gas e n t e r s t h e empty chamber above t h e s t e e l wool and t h u s o f f e r a l a r g e r cross s e c t i o n of s t a i n l e s s s t e e l wool t o gas flow as i n t h e MK-11 t r a p . Another argument t h a t f i s s i o n products, as w e l l as hydrocarbons, are involved i n t h e plugging mechanism a t . t h e charcoal bed entrance, is t h e f a c t t h a t t h e beds remained f r e e of plugging u n t i l power operations were begun. For a b e t t e r understanding of t h e plugging mechanism, t h e MK-I1 p a r t i c l e t r a p and t h e entrance s e c t i o n t o one of t h e charcoal beds would need t o be examined.

Also a t least one bed could be examined along i t s length t o de-

termine t h e f i s s i o n product adsorption c h a r a c t e r i s t i c s of t h e bed. The probable reason t h a t t h e overflow tank exit l i n e remained clear

f o r such a long time was the f a c t t h a t pressure d i f f e r e n t i a l s (Q3.5 p s i )

were periodically released through t h e l i n e each t i m e s a l t was returned t o t h e pump bowl.

The pressure pulses probably cleared any i n c i p i e n t plug

which may have been i n i t s formative stage. The r e s t r i c t i o n a t t h e f u e l pump bowl does not seem t o be r e l a t e d t o f i s s i o n products s i n c e t h e plug seems t o form as r e a d i l y i f not more so while s u b c r i t i c a l and a t low power as i t does a t f u l l power.

The gas e x i t

l i n e from t h e pump bowl should be modified t o eliminate m i s t o r l i q u i d carryover i n t o t h e off-gas stream.

A heated cyclone-type separator"

or a

l a r g e w e l l baffled and cooled region which could b e l a t e r heated t o m e l t down any s a l t formation are two p o s s i b i l i t i e s . I f hydrocarbons had not been present i n t h e pump bowl, f i s s i o n gas behavior i n t h e MSRE might have been somewhat d i f f e r e n t from t h a t experienced. Some speculations on probable r e s u l t s are (1) less s e r i o u s o v e r a l l plugging because t h e r e would be no semisolid varnish-like buildup due t o hydrocarbons, (2) t h e York mesh would probably be less e f f i c i e n t a t trapping t h e s o l i d decay daughters and more of t h i s material would then be trapped on t h e f e l t metal f i l t e r , (3) t h e f i s s i o n product "cake" would probably be loosely packed on t h e f i l t e r and more e a s i l y disrupted by backflow s i n c e i t would not contain a 'tbinder" o r p a s t e material (hydrocarbons).

215

FUEL AND COOLANT PUMP LUBE OIL SYSTEMS

J. K. Franzreb

9.1

Description

Two i d e n t i c a l o i l systems served t o ~ l u b r i c a t eand cool t h e f u e l and

coolant pump bearings and t o cool t h e s h i e l d plugs located between t h e bearings and t h e bowls of t h e pumps. meet containment requirements.

Each was a closed loop designed t o

The o i l used was Gulf-Spin 35.

Each system consisted of two 5-hp, 60-gpm a t 160-ft head, 3500-rpm Allis Chalmers Electri-Cand pumps, (one normally i n operation with t h e other

i n standby), an in-line Cuno EFS o i l f i l t e r , and an o i l tank of 22-gal ope r a t i n g capacity having "brazed on'' water c o i l s capable of removing. 41,000 Btu/hr

The o i l flow t o t h e bearings of each of t h e s a l t pumps was about 4 gpm with about 8 gpm t o each s h i e l d plug.

The remaining 4 8 q g m was recycled

through t h e o i l tank t o a i d i n cooling.

A scavenging J e t was i n s t a l l e d i n

the o i l l i n e s near t h e s a l t pump s o t h a t t h e s h i e l d plug o i l flow aided t h e return of t h e bearing o i l t o t h e storage tank. Lubricating o i l seeping past t h e lower s h a f t s e a l of t h e s a l t pump was piped t o t h e o i l catch tank f o r measuring t h e r a t e o f leakage. The pumps, storage tank, f i l t e r and much of t h e instrumentation and valving f o r each system were mounted i n an angle i r o n frame t o f a c i l i t a t e moving t o t h e s i t e .

This was commonly r e f e r r e d t o as a lube-oil package.

The two packages were interconnected s o t h a t e i t h e r could be used i n an emergency t o supply both salt pumps.

9.2 Before i n s t a l l a t i o n i n a t e s t stand.

e MSRE, both lube-oil packages were operated

Heat load and pressure drop d a t a were obtained and t h e

performance of t h e systems was checked.

A problem o f gas entrainment w a s

found which affected t h e priming of t h e standby pumps.

Approximately

30 seconds were required t o prime a spare pump i f it had not been operated f o r several hours.

Priming time w a s reduced t o 5 sec by i n s t a l l i n g gas

216 -

vents from t h e pump volute casing and t h e pump discharge l i n e t o t h e gas space i n t h e o i l tank.

This experience l e d t o a modus operendi of' s t a r t i n g

t h e spare pump of each package once per s h i f t and running it f o r about

1 5 min. The packages were i n s t a l l e d at t h e MSRE i n 1964.

I n i t i a l operation

was hampered by f a i l u r e of t h e s t a t o r i n s u l a t i o n i n one of t h e pump motors and by low resistance t o ground i n s t a t o r s of others.

Since moisture w a s

suspected, a p o t t i n g compound (Dow Corning S i l a s t i c RTV-731) was used t o s e a l t h e motor housing j o i n t s and a moisture r e s i s t a n t coating of paint (Sherwin W i l l i a m s epoxy .white B69W6) w a s applied t o t h e e x t e r i o r surfaces of a l l four operating motors, plus two spares. The loss of prime of t h e standby pumps continued t o be a problem u n t i l

t h e scavenging J e t s used t o r e t u r n o i l t o t h e reservoirs from t h e salt pumps were replaced with j e t s of lower capacity t o reduce t h e entrainment of gas i n t h e r e t u r n o i l streams.

These replacements were done i n September 1965.

A f t e r modifications, it was possible t o reduce t h e frequency of priming of t h e standby pumps t o once a week. Because of t h e lowered capacity of t h e new r e t u r n j e t s , it was found necessary t o l i m i t t h e f l o w of o i l t o t h e fuel-pump motor t o

4 gpm, as a

flow of 5 gpm would result i n a buildup of o i l i n t h e s a l t pump motor cavity. ,

9.3

Addition of Syphon Tanks t o t h e O i l Catch Tanks

The o i l catch tanks were f a b r i c a t e d of a 46-1/2-ine-1ong 2-in. pipe topped by a 20-in.-long

section of 8-in. pipe.

section of

The lower sec-

t i o n allowed accurate measurement of t h e o i l accumulation rate whereas t h e upper portion provided s u f f i c i e n t volume t o handle possible gross leakage. The catch tanks were drained periodically t o keep t h e l e v e l i n t h e lower section,

Radiation would not permit doing t h i s during power operation and

since t h e volume of t h e lower section was around 2500 cc and t h e allowable

seal leak rate was 100 cc/day, it w a s possible t h a t t h e r e a c t o r would have t o be shut down t o drain t h e o i l catch tanks during extended runs. Therefore, i n October 1965, equipment was i n s t a l l e d t o automatically siphon t h e

o i l from t h e catch tanks when they became f u l l .

217

After i n s t a l l a t i o n , t e s t s were run on these and they performed w e l l . However, they f a i l e d t o function properly a t t h e low o i l leakage r a t e s that a c t u a l l y occurred during operation.

The o i l flowed over t h e high point of

t h e syphon tubes, as over a weir, without bridging t h e tube t o form a syphon.

We therefore reverted t o manual draining during shutdowns.

Fortu-

nately t h e leak r a t e s did not get high enough t o i n t e r f e r e w i t h operation of t h e reactor. 9.4

O i l Leakage

Continuous records of o i l tank a n d ~ o i lcatch tank l e v e l s were maint a i n e d during all of t h e MSRE operating l i f e t o determine what l o s s e s were taking place.

Samples removed from t h e systems were c a r e f u l l y measured,

as were additions.

The o i l storage tanks were l a r g e and therefore s m a l l

changes i n l e v e l indication caused.large e r r o r s i n inventory. log readings could vary by 500 t o 600

CC.

Successive

F a i r l y accurate indication of

t h e leakage through t h e r o t a t i n g o i l seals of t h e s a l t pumps t o t h e o i l

s e a l s of t h e s a l t pumps t o t h e o i l catch tanks were possible over one month o r longer periods.

The leak rates during operation varied from a few cc

.per day f o r e i t h e r s a l t pump t o around 25 cc/day.

The s e a l s seemed t o get

worse o r improve f o r no apparent reason. For t h e 24-month period through August 1969 , inventories showed unaccounted f o r l o s s e s from t h e f u e l pump o i l ' s y s t e m of 5.4 the coolant o i l system of 5.6

5e'+

-3.0

l i t e r s ; from

l i t e r s or average d a i l y l o s s e s of 7 . 5 cc -3.0 (5.25 gms), and 7.8 cc (5.4 gl~ls) respectively. Analyses indicated that

t h e r e were approximately 1 t o pump off-gas stream.

5 0 ' +

gms of o i l products per day i n t h e f u e l

Some o i l may have been held up i n t h e l i n e s , 6-ft3

holdup volime, and p a r t i c l e t r a p s t h a t were upstream of t h h sampling point. I n conclusion, t h e b e s t estimates showed an o i l loss of 5+ gms/day, of which only 1

-2

@~ms co

gas f r o m t h e primary s y s t e for.

e found as'hydrocarbons i n samples of off-

aving approximately 3 gms/day unaccounted

218

9.5

Change-Out of O i l Pumps

A f t e r t h e i n i t i a l trouble with t h e e l e c t r i c a l i n s u l a t i o n due t o mois-

ture, t h e pumps gave very s a t i s f a c t o r y service.

Two pumps were removed

from service, one because of excessive vibration and t h e other because of an e l e c t r i c e short i n t h e motor winding. month period ending August 31,

This w a s done during t h e six-

1967.

The p\m?p with excessive vibration had one of t h e two balancing d i s k s

loose on t h e shaft.

This loose disk was reattached, and t h e s h a f t assembly

was dynamically balanced.

The pump was reassembled, t e s t e d under operating

conditions and made ready f o r service.

The pump with t h e shorted winding

was rewound and put back i n t o service. Test Check of One O i l System 9.6 Supplying Both t h e Fuel and Coolant S a l t Pumps On .March 20, 1966, t h e two o i l systems were valved so t h a t t h e f u e l o i l package (FOP-2 running) supplied o i l t o both salt pumps.

This was done

t o check calculations, and t o be sure t h a t t h i s could indeed be done i n some future emergency.

The coolant pump was c i r c u l a t i n g salt a t 1200'F;

t h e f u e l pump was idle and at a temperature of 125'F.

Adequate flaws t o

both s a l t p p p s were maintained (3.35 gpm t o t h e bearings and 6.5 and 7 . 3 'gpm t o t h e coolant and f u e l pump s h i e l d plugs). o i l supply temperature e q u i l i b r a t e d a t l27'F.

Under t h e s e conditions t h e When t h e f u e l pump o i l flow

was stopped, t h e o i l supply temperature increased

7'F,

indicating t h a t

t h e f u e l p u p was removing p a r t o f t h e heat load of t h e o i l system. Although t h i s test did not prove conclusively t h a t one o i l cooling system was adequate f o r an emergency wherein one package would be used t o supply both-operating salt pumps, subsequent heat exchanger tests and calculations indicated t h a t one o i l cooling system would be adequate.

9.7

O i l Temperature Problems

The t o t a l o i l f l o w from one hibe o i l pump w a s normally 60 gpm.

t h i s , 8 gpm-w a s directed t o t h e s h i e l d plug of t h e salt pump i n t h e par'

t i c d a r system, 4 gpm t o t h e salt pump bearings, and t h e balance was bypassed back i n t o t h e o i l tanks and caused t o flow down khe outer tank w a l l s ,

219 This amount plus t h e warm o i l from t h e s h i e l d and bearings was cooled by means of water flow through c o i l s brazed t o t h e outside of t h e carbon s t e e l tanks. A s t h e system was subjected t o extended service, s c a l e b u i l t up on t h e

inside surfaces of these c o i l s , causing a drop i n t h e cooling capacity. This i n t u r n caused t h e temperature of t h e supply o i l - t o climb from i t s

This higher temperature w a s reached during March

normal 134'F t o 140'F. of 1966.

s c o i l s were flushed w i t h a 15-wgt % solution

Both of t h e lube o i l of a c e t i c acid.

Considerable material w a s removed, and t h e o i l tempera-

t u r e s , upon r e s t a r t i n g , were reduced 7'F

o i l system, and 3'F

i n t h e case of t h e coolant pump

i n t h e case of t h e fuel-pump o i l system.

During subsequent operatlon t h e temperature o f t h e f u e l pump o i l gradu a l l y rose, and i n August 1966, t h i s a c e t i c acid treatment w a s again given both coolers. This problem of gradual fouling of t h e water s i d e of t h e cooling c o i l s

r e curred throughout subs eque

operations, and by December 1967, t h e tem-

perature of t h e o i l supplied t o t h e pumps was up t o 150'F. ply was changed from tower

The water sup-

e r t o cooler untreated "process water" which

i n t h e case of t h e E R E was taken d i r e c t l y from t h e potable water system

via a backflow preventer.

The system w a s l e f t on process water f o r two

months and t h e heat t r a n s f e r improved enough, probably by physical flushing out of the s c a l e , so t h a t tower water could again be used and a f f o r d satisfactory cooling of t h e o i l

i s switching had t o be done at l e a s t twice

more before t h e r e a c t o r w

a l l y shut down i n December 1969.

permanent hose and piping d i t ious l y

9.8

A semi-

w a s provided s o t h i s could be done expe-

Increase i n Radiation Level at t h e Lube O i l Packages

A t the beginning of power.operation i n August 1969, a r a d i a t i o n moni-

t o r at t h e r e s e r v o i r of t

1pump o i l system i n d i c a t

t h e upper p a r t of t h e t

increasing and decreasing

U s i n g t h e on-site gamma

w a s 41Argon.

r a d i a t i o n from i t h r e a c t o r power.

t e r , it was determined t h a t t h e a c t i v i t y

This was apparently produced by a c t i v a t i o n of t h e blanket gas

i n t h e upper p a r t of t h e f u e l pump motor cavity.

Prior t o t h i s time only

helium had been used during power operation, and no gas activation had occurred. '

Argon was being used at t h i s time t o investigate bubbles i n t h e

fuel loop.

Since radiation levels did not 'exceed 2 mR/hr and decreased

when t h e cover gas was changed back t o helium, t h i s was not a serious problem.

9.9

Analysis of O i l

Samples were taken of each new supply drum and f r o m t h e two lube o i l systems after c i r c u l a t i n g f o r a short time.

Analyses of these were used as

controls, t o compare with periodic samples taken from t h e systems during operation.

The main concern was t h a t t h e o i l , especially i n t h e f u e l pump

system, might undergo some changes due t o long-time circulation through t h e high flux f i e l d i n the pump o r t h a t there might be some thermal decompo-

sition. Table 9.1 i s a compilation of some t y p i c a l o i l analyses performed during t h e l i f e of t h e MSRE.

The only s i g n i f i c a n t o r even minor changes

r a d i c a l increased somewhat and some C=O w a s formed, probably indicating some small degree of oxidation. T r i t i u m t h a t were found were t h a t the

buildup w a s insignificant. The o i l which leaked through t h e s e a l s t o t h e o i l catch tanks, although dark i n color, showed no s i g n i f i c a n t chemical o r physical changes. An addition f r a c t i o n a l d i s t i l l a t i o n t e s t analysis w a s run on a sample of new Gulfspin-35 o i l on September 2, 1964.

This was d i s t i l l e d under re-

duced pressure (3-4 mm Hg) i n an 18-in. reflux column.

As indicated below,

most of t h e d i s t i l l a t i o n occurred between 148 and 178OC. 9.10

Replacement of O i l

The first charge of Gulfspin-35 o i l was added i n September 1964 as t h e o i l tanks were calibrated.

Each tank took about 35 gallons.

Periodic ad-

ditions were made t o compensate f o r losses.

As indicated previously, no s i g n i f i c a n t physical ok -chemical changes occurred which would indicate t h a t t h e o i l s h o u l d b e replaced.

Various o i l

companies were contacted and they d i d not have any suggestions as t o other

221

Table 9 -

MSRE Sample No.

MSRE

Run No.

LO-1 thru LO-5 and LO-9 t h m LO-13

Date Sampled

Description

Viscosity Centistokes At 25OC

8125165

5 ea 55-bal drums new Gulfspin-35

15.44 to 15.99

LO-6

2118166

FOP-CONTROL

Lo-7

21la/66

COP-CONTROL

8/8/66

New Oil for FOP

8/13/66

New Oil for COP

Lo-47

Viscosity

At lOO'F

At 210'F

ssu

Carbon

85.52 to 86.43

9.94

2.64

86.88

-N O T A N A L Y Z E D

N O T A N A L Y Z E D -

Lo-49

LO-69

2113167

From COP

9.94

2.64

86.21

-70

2113167

From FOP

10.26

2.67

86.6

N O T A N A L Y Z E D

LO-140

5116168

-ul-141

5/16/68

** ***

LO-154

Sf 19169

From FOP

Solids ppm 180

9.94

2.58

78.0

LO-155

5/19/69

From COP

180

10.52

2.67

76.7

N O T A N A L Y Z E D

*References: MSRE Sample Log for Water, Helium, Miscellaneous; and MSRE Data Pile - Section 6E-1 (Lube 0 **Oil service from 5120167 to 5/16/68 - Collected in the Waste Oil Receiver from the fuel pump oil cat1 ***Oil in in service from 5120167 to 5/16/68 - Collected in the Waste Oil Receiver from the coolant pump oil

MSRE

Sulphur

GULFSPIN 35 LUBE O I L

Bromine Number

Flash Point

Sediment x v/v

- 0

TYPICAL ANALYSIS

1 -- Microp;rams/KL

SPECTROGRAPHIC ANALYSEd Fe La Mg Mo ti

1.56 to 1.95

316'F to 325'F

. e0.04

Ha+% 1.9

322'F

0.06 HaO-X n nn

0.05

1.55

Analyses) tank. tch tank.

_I_.

326'F 333'F

HaO-ppm 150 bo-ppm 200

<20

0.5

<0.6

<On6

250

<20

2.5

<0.6

136

<0.6

-- {-

108

---

124 <10

0.3

0.2

0.1

<0.4

<0.02

352 e10

0.51 ~0.4 1.9

0.2

<0.4

3.9 <I

---

520

--- ---- --

%ZOO

1.0

648

~0.4

Li 4.0

---

K 10 <0.5

5.0 %2

1.0 (0.5

<Os02

0.2

1.29

0.012

Comments and Miscellaneous

ANALYZED AS ESSENTIALLY TtI SAME AS LO-6 and LO-?

(all)

0.10

Tr 'a 5

'a 3

5.0

s1 1.0 s1 1 s1 10

3.4

-- --- --

---

-- --

4.6

130

---

200

2.0

Interfacial tension 15.4 (normal Interfacial tension 16.0 (normal

18) 18)

Dark appearance

%loo

Had C-0 band indication some oxidation

%loo

Had no C-0 band increase over new oil

222

means of checking f o r damage.

However, as protection against undetected

changes, t h e o i l was replaced i n August 1966 and again i n May 1967.

9.11

Discussion and Recommendations

Except f o r t h e priming d i f f i c u l t i e s w i t h t h e pumps and t h e moisture t r o u b l e s with t h e motors, t h e systems operated very s a t i s f a c t o r i l y . No noticeable change occurred i n t h e l u b r i c a t i n g o i l .

However, more

information i s needed as t o what chemical analysis o r physical t e s t s should be made t o d e t e c t undesirable changes.

A more r e a l i s t i c sampling schedule

needs t o be established. Improvements i n future heat exchangers f o r t h i s s e r v i c e should be made. Excess capacity should'be provided t o compensate f o r fouling of t h e heat t r a n s f e r surfaces o r better cooling water i n h i b i t o r s should be used.

223 10.

C,OIWONEN!T COOLING SYST?!NS P. H. Harley

The primary and secondary component coolant systems consisted of compressors and associated equipment required t o gas-cool various components i n t h e f u e l and coolant systems. Although they performed similar functions, , t h e two were completely independent and a r e therefore described separately. 10.1

10.1.1

Primary System

Description

Cell atmosphere gas (95% N2, 5% 02), commonly c a l l e d " c e l l air" w a s r e c i r c u l a t e d by one of two 75-hp belt-driven p o s i t i v e displacement pumps (component coolant pumps). a l l y s t a r t e d i f needed.

The o t h e r pump w a s a spare which could be manu-

Each pump was located i n a containment enclosure

o r dome which could be opened f o r maintenance.

After compression by the

component coolant pump, t h e c e l l a i r passed through a water-cooled s h e l l and-tube heat exchanger and then t o t h e reactor and drain tank c e l l s f o r d i s t r i b u t i o n t o t h e freeze valves, control rods, r e a c t o r neck, and fuel-pump cooling shroud.

A s i d e stream w a s c i r c u l a t e d past a r a d i a t i o n moni-

t o r t o i n d i c a t e c e l l - a i r a c t i v i t y and could be exhausted t o t h e containment stack f o r evacuating t h e c e l l s .

10.1.2

I n i t i a l Testing;

"he i n i t i a l t e s t i n g w a s primarily a check of t h e system design.

t o equipment were measured and interdependence of flows w a s checked. system was leak-tested and

Flows

The

heat t r a n s f e r c o e f f i c i e n t of t h e gas cooler

was measured. Permanent flowmeters were not i n s t a l l e d since most of t h e flows were s e t i n i t i a l l y and then not changed.

During i n s t a l l a t i o n of piping, various

flows were measured with temporar l y i n s t a l l e d rotameters.

t o t h e fuel-pump shroud nee s t a l l e d when t e s t s i n d i c a t e

Since a i r flow

e varied, a permanent o r i f i c e w a s in-

a i r loading on t h e control valve w a s not

an adequate indication of a i r flow. I n i t i a l flows were adequate except f o r cooling t h e control rods.

f i c i e n t flow w a s obtained by removing a discharge block valve.

Suf-

This block

224

valve w a s intended t o close on high c e l l - a i r a c t i v i t y t o prevent gross contamination of t h e c e l l i n case a control rod thimble ruptured.

I n l i e u of

the block valve, t h e discharge a i r was routed through a 55-gallon s t a i n l e s s s t e e l drum located i n t h e r e a c t o r c e l l t o r e t a i n any salt which leaked due t o the rupture. The measured flows a r e given i n T a b l e 10.1

Tests indicated t h a t chang-

ing one flow had no e f f e c t on t h e o t h e r flows even when t h e c e l l was being evacuated at about 200 scfm.

The pressure control valve (PCV-960) which

discharges excess a i r back t o t h e c e l l adequately compensated f o r t h e varying flows and maintained a constant discharge pressure. 10.1.3

Blower Capacity

Design of i n - c e l l MSRE components i n i t i a l l y required a gas-cooling capacity of 885 scf'm.

However, development t e s t s indicated t h a t cooling

air would not be required on t h e freeze flanges.

Therefore, t h e blower

capacity was lowered t o 6 1 7 scfm by i n s t a l l i n g a smaller sh'eave on t h e

drive motor. After about two-years operation, more c i r c u l a t i o n was deemed necessary between drain tank c e l l and r e a c t o r c e l l .

This was t o give a more rapid

response on t h e c e l l air a c t i v i t y monitors i n case of a leak i n t o t h e drain tank c e l l .

A l a r g e r diameter drive sheave w a s i n s t a l l e d t o give a blower

capacity of 740 scf'm and a 2-in.

l i n e w a s i n s t a l l e d from t h e freeze valve

supply header t o t h e drain tank c e l l .

This gas (%200 scfm) flowed from t h e

drain tank c e l l i n t o t h e r e a c t o r c e l l and back t o t h e component coolant pump suction.

This

740 scf'm w a s

used f o r all subsequent operation.

During

rapid evacuation of t h e c e l l , it w a s necessary t o close a hand valve i n t h i s l i n e i n order t o maintain an adequate flow t o other components. 10.1.4

Gas Cooler

The heat removal by t h e gas cooler was adequate throughout t h e operation.

The i n i t i a l heat t r a n s f e r c o e f f i c i e n t calculated t o be

f't2 OF w i t h a flow rate of 885 scfm.

After two years of operation and a t

a flow r a t e of 740 scfm, it was 28 Btu/hr f t 2 proximately t h i s same value.

36.5 Btu/hr

and has remained at ap-

No inspection has been made t o determine how

much o i l and/or b e l t dust i s deposited on t h e gas s i d e ( s h e l l s i d e ) of .the tubes.

225

Table 10.1 Flaw Distribution of Primary Component Cooling System

Measured Flow scf’m

Equipment Supplied

No. 1 Control Rod Motor

1.33

No. 1,Control Rod

3.8

Reactor Neck Inside

1.45 3.9 1.4 3.9 18 18

Graphite Sampler

16.2

No. 2 Control Rod Motor No. 2 Control Rod No. 3 Control Rod Motor

No. 3 Control Rod Reactor Neck Outside

FV-104

Design

5 5

5 15

15 15

FV-104

26.6

FV-105

27.7 24.8

15 15 15

FV-106 FV-107 FV-108

15 15

FV-109 Cell A i r Radiation Monitor (~!&565)

100

W52* ** Fuel Pump

260

*** PdcV-960

100

*L-952

was i n s t a l l e d i n September 1966 t o increase circul a t i o n between reactor and drain tank c e l l s .

**I n i t i a l

design I n i t i a l normal use w a

***PCV-960

scfm was lowered t o 100 scf’m. cfm but l a t e r no flow w a s required.

discharged excess capacity t o maintain a cons t a n t discharge pressure.

226

10.1.5

CCP Drive Units

Originally t h e component coolant pumps were driven by f i v e matched b e l t s using a 15.9-in.

drive sheave an@ a 21-in.

drive sheave.

blower capacity w a s reduced by changing t o a 10.9-in. d i f f i c u l t i e s developed.

When t h e

drive sheave, b e l t

The first s e t of b e l t s f a i l e d after 1450 hours of

operation and t h e second s e t only 11 hours later.

The smaller diameter

drive sheave increased the bending s t r e s s e s t o values i n excess of t h e b e l t To correct t h i s , i n February 1966, t h e drive and drive sheave dia-

ratings.

meters were increased t o 12.5 and 24 in. respectively.

To obtain a b e t t e r

b e l t loading, t h e i n i t i a l matched b e l t sets were replaced by a poly-V-belt. Then i n August 1966, the drive sheave was increased t o 15-in. diameter t o increase blower capacity.

This f u r t h e r reduced bending s t r e s s e s on the

belt e The original b e l t s were rated a t 200째F whereas the poly-V-belts were

only rated at 130째F.

This caused d i f f i c u l t y since the ambient temperature

i n the domes was about 170째F due t o motor heat and leakage of 320째F gas from t h e discharge relief valves.

hsi

Rupture discs were i n s t a l l e d t o prevent

leakage but they would not withstand t h e shock received when t h e blowers were s t a r t e d .

Baffles were then i n s t a l l e d which deflected t h e cooler in-

coming gas across the b e l t s and thus lowered t h e ambient temperature t o

around 150'F. To magnify t h e drive b e l t d i f f i c u l t i e s , t h e motor support w a s not satisfactory.

The motor could s l i p on i t s support p l a t e and flexing of t h e

motor support legs caused a pulsating load on t h e drive b e l t s .

Small blocks

were welded t o t h e support bracket t o keep t h e motor from slipping and heavy bracing of t h e stand minimized t h e vibration.

Since t h e l a r g e s t l s t r a i n m-

curred on t h e drive b e l t s when t h e blower w a s s t a r t e d , one blower w a s operated continuously during a run instead of a l t e r n a t i n g t h e blowers twice a month as was done i n i t i a l l y .

This increased t h e b e l t l i f e t o more than

8000 hours and uninterrupted runs of over 3000 hours were made without b e l t

adjustments 10.1.6

O i l System

One of t h e component coolant pumps (CCP-1) had t o be stopped during

Run 10 i n January 1967 because of low o i l pressure, and CCP-2 w a s used f o r

227

t h e remainder of t h e run.

Af'ter Run 10, a cracked copper f i t t i n g i n t h e

o i l system was repaired, and 2 gallon of o i l w a s added t o bring t h e o i l l e v e l back t o normal.

Drains were a l s o

n s t a l l e d t o r e t u r n any o i l s e a l

leakage back t o t h e o i l reservoir. More trouble was encountered i n t h e CCP-1 o i l c i r c u l a t i n g system during Run 12.

F i r s t , a loose tubing connection c a u s e d t h e loss of ~2 gal. of o i l .

This was repaired i n August $1967. When t h e blower was r e s t a r t e d , intermit-

t e n t low oil-pressure alarms again occurred,

An investigation indicated

no s i g n i f i c a n t l o s s of o i l , but a slow oil-pressure response w a s observed when t h e blower was s t a r t e d .

The suspected o i l pump and pressure-relief

valve w e r e replaced with spare units t o correct t h e trouble.

Later inspec-

t i o n indicated t h a t t h e o i l pump was not damaged but t h e pressure-relief valve was r e l i e v i n g before the normal o i l pressure developed.

I n s p i t e of

t h i s d i f f i c u l t y , adequate l u b r i c a t i o n had been maintained and t h e r e was no

damage t o t h e blower,

The o i l pressure r e l i e f valve on CCP-2 was inspected

and a l s o found t o r e l i e v e a t too l o w a pressure and was readjusted. I n September 1967, a f t e r s i x days of flush-salt c i r c u l a t i o n and two

days of nuclear operation, CCP-2 was shut down by low o i l pressure.

Since

t h e discharge valve would not close leak-tight, t h e run was terminated. Repairs consisted of repairing t h e leaking discharge valve, capping a leaking drain l i n e and tightening several packing nuts and f i t t i n g s . blower, CCP-1,

w a s used f o r 873 hours i n t o November 1967.

The other

A t t h i s time a

serious o i l leak was indicated by low o i l pressure and accumulation of about

2 gal. of o i l i n t h e condensate c o l l e c t i o n tank connected t o t h e blower containment.

The standby u n i t , CCP-2,

w a s immediately s t a r t e d up and oper-

ated without further d i f f i c u l t i e s throughout t h e remainder of t h e run,

3,036 hours. I n J u l y 1968, t h e brazed tubing o i l system on both blowers was replaced by welded pipe and blems w i t h t h e o i l system.

l e hose.

This eliminated most of t h e pro-

eaky shaft o i l s e a l had t o be replaced i n

December 1969.

10.1.7

Strainer

I n August 1966, a temporary s t r a i n e r with a 1/16-in, screen was in-

s t a l l e d i n t h e component-cooling-pump discharge l i n e t o c o l l e c t rubber dust

228

from t h e drive b e l t s , and prevent blowing t h i s material onto t h e hot metal components being cooled.

Although t h e temporary s t r a i n e r worked s a t i s f a c -

t o r i l y a new s t r a i n e r made of 100-mesh screen, which had been on order f o r

a year, w a s i n s t a l l e d i n t h e l i n e i n May 1967.

Over an eight-month period,

t h e temporary s t r a i n e r had accumulated 30 t o 50 g of black, dry powdery ma-

t e r i a l t h a t appeared t o be dust from abrasion of t h e drive b e l t s .

After

being decontaminated, the s t r a i n e r w a s examined and was found t o be i n very good condition.

The surface was s l i g h t l y etched, but no more than would

be caused by t h e decontamination process. The new s t r a i n e r developed excessive pressure drop i n September 1967

and t h e screen was replaced.

I n 1,800 hours of operation t h e 100-mesh screen

had become p a r t i a l l y plugged with rubber dust f r o m t h e d r i v e b e l t s .

It w a s

damaged i n removal and a replacement basket w a s made using 16-mesh, 0.023in. w i r e screen.

This showed no pressure buildup over t h e subsequent 2

years of operation.

10.1.8

Condensate Collection

A f t e r a reactor c e l l space cooler began t o leak water i n t o t h e c e l l i n

March 1966, condensate accumulated i n t h e 10-in. suction l i n e t o t h e component coolant pumps at a rate of 1 t o 2 gpd.

The water w a s vaporizing i n

the reactor c e l l and condensing on cold surfaces at t h e component coolant pumps. Simple drains were i n s t a l l e d , but these could be used only when the reactor was s u b c r i t i c a l due t o r a d i a t i o n i n t h e coolant d r a i n tank c e l l . Handling of t h e drained water w a s complicated by the tritium (up t o 915 vc/ml. produced from the 6 L i i n t h e treated-water corrosion i n h i b i t o r o r from diffusion through primary piping.

During t h e shutdown i n August 1966,

piping w a s i n s t a l l e d t o permit draining t h e domes during power operation t o a tank i n t h e sump room.

It was periodically pumped from t h i s tank t o

t h e l i q u i d w a s t e tank. l0,l.g

Electrical

I n June 1966, t h e wiring t o CCP-1 drive motor shorted out i n t h e penet r a t i o n i n t o t h e containment enclosure.

The s h o r t t r i p p e d t h e main breaker

t o generator bus No. 3 as w e l l as t h e component coolant pump breaker. pump and other equipment on bus No. 3 stopped.

The

Emergency d i e s e l power was

229

bsi

s t a r t e d immediately, but because of a misinterpretation of t h e problem, the r e a c t o r drained before cooling was restored t o t h e freeze valves by switching t o CCP-2.

The short occurred i n an epoxy s e a l on t h e end o f t h e

copper-sheathed mineral insulated cable and w a s caused e i t h e r by moisture from i n - c e l l leakage o r a breakdown of t h e epoxy p o t t i n g compound. Wiring changes were made t o both component cooling drive motors t o prevent a reoccurrence.

Each phase of t h e t h r e e phase c i r c u i t s was brought

through a separate penetration and t h e epoxy w a s omitted from t h e end s e a l s .

No f u r t h e r d i f f i c u l t i e s were encountered. 10.1.10

Valve Problems

I s o l a t i o n valves were provided on each component coolant pump t o per-

m i t r e p a i r i n g one while the other w a s i n operation. (10 in. on t h e i n l e t and

These l a r g e valves

6 i n . on t h e o u t l e t ) had back-seating stems on t e f -

lon seats t o prevent leakage when t h e valves were open. be very s a t i s f a c t o r y .

A f t e r about 3-years service, one of t h e 6-in.

l a t i o n valves leaked i n excess of 100 cfd.

These p r o v e d t o iso-

An fnspection revealed several

l a r g e p i t s i n t h e s e a t i n g surface of t h e valve which had been f i l l e d with epoxy during manufacture.

These were repaired by f i l l i n g w i t h weld metal

and then r e f i n i s h i n g t h e seating surface.

When reassembled, t h e leakage

through t h e valve was $0.5 cfd which was considered t o be s a t i s f a c t o r y . Check valves were i n s t a l l e d on t h e discharge of t h e blowers t o prevent These were 6-in. butterfly-type valves w i t h s i l i c o n e rubber hinges holding two flappers i n place. I n January 1966,

backflow through t h e stand-by unit.

after 1640 hours of s e r v i c

he hinge on t h e CCP-2 valve broke allowing

one flapper t o f a l l o f f ,

pares were available s o flappers and hingers

were made at ORNL.

This had not f a i l e d a f t e r 1800 hours but w a s replaced

by factory-built p a r t s .

I n November 1966, t h e b u t t e r f l y check valve from

CCP-1 was found t o be inoperative and was repaired.

The f a i l u r e a l s o oc-

curred at t h e s i l i c o n e rubber hinge which supports t h e two wings of t h e check valve. On one occasion during t h e period when b e l t problems were causing l o s s of blower capacity, t h e stem of t h e i n - c e l l pressure control valve (PdCV960) froze i n a p a r t i a l l y open position and f a i l e d t o close when t h e valve

operator a i r loading increased.

This was t h e only f a i l u r e of an i n - c e l l

230

component i n t h i s system.

The valve was removed remotely and repacked.

The stem was polished and l i g h t l y l u b r i c a t e d before r e i n s t a l l a t i o n .

prevent t h e valve from remaining i n one t h r o t t l e d p o s i t i o n f o r an extended

t i m e , t h e pressure c o n t r o l l e r was subsequently cycled periodically t o be sure t h e valve w a s functioning.

Moisture i n t h e c i r c u l a t i n g a i r probably

contributed t o t h e valve failure, and t h e periodic draining of t h e collected moisture has helped prevent a reoccurrence. 10.1.11

Conclusions and Recommendations

Although many di'fficulties have been encountered with equipment, only two r e a c t o r drains have been caused d i r e c t l y by component cooling system failures.

One w a s l o s s of one blower while the other w a s i s o l a t e d f o r main-

tenance and t h e second w a s a f a i l u r e of e l e c t r i c a l cable t o t h e motor of one compressor (CCP-1).

One run w a s r e s t a r t e d a f t e r r e p a i r i n g an o i l leak

which.developed early in the run.

Better design s p e c i f i c a t i o n s would have assured more serviceable items. For instance, t h e pressure relief valves had a s p e c i f i e d capacity a t 1 5 psig, however, they discharged a small amount of gas continuously a t norm a l operating pressure.

containment domes.

This caused excess heating of t h e equipment i n t h e

Combining overpressure protection w i t h an unloading

valve and controls would have permitted blower s t a r t u p without load and caused less w e a r on the drive b e l t s .

A more durable hinge would have pre-

vented t h e check valve f a i l u r e s . More instrumentation on gas flows and on t h e o i l system would have given earlier indication of pending d i f f i c u l t i e s .

O i l sump l e v e l and dis-

charge pressure with indicators outside t h e containment would have helped here. Although s u f f i c i e n t capacity was a v a i l a b l e , only minimum capacities

were available f o r cooling at some locations because of l i m i t i n g pressure drops of piping and control valves.

More design consideration should have

been given t o t h e p o s s i b i l i t y of needing a d d i t i o n a l cooling.

A f t e r cor-

r e c t i n g blower drive problems during t h e first two years of operation, sev-

eral r e a c t o r runs were made i n excess of 3000 hours w i t h only preventative maintenance between runs.

231

10.2

Secondary System

,1 Description

Ambient a i r f o r cooling out-of-cell

~ O - S Cp ~o s i t i v e

pressor (AC-3).

components w a s supplied by e i t h e r

displacement pump (CCP-3)

o r a piston-type a i r com-

"he a i r was d i s t r i b u t e d t o t h e coolant system drain freeze

valves and t o t h e chemical processing plant freeze valves. 10.2.2

Operating; Experience

Early t e s t i n g revealed t h a t t h e flow t o FV-204 and FV-206 was insufThe air-operated

f i c i e n t t o e s t a b l i s h good plugs i n t h e freeze valves.

control valve t r i m w a s enlarged t o reduce t h e pressure drop and t h e blower discharge pressure w a s increased from 8 t o 9.25 psig.

With these changes,

adequate a i r capacity, 25 scfh, was obtained t o each freeze valve. I n April 1967, t h e drive s h a f t of t h e r o t a r y blower, CCP-3,

seized i n

t h e brass sleeve bearing of t h e blower housing and d i d extensive damage t o t h e drive s h a f t .

Operation was continued using t h e service a i r compressor,

ACT-3, which w a s t h e standby supply t o keep FV-204 and FV-206 frozen.

second bearing f a i l u r e occurred i n CCP-3 a f t e r only 200 hours of operation following t h e first f a i l u r e .

Since t h a t time, AC-3 has been used as t h e

normal cooling a i r supply with CCP-3 as a standby unit.

CCP-3 was used

during programmed maintenance on AC-3 and while freezing FV-204 and FV-206 when more a i r w a s required than t h e a i r regulator from AC-3 would supply. A brass swinging gate check valve which prevented backflow through CCP-3 when t h e service a i r

ressor was used, failed because of wear

caused by pulsating flow wh

CP-3 w a s i n service.

replaced by a hand valve.

This check valve was

Then, when t h e a i r compressor was changed t o t h e

normally operated supply, a new check valve w a s .installed s o t h e blower could be r e s t a r t e d remotely i f a low header pressure a l a r m occurred.

The

hand valve was l e f t i n s t a l l e d so t h e blower could be i s o l a t e d i f t h e check valve f a i l e d again. 10.2.3

Conclusions

The p o s i t i v e displacement pump (CCP-3)

air mover functioned poorly.

which was intended as t h e prime

However, t h e service a i r compressor (AC-1)

was very r e l i a b l e and thus l i t t l e o r no operating time was l o s t due t o t h i s system.

232

11.

VENTILATION SYSTEM P. H. Harley

11.1

Description

The purpose of t h e v e n t i l a t i o n system w a s t o exhaust a i r from areas where there was a p o t e n t i a l danger of release of radioactivity.

The a i r

flowed through a network of ducts t o t h e f i l t e r p i t which contained 3 paral-

l e l banks of roughing f i l t e r s and absolute f i l t e r s .

It w a s pulled through

t h e f i l t e r p i t by one of t h e two stack fans and w a s then discharged up a

100-ft stack t o t h e atmosphere.

The stack was continuously monitored t o Sen-

assure t h a t t h e r a d i o a c t i v i t y released w a s within acceptable limits.

s i t i v e d i f f e r e n t i a l pressure gages w e r e i n s t a l l e d t o i n d i c a t e t h e d i r e c t i o n of.air flow between t h e various areas. 11.2

OperatinK Experience

After i n s t a l l a t i o n of t h e v e n t i l a t i o n system w a s complete, a series of t e s t s w a s run t o insure t h a t t h e c r i t e r i a would be met. Several minor changes were made i n t h e ducting, c e l l blocks were caulked, and leaks repaired.

The biggest problem encountered w a s s e a l i n g t h e hi-bay.

Gaskets

were i n s t a l l e d on t h e doors, all j o i n t s i n t h e l i n i n g were taped, and a number of l a r g e holes were plugged. ured as follows.

The leakage i n t o t h e hi-bay w a s meas-

The dampers i n t h e v e n t i l a t i o n ducts from all other areas *-

were closed.

All i n l e t vents and doors t o t h e hi-bay were closed.

The

dampers between t h e hi-bay and t h e stack fans were t h r o t t l e d u n t i l t h e hibay pressure was 4 . 3 inches of water. found t o be 4250 ft3/min.

The in-leakage ( s t a c k flow) was

Although t h i s w a s above t h e design value of

1000 ft3/min, it w a s accepted. The p i t o t t d e stack gas flow meter was c a l i b r a t e d using a hot wire anemometer.

Several t r a v e r s e s were made under various f l o w conditions.

The flow under normal operating conditions w a s found t o be around 23,000 cFm.

The stack c a l i b r a t i o n curve and d e t a i l s o f t h e c a l i b r a t i o n a r e given

i n MSRE Test Memo and T e s t Report 2.3.15. Tests indicated t h a t adequate v e n t i l a t i o n could be maintained with 2 of t h e 3 p a r a l l e l f i l t e r banks i n service. 9

Therefore, it w a s possible t o

replace one bank of f i l t e r s without i n t e r r u p t i n g operations.

I \ W

233

After i n s t a l l a t i o n , t h e absolute f i l t e r s d i d not pass t h e standard ORNL-DOP smoke test.45

(99.95% e f f i c i e n c y i s required.)

Visual inspection

indicated leakage around t h e frames which required considerable caulking and some minor revisions.

On 11/14/64 a l l t h r e e banks passed t h e DOP t e s t

(East = 99.970%; Center = 99.972%; and West = 99.967% e f f i c i e n t ) .

The next

t e s t i n May of 1965 indicated t h a t a l l t h r e e banks had dropped t o less than 99.90% e f f i c i e n t .

Therefore, they were removed and extensive recaulking

and revisions were made t o could be replaced remotely.

rove t h e i r efficiency and assure t h a t they e results of subsequent DOP tests a r e shown

i n Table 11.1. It has not been necessary t o replace t h e absolute f i l t e r s since 1965.

The pressure drop through them had changed very l i t t l e during t h e 4 years of operation which indicates t h a t t h e roughing f i l t e r s performed well and removed most of t h e p a r t i c u l a t e s . The roughing f i l t e r s were changed when t h e pressure drop across them I

reached 5 t o 6 inches of water at which time t h e s t a c k flow would be down t0-'~18;000cfm.

During e a r l i e r operation while considerable maintenance

and construction was s t i l l being done, these plugged r a t h e r rapidly.

Small

e a s i l y changed f i b e r g l a s s f i l t e r s i n s t a l l e d on s e v e r a l of t h e i n l e t ducts reduced t h e plugging r a t e . 11.2.

Subsequent f i l t e r changes a r e shown i n Table

The l a s t set of roughing f i l t e r s read %5OO mEt/hr a t 3 inches a f t e r

they were removed. Some d i f f i c u l t y was encountered i n e a r l y operation w i t h s t a c k fan No. 1 (SF-1) bearings.

The bearings were i n i t i a l l y grease-lubricated and t h e r e

w a s d i f f i c u l t y i n maintaining proper l u b r i c a t i o n ,

was about four months,

The normal bearing l i f e

I n 1966, an o i l - l u b r i c a t i n g system was i n s t a l l e d

and t h e subsequent bearing l i f e has been about 18 months. continuously except when necessary t o do maintenance on it.

SF-1 was run SF-2 served

as a standby u n i t . Sensitive d i f f e r e n t i a l pressure gages ( f u l l s c a l e = 2 inches of water) were provided t o indicate d i r e c t i o n of a i r flow.

These were recorded each

s h i f t on t h e building l o g and dampers were adjusted as necessary.

Venti-

l a t i o n flow was from t h e less hazardous t o t h e more hazardous area except

h.,

i n t h e coolant c e l l during power operation.

When t h e main r a d i a t o r blowers

were operating, considerable air leaked from t h e r a d i a t o r and pressurized

234

U Table 11.1

Results of DOP Tests of Absolute F i l t e r s

Efficiency Bank

East Center West

99 0999% 99 994 99.994 9

3/66

6/67

99 998% 99 -995 99 997

99 979% 99 -998 99 994

Table 11.2

99.994% 99 9% 99 994

99 993% 99 995 99 985 9

Roughing F i l t e r Changes

East -

Center

West -

October 1964

October 1966

August 1967

Septeniber 1968 August 1969

X X

235

t h e coolant c e l l .

In an e f f o r t t o correct t h i s , t h e i n l e t t o one of t h e

r a d i a t o r a u x i l i a r y blowers (MB-2) was connected t o t h e coolant c e l l .

This

reduced t h e pressure below atmospheric but it w a s s t i l l above t h e hi-bay pressure.

Since t h e only hazard was fromberyllium and since t h e coolant

stack was continuously monitored f o r beryllium, t h i s w a s considered acceptable. During maintenance, containment i s provided by t h e flow of a i r i n t o the cells.

The system has adequately supplied t h i s need f o r a l l mainte-

nance operations. There has been very l i t t l e a c t i v i t y released t o t h e atmosphere except during maintenance periods. ORNL permissible limits.

Even then, t h e releases have been below t h e

The permissible r e l e a s e r a t e f r o m t h e 5 ORNL

stacks i s %l c u r i e per week.

Table 11.3 i s a tabulation of t h e amount of

a c t i v i t y released from t h e MSRE stack.

The maximum released i n any week

at t h e MSRE was 568 m C i during t h e week ending December 21, 1969.

This

w a s mainly from t h e core specimen removal plus about 50 m C i from t h e p r i -

mary system leak which occurred during t h e f i n a l shutdown.25 Due t o a galled b o l t , t h e core specimen hold-down flange could not be r e i n s t a l l e d and was l e f t hanging i n t h e standpipe. s t a l l e d t o close t h e system.

Another flange was in-

S a l t on the hold-down assembly probably con-

tributed t o t h e l a r g e r than usual release.

The standpipe vacuum w a s a l s o

inadvertently l e f t o f f on two occasions when it should have been on.

11.3 Conclusions and Recommendations The v e n t i l a t i o n system has functioned.satisfactorilyduring operation and maintenance.

The i n s t a l l t i o n of e a s i l y removable f i l t e r s a t i n l e t s t o

t h e v e n t i l a t i o n ducts prolonged t h e l i f e of t h e roughing f i l t e r s a n d ' s h o u d

be considered i n f u t u r e designs. The stack monitoring system could be improved. were very d i f f i c u l t t o m a k e notes on.

The recorders used

This, together with t h e range chang-

ing f e a t u r e of t h e system, made l a t e r i n t e r p r e t a t i o n of t h e charts very difficult.

I n addition t o t h i s , t h e monitors were never calibrated.

This

should have been done and limits on t h e i r readings should have been est a b l i s h e d before s t a r t i n g operation.

236

U Table 11.3

Dates

Particulate

Gaseous

mCi

From

3/66 9/66

8/66

2/67

206

3/67

8/67

<0.3

4.0

9/67

2/68

2.3

3/68

8/68

<0.3 <0.3

9/68

2/69

<0.3

3/69

8/69 1/70

~0.3

9/69

Semiannual Stack Release

<0.3

343 0.4 12.5 8.2

Main Source of Activity Miscellaneous maintenance Off-gas line maintenance

Off-gas line maintenance

OFT vent maintenance Removal of core specimens and primary system leak

237 WATER SYSTEMS

12.

P. H. Harley

Water f o r various types of usage w a s supplied at t h e MSRE by t h e following systems:

(1)t h e potable water system d i s t r i b u t e d ORNL water f o r

f i r e protection, general building services, and was t h e supply f o r t h e process water system, ( 2 ) t h e process water system d i s t r i b u t e d water (potable water which had gone through a backflow preventer t o provide i s o l a t i o n ) f o r process usage, such as d i l u t i o n of t h e l i q u i d waste, f o r temporary o r emergency cooling and f o r makeup t o t h e cooling tower, ( 3 ) t h e cooling tower

water system r e c i r c u l a t e d t r e a t e d process water through various out-of-cell components and provided cooling f o r t h e t r e a t e d water, ( 4 ) t h e t r e a t e d wat e r system r e c i r c u l a t e d t r e a t e d steam condensate i n a contained loop t o provide cooling f o r a l l i n - c e l l o r contained components, ( 5 ) t h e steam condensate system condensed ORNL steam t o produce pure condensate which was used i n t h e drain tank a f t e r h e a t removal system and f o r t r e a t e d water system makeup, and (6) t h e nuclear instrument thimble water system recirculat e d t r e a t e d steam condensate t o provide cooling f o r t h e nuclear instruments

b.,

and b i o l o g i c a l shielding f r o m t h e r e a c t o r vessel. The operating experience, d i f f i c u l t i e s encountered, and recommenda-

t i o n s a r e given separately f o r each sub-system. . 12.1

Potable Water System

Two six-inch cast i r o n l i n e s supply w a t e r t o t h e MSRE area from t h e

supply main,

One l i n e suppl

water t o t h e f i r e sprinkler system and

general building services and t h e o t h e r supplies makeup t o t h e process wat e r system.

The underground supply l i n e t o t h e f i r e s p r i n k l e r system and building services ruptured i n t h e f a l l of 1967.

Repairs required s h u t t i n g off both

potable water supplies t o t h e a r e a so process water makeup w a s supplied by

a f i r e hose from a vent l i n e on t h e valley water main.

F i r e protection w a s

provided by c1/4 mile of f i r e hose from t h e Nuclear Safety P i l o t Plant and posting of a f i r e watch a t t h e l o c a l f i r e alarm box i n t h e building.

Iso-

l a t i o n valves where t h e l i n e s t e e d o f f t h e water main would have simplified

t h e - r e -p- a i r s . ground.

I n December 1969, t h e other 6-in. supply l i n e ruptured under-

Process water was supplied from t h e other 6-in. main through a

238 temporary backflow preventer during repairs.

I n both cases, r e p a i r s were

made using a s t a i n l e s s s t e e l A d a m clamp. Although no serious malfunctions have occurred i n t h e automatic f i r e The

sprinkler system, a b e t t e r designed system could have been provided. MSRE system was i n s t a l l e d from national f i r e protection specifications

without regard t o area services.

I n some areas containing e l e c t r i c a l equip-

ment, more &anger existed from possible water damage than f r o m t h e possib i l i t y of fire.

I n these areas the s p r i n k l e r s were disconnected o r iso-

l a t e d with manual valves.

Portable extinguishers s u i t a b l e f o r e l e c t r i c a l

f i r e s have been provided i n areas containing e l e c t r i c a l equipment. 12.2

Process Water System

This system has performed very w e l l .

On a f e w occasions, process wa-

t e r was used f o r cooling c r i t i c a l equipment when r e p a i r s necessitated shutt i n g down t h e cooling tower water system. Process water which w a s %20째F cooler than t h e cooling tower w a t e r was e f f e c t i v e i n cooling t h e f u e l pump and coolant pump o i l tanks when t h e c o i l s became fouled with scale.

The

use of process water tended t o f l u s h t h e c o i l s so t h a t adequate cooling could be obtained using cooling tower water. The process water system w a s supplied by ORNL potable water through one backflow preventer.

Another backflow preventer was i n s t a l l e d i n s e r i e s

with t h i s i n t h e process water supply t o t h e l i q u i d waste tank. backflow preventers were scheduled t o be t e s t e d semiannually. i n t occurred a t i n t e r v a l s of 3 t o 14 months. never required any repairs.

These Actual t e s t -

The main backflow preventer

"he one i n t h e l i n e t o t h e waste tank required

minor r e p a i r s on two occasions.

Valving w a s provided t o i n s t a l l a second

u n i t i n p a r a l l e l i f r e p a i r s of t h e main operating backflow preventer were required.

However, t h e second u n i t w a s never i n s t a l l e d .

cess water flow was shut o f f t o make t h e periodic tests.

Instead t h e proFor a more com-

plex water system, two backflow preventers should be i n s t a l l e d . 1

239

12.3

Cooling Tower Water System

Standard c e n t r i f u g a l pumps were used t o c i r c u l a t e t h e cooling tower water.

Flow was adequate and no pump performance t e s t s were made.

The

cooling tower w a s a packaged commercial u n i t and provided adequate cooling. Very adequate temperature control w a s obtained using a temperatyre cont r o l l e d valve which allowed p a r t of t h e water t o bypass t h e cooling tower. One major pipe leak occurred which required shutting t h e e n t i r e system down. This leak w a s i n t h e underground six-inch r e t u r n l i n e t o t h e cooling tower.

Nalco-360 b a l l s , a mixture of sodium and potassium chromate containing some phosphates, were used as a corrosion i n h i b i t o r . I n t h e summertime, a one-pound b a l l of Nalco-2lS, an algae i n h i b i t o r , w a s added d a i l y and w a s s a t i s f a c t o r y i n preventing algae formation on t h e

cooling tower.

Hardness was controlled by bleeding o f f about 2 gal. per

minute of cooling tower water.

This water plus about 8 gallons per minute

of process water w a s used t o cool t h e charcoal beds. Over 1500 samples of cooling tower water were taken and analyzed.

Most of these were analyzed at t h e s i t e t o assure adequate treatment. odically t h e samples were sent t o t h e laboratory f o r i r o n analysis. remained l e s s than 1 ppm throughout the operation.

PeriThis

Due t o t h e continual

makeup and bleedoff f r o m t h e system, good corrosion r a t e figures a r e not available, however, no .leak has occurred which could be a t t r i b u t e d t o corrosion. A considerable amount of calcium phosphate s c a l e coated t h e rotameter

tubes and cooling c o i l s and

ludge i n t h e cooling tower basin and low

velocity areas of t h e t r e a t

r cooler.

t e r a solution of p o t a s s i

mate and zinc-sulfate i n h i b i t o r i n t o t h e

cooling tower water by a prop makeup water flow.

d e n this

i n i n h i b i t o r w a s abandoned. t o f a c i l i t a t e reading and

Equipment was i n s t a l l e d t o me-

ioning pump powered by t h e cooling tower f a i l e d t o operate propertly, t h e change Lights were i n s t a l l e d behind t h e rotameters e frequency of cleaning.

240

12.4 12.4.1

Treated Water Systems

I n i t i a l F i l l and Operation

The t r e a t e d water system w a s f i l l e d by adding steam condensate and corrosion i n h i b i t o r batchwise.

The t o t a l volume w a s about 4000 gallons

with about 2300 gallons i n t h e thermal shield.

12.4.2

Circulating PumpS

Standard c e n t r i f u g a l pumps were used t o c i r c u l a t e t h e t r e a t e d water i n a closed loop. were made.

The capacity w a s adequate and no pump performance t e s t s

The pumps have operated s a t i s f a c t o r i l y w i t h l i t t l e maintenance

other than repairing t h e r o t a t i n g seals.

Leakage from these was collected

i n p l a s t i c b o t t l e s due t o t h e induced a c t i v i t y i n t h e water.

12.4.3

Thermal Shield

I n i t i a l t e s t i n g i n d i c a t e d t h a t eFcessive pressure would e x i s t i n t h e thermal s h i e l d and that the flow was not adequate t o t h e thermal s h i e l d s l i d e s (sections of t h e s i d e s which could be removed f o r replacing t h e rea c t o r vessel).

Piping changes were made t o lower t h e back pressure and t o

provide a separate supply l i n e and pump t o assure adequate flow i n t h e slides.

During e a r l y operation, several rupture d i s c s i n t h e thermal I

s h i e l d discharge l i n e s were ruptured due t o t h e supply and discharge valve closing simultaneously.

To prevent t h i s , t h e supply valve w a s modified so

t h a t it would close before t h e discharge valve closed. The water level w a s noted t o increase i n t h e surge tank when t h e rea c t o r power w a s increased and t h e l e v e l decreased when t h e r e a c t o r w a s subcritical.

Decomposition of t h e water was producing %2 cubic feet per hour

of H2 gas a t full power and about 8 f t 3 of t h i s gas collected i n one of t h e TS s l i d e s i n which it i s postulated t h e water piping was reversed, i.e., water entering t h e t o p and leaving from a d i p l e g near t h e bottom.

A small

deaerating t a n k , i n s t a l l e d i n the water supply t o t h e TS s l i d e permitted

%2 f t 3 of gas t o redissolve when t h e system reached equilibrium.

Later a

l a r g e degasing tank was- i n s t a l l e d t o deaerate t h e water leaving t h e TS t o permit more H2 t o be dissolved.

This reduced tlie accumulate gas t o Q3 f t 3 .

I n i t i a l l y a purge of 32 and later a i r was introduced i n t o t h e surge tank

241

bi'

and degasing tank t o flush t h e gas space and d i l u t e t h e Hp t o below t h e explosive l i m i t .

No f u r t h e r d i f f i c u l t i e s were encountered, however,

%3/4-gal. of water/day w a s l o s t by evaponation i n t h e purge gas. 12.4.4

Reactor and Drain Tank Cell Space Coolers

Two space coolers were i n s t a l l e d i n t h e r e a c t o r c e l l and one i n t h e drain tank c e l l t o keep t h e c e l l a i r temperatures below 150째F.

The meas-

ured capacity was only 40 t o 60 kW compared t o t h e design value of 75 kW. However, t h e c e l l temperatures normally ran at about 140째F f o r t h e r e a c t o r c e l l and 130째F f o r t h e drain tank c e l l . The 3-hp fan motors o r i g i n a l l y i n s t a l l e d on t h e space coolers, were adequate f o r t h e normal operating pressure of

+ p s i g but

were too small

f o r higher pressures such as could be encountered i n t h e event of an acciThey were, therefore replaced with 10-hp motors.

dent.

One of these on

t h e drain tank c e l l cooler f a i l e d a f t e r 2900 hours of operation due t o t h e r o t o r s l i p p i n g on t h e shaft.

Since t h e trouble app'eared t o be caused by

f a u l t y construction r a t h e r than poor design, it was replaced by an identic a l motor. Before t h e c e l l s were sealed, a reactor c e l l cooler (RCC-2) developed a leak a t a header-to-tube brazed f i t t i n g .

Repairs proved t o be d i f f i c u l t

because,heat required f o r brazing one leak would cause an adjacent tube t o

start leaking.

After t h e c e l l was sealed, a 1-1/2 gallon per day water

leak w a s traced t o another brazed f i t t i n g ,leak on RCC-1. paired.

I n , J u l y 1967, RCC-2 again s t a r t e d leaking and w a s replaced by one

of t h r e e newly obtained un

n which t h e copper tubes were heliarc-welded header.

( i n s t e a d of brazed) t o t h 12.4.5

This was re-

In-Cell Water Leaks

In-cell water leaks have p e r s i s t e d throughout t h e r e a c t o r operation. Before t h e c e l l s were sealed, components could be checked Visually but aft e r power operation, l o c a t i n g t r e a t e d water leaks required hydrostatic t e s t i n g of i n - c e l l components

Although all known leaks such as a t t h e

space coolers were repaired,

continued t o have a leak i n t o t h e c e l l of

1 / 2 t o 1 gallons per day.

kid

Water from t h i s small leak d i d not c o l l e c t i n

t h e c e l l sump but evaporated i n t h e 130 t o 140째F c e l l gas and then condensed i n a colder portion of t h e containment, t h e suction l i n e s and gas

242

cooler of t h e component cooling system. t a i n considerable amounts of tritium.

This condensate w a s found t o con-

Drains were i n s t a l l e d t o permit-

draining t h e condensate t o a measuring tank i n t h e sump room from which it could be pumped t o t h e l i q u i d waste storage. The nuclear instrument penetration was one of t h e suspected sources

of t h e leak.

Since it could not be hydrostatically t e s t e d , 12 kg of D 2 0

was added as a t r a c e r .

This increased t h e deuterium concentration t o about

seven times the normal water concentration.

Since t h e deuterium concentra-

t i o n d i d not increase i n samples of t h e c e l l condensate, a leak f r o m t h e nuclear instrument penetration was discounted. 12.4.6

Treated Water Heat Exchanger

The heat exchanger i n s t a l l e d t o t r a n s f e r t h e heat from t h e treated wa-

t e r t o t h e cooling tower water w a s a surplus tube-and-shell

unit.

During

preoperational tests t h e tubes had t o be r e r o l l e d and tube sheet gasket re-

placed t o eliminate intermixing of t r e a t e d water and cooling tuwer water.

A f t e r a year of operation, 17 tubes which had developed cracks where t h e tubes were r o l l e d i n t o t h e tube sheet were plugged. caused by t h e ' e a r l i e r tube repairs.

This w a s probably

When leaks again developed i n t h e Tw

cooler, more tubes were plugged but when attempts t o s e a l t h e flanged tube sheet were unsuccessful, t h e heat exchanger w a s replaced w i t h another sur-

plus u n i t which w a s smaller but s t i l l had adequate cooling capacity.

Af-

t e r replacing t h e heat exchanger, sodium from t h e cooling tower system had t o be removed f r o m t h e t r e a t e d water by d i l u t i o n with steam condensate and new i n h i b i t o r was added.

The a c t i v a t i o n of t h e 1.6 ppm of sodium l e f t i n

t h e treated water a f t e r t h e d i l u t i o n raised t h e r a d i a t i o n l e v e l of t h e sys-

t e m t o 80 mR/hr at surge tank and 260 mR/hr a t t h e f i l t e r but d i d .not s e r i ously a f f e c t r e a c t o r operation. 12.4.7

Valves

Containment of the t r e a t e d water system depended upon air-operated block valves and check valves. annual containment checks, leaked excessively.

None of t h e block valves leaked during t h e

However, several of t h e piston-type 'check valves

Foreign material and s c a l e caused some of t h e valves

t o bind i n t h e open position due t o t h e close clearance between t h e valve

24 3

body and t h e plug.

The valves were mounted i n horizontal pipes which

probably magnified t h e problem. Pressure r e l i e f valves were i n s t a l l e d t o prevent damage t o in-cell components i n case t h e block valves closed.

waste tank.

These discharged t o t h e l i q u i d

Leakage from these occurred several times.

The first indi-

cation was usually an increase i n r a t e of condensate makeup.

Pinpointing

the l e a k involved disconnecting piping from each r e l i e f valve.

Power re-

duction w a s necessary t o accomplish t h i s .

12.4.8

Corrosion I n h i b i t o r

The t r e a t e d water system w a s first f i l l e d w i t h condensate containing

2000 ppm of a 75+51

mixture of potassium n i t r i t e - p o t a s s i m tetra-borate. Induced a c t i v i t y 42K a t low reactor power w a s extrapolated t o radiat i o n l e v e l s of 400 mR/hr at t h e surge tank i n t h e water room and f i l t e r i n t h e d i e s e l house when t h e r e a c t o r was at f u l l power. To prevent th3s radia t i o n , t h e potassium i n h i b i t o r w a s changed t o lithium.

Lithium-'7 w a s used

t o minimize t h e production of tritium which would be formed i n t h e thermal

shield.. The potassium was removed by d i l u t i o n with demineralized water from ORmL but had t o be repeated using steam condensate because t h e demine r a l i z e d water which contained 1 ppm of sodium became radioactive during t h e next run.

The new i n h i b i t o r w a s composed of lithium n i t r i t e , b o r i c

acid, and.lithium hydroxide. During e a r l y power operation, t h e t r e a t e d water l o s t a c t i v e i n h i b i t o r . Samples indicated t h a t 300500 ppm H202 was being produced by r a d i a t i o n damage which oxidized L i N O

equilibrium of -NO2--NO3 Approximately 365 s these were analyzed at t

iNO3.

Additional LiN02 w a s added u n t i l

ched with Q700 ppm of -NO3 i n t h e water. f t h e t r e a t e d water were taken.

Most of

o assure t h a t adequate treatment was be-

i n g maintained. 12.4-9

Corrosion

Corrosion - i n ,t h e treat i n g most of t h e operation, h i b i t o r , pH and.iron.

r system has not been s i g n i f i c a n t . Dures were analyzed monthly f o r loss of in-

I n 1969, chromium and nickel analyses were added

t o f u r t h e r check f o r corrosion on t h e in-cell s t a i n l e s s s t e e l .

The i r o n

244

analysis has remained near t h e lower detectable l i m i t of t h e analysis indi-

cating no corrosion and t h e C r on N i analysis a l s o indicates no s i g n i f i c a n t corrosion.

Although t h e chromium and nickel analyses haven't been followed

continuously, any corrosion would have caused an increase i n t h e amount present,

The present chromium analysis of .4 ppm i s equivalent t o 0.05

pounds of s t a i n l e s s s t e e l . One f a c t o r which-clouds t h e i n t e r p r e t a t i o n of t h e corrosion results has been t h e continual leakage from t h e system which has required periodic makeup.

The decrease i n concentration of i h i b i t o r of 25% i n two years

agrees reasonably with t h e known leakage.

An increase of t h e i r o n , chrom-

i u m , and nickel by t h e same amount would s t i l l i n d i c a t e a negligible amount of corrosion. If corrosion products p r e c i p i t a t e d as oxides, they would be caught on t h e t r e a t e d water f i l t e r .

This f i l t e r , located i n t h e d i e s e l

house, b u i l t up excessive pressure and w a s cleaned twice during p r e c r i t i c a l t e s t i n g but has not required f u r t h e r cleaning i n t h e past four years. 12 4.10

Recommendations

The biggest problem i n t h e t r e a t e d water system w a s t h e r e s u l t of t h e

l o w design pressure of t h e thermal s h i e l d (20-30 p s i ) .

Care should be ta-

ken i n f u t u r e designs t h a t all components withstand t h e dead-head pressure

of t h e c i r c u l a t i n g pump.

The piping changes required f o r protection of t h e

thermal s h i e l d r e s u l t e d i n other p a r a l l e l flow paths having marginal capacity. Consideration should be given t o t h e design of a system which i s comp l e t e l y contained.

This would eliminate t h e need f o r so many block valves.

Check valves, of t h e type i n s t a l l e d a t t h e MSRE, should not be r e l i e d on as containment block valves.

If similar valves are used, they should be i n

v e r t i c a l piping and have s u f f i c i e n t clearances t o insure t h a t they w i l l close when needed. A corrosion i n h i b i t o r should be s e l e c t e d t o minimize induced a c t i v i t y .

Provisions f o r handling gaseous a c t i v i t y such as tritium and r a d i o l y t i c hydrogen should be provided.

Large components should be shielded o r lo-

cated so as t o minimize personnel exposure t o t h e induced a c t i v i t y .

245

12.5

mdensate has been a very s a t i s f a c t o r y supply of de-

Building steam mineralized water.

QWdensate System

The steam i s condensed i n a s t a i n l e s s s t e e l tube-and-

s h e l l heat exchanger and stored i n two s t a i n l e s s s t e e l tanks. I n i t i a l l y condensate makeup was added by gravity feed through an auto-

matic l e v e l control valve t o maintain a constant l e v e l i n t h e t r e a t e d water To obtain b e t t e r t r e a t e d water inventory data, t h i s was

surge tank.

changed t o periodic manual additions when t h e surge tank l e v e l decreased below 50 gal. Condensate i s used t o f i l l feedwater tanks which supply cooling water f o r r e c i r c u l a t i n g evaporative cooling of t h e f u e l drain tanks. i n h i b i t o r i n t h i s service would p l a t e out during boiling.

A corrosion

The system has

worked well except f o r 2 occasions when leaking valves between t h e feedwater

tanks and steam dome cause

i d e n t a l addition of water t o a steam dome.

A t one time t h e high drain

temperature ( 1 3 O O O F ) interlocks actuated

t h e supply valves t o FD-2 steam dome following a normal drain.

6.i

This could

have been prevented by turning off e l e c t r i c heaters t o t h e drain tank s h o r t l y after t h e drain.

o refluxing, all of t h e water could not be

removed from t h e steam dome when cooling was no longer required.

To cor-

r e c t t h i s , a pump and storage drum were i n s t a l l e d i n t h e north e l e c t r i c service area t o permit complete drainage of t h e steam domes when required. 12.6

Nuclear Instrument Penetration

Instruments f o r l o c a t e d i n a 36-in.-di t o t h e inner edge of t h e about 1700 gallons

tassium n i t r i t e and potass 2000 ppm.

A s with

ithium n i t r i t e - l i t h i u m t e w a s i n i t i a l l y by natur

uclear c h a r a c t e r i s t i c s of t h e system were

a f t which extended f r o m t h e high-bay f l o o r shield.

This was o r i g i n a l l y f i l l e d with

water t o which a

75+5% mixture of

e t r a b o r a t e w a s added t o a concentration of

r system, t h e i n h i b i t o r was changed t o r a t e t o lower t h e induced

ivity.

a t i o n but gamma heating wh

the reactor was

taken t o full power increased t h e water temperature %7O0F.

PO-

Cooling

This decreased

t h e neutron attenuation causing a divergence between t h e indicated nuclear

246 J

power and t h e heat balance power.

A cooling system using a 5-gpm pump and

a small available heat exchanger were i n s t a l l e d which reduced t h e increase

i n water temperature t o

and kept t h e nuclear power measurements and

% 1 8 O ~

heat balance calculations within %5%. I n i t i a l l y t r e a t e d water w a s used f o r makeup t o t h e nuclear instrument penetration.

Since t h e main losses were by evaporation, t h i s caused t h e

i n h i b i t o r concentration t o increase.

Steam condensate w a s then used f o r

water makeup. Approximately 215 samples were t a k e n of t h e nuclear instrument penet r a t i o n water.

Most of these were analyzed at t h e s i t e t o assure t h a t ade-

quate treatment w a s maintained.

As i n t h e treated water system, t h e i r o n

analysis of samples sent t o t h e laboratory remained near t h e lower detectable l i m i t and t h e chromium content never exceeded 0.4 ppm indicating very

l i t t l e corrosion. 12.7

General Water Systems Conclusions

The water systems were q u i t e adequate and as a whole presented very f e w problems.

The biggest problems resulted from t h e marginal pressure

coâ&#x20AC;&#x2122;

r a t i n g of t h e thermal s h i e l d and piping e r r o r t o one of t h e thermal s h j e l d s l i d e s which resulted i n t h e formation of a l a r g e gas pocket.

The piping

changes required t o prevent over-pressurizing t h e TS and additional equipment needed f o r gas s t r i p p i n g weakened t h e performance of t h e t r e a t e d water system. Corrosion i n h i b i t o r s i n both cooling tower water and t r e a t e d water systems proved t o be excellent i n h i b i t o r s but they had undesirable s i d e effects -the

s c a l e and sludge i n t h e CTW system and induced a c t i v i t y and

tritium formation i n t h e TW system. Equipment and instrumentation performance was very good,

The only

equipment troubles were the t r e a t e d w a t e r heat exchanger and brazed tubing i n t h e c e l l space coolers.

The in-cell water leakage (presumably from t h e

space cooler brazed f i t t i n g s ) , w a s annoying, but afâ&#x20AC;&#x2122;ter i n s t a l l a t i o n of t h e condensate drainage system, did not i n t e r f e r e w i t h performance of the reactor.

247

13.

LIQUID WASTE SYSTEM P. H. Harley

13.1 Description The l i q u i d waste system consisted of equipment which handled t h e aqueous wastes from t h e area. Non-hazardous wasts were pumped t o small catch bas$ns and then t o a drainage f i e l d west of t h e area.

This water was from

building foundation drains and various f l o o r drains i n t h e building, over-

SLOW from t h e charcoal bed c e l l , etc.

Hazardous wastes, primarily ,the

leakage i n t o t h e reactor c e l l s , and auxiliary c e l l s were pumped o r j e t t e d t o a waste storage tank f o r sampling and treatment and then transferred t o t h e Melton Valley Waste Disposal System. The waste system design included a component decontamination f a c i l i t y

and a water-shielded c e l l i n which maintenance of radioactive components could be performed.

Some of t h e equipment w a s i n s t a l l e d but as an economy

measure, t h i s section of t h e system was not completed.

13.2

(bi

Preliminary Testinq

Piping and tanks were leak-tested while running performance t e s t s on t h e pumps and j e t e j e c t o r s used t o t r a n s f e r liquids.

The l i q u i d waste

storage tank and c e l l sumps which have bubbler-type l e v e l indicators were calibrated and t h e e j e c t o r s were then t e s t e d t o remove t h e water.

No d f f f i c u l t i e s w e r e encountered w i t h t h e steam-operated ejectors i n t h e auxiliary c e l l s but t h e capa of t h e air-power j e t s i n t h e reactor and drain-tank c e l l was excee formance of t h e e j e c t o r

low. Attempts were made t o improve the ging t h e venturi diameter but no sig-

cant improvement was

pressure of t h e long discharge l i n e

limited t h e j e t capacity. but t o obtain more capacit

l i n e s a t t h e Denetration use e i t h e r a i r o r steam j e t s using a i r w a s 25+3

e out o f t h e c e l l , a i

if'ying piping i n t h e waste c e l l helped some

steam supply w a s piped t o the j e t supply -tank c e l l .

Provisions were made t o

) f o r t h e j e t supply.

and using steam 8.5 d t h e preferred supply.

Although t h e decontamination f a c i l i t y was not completed, t h e waste f i l t e r f o r clarif'ying shielding water w a s i n s t a l l e d and tested.

Filtering

248

required pumping 30 g p m through t h e f i l t e r using t h e waste pump but backflushing t h e f i l t e r required %140 gpm.

Process water was used f o r t h i s

operation and t h e flow was c a l i b r a t e d versus t h e process water pressure required (15 p s i = 140 gpm)

I n i t i a l t e s t i n g of t h e waste pump f o r c i r c u l a t i n g (mixing) waste tank contents and t r a n s f e r r i n g w a s t e s t e d s a t i s f a c t o r i l y but t h e pump had a tendency t o gas bind a t low tank l e v e l s and process water was required f o r priming These i n i t i a l t e s t s w e r e used t o check and improve operating procedures as well as t e s t t h e equipment.

13.3

Operating Experience

The primary aqueous a c t i v i t y found i n t h e MSRE waste has been 3H. A * continuing water leak i n t o t h e c e l l since e a r l y power operation has been drained i n t o a condensate c o l l e c t i o n tank and t r a n s f e r r e d t o t h e waste tank a t l e a s t once a week. t h e c e l l sump.

The water from t h e i n - c e l l leak did not c o l l e c t i n

The water evaporated i n t h e 135'F

gas i n t h e c e l l and then

condensed out a f t e r being compressed t o 6 p s i g from +2 p s i g and cooled t o

%11O0F i n t h e component cooling system gas cooler. up t o 6 curies/gal.

This water contained

Most of t h e water i n t h e waste tank was from r a i n wa-

t e r which drained from t h e v e n t i l a t i o n stack and f i l t e r e d i n t o a c o l l e c t i o n tank and w a s pumped t o t h e waste tank.

Although t h i s water could have been

contaminated from material on t h e f i l t e r s , very l i t t l e a c t i v i t y was detected.

Other sources of contaminated water i n t h e l i q u i d waste were from

t h e c a u s t i c scrubber during f u e l processing and wash water from cleaning t h e hi-bay f l o o r and a u x i l i a r y c e l l s following r e a c t o r maintenance periods. Table 13.1 shows t h e calculated a c t i v i t y of material t r a n s f e r r e d t o t h e Melton Valley Waste Disposal System.

The 3H a c t i v i t y c o r r e l a t e s t o

reactor pawer f a i r l y w e l l as being one c u r i e of 3H f o r every 45 MWh of power operation. The equipment i n t h e l i q u i d waste system worked q u i t e well.

Some

d i f f i c u l t i e s were encountered i n keeping t h e building sump l e v e l switches

See Section 12 on Water System.

249

Table 13.1 Liquid Waste Tank Transfers And Total Contaminants

Date

C Tritium

Curies

---

C F P

Vol

Curies

Gal.

0.092

5030

141

1800

8850

457 868

5060

_--

7000

0.031

5020

0.005

7440

0.0002

6500 6400

aligned.

These were operated by f l o a t s w i t h %5 f e e t v e r t i c a l tube which

actuates an ON-OFF switch.

Misalignment usually caused t h e f l o a t t o "hang"

and t h e pump would f a i l t o shut o f f on low water level. On several occasions, t h e water l e v e l i n t h e charcoal bed c e l l w a s lowered so t h e i n l e t s t o t h e beds could be heated.

This was accomplished

by draining % l 5 O O gal. of water i n t o t h e sump and allowing t h e two sump pumps t o pump t h e water t o t h e catch basin. During t h e r e a c t o r construction, t h e r e w a s only one building sump pump and it failed causing t h e coolant drain c e l l t o flood through a f l o o r drain. (The CD c e l l and sump room f l o o r are at t h e same elevation, 820 ft.) A s m a l l sump was cut i n t o t h e coolant drain c e l l f l o o r and a float-operated

alarm and steam j e t were i n s t a l l e d . failure of both sump pumps

The system was maintained t o warn of

d provide the steam j e t t o keep t h e sump l e v e l

down during a possible e l e c t r i c a l power f a i l u r e . A second sump pump was ordered t o replace t h e u n i t which was l e f t from

earlier building operations.

Before receiving t h e new pump, t h e o r i g i n a l

250

one was overhauled and t h e new pump was i n s t a l l e d t o provide an operating spare.

The tritium t r a n s f e r r e d through t h i s system w a s %lo% of t h e dis-

charge from ORNL and no effo& w a s made t o l i m i t t h e r a t e of discharge. However, f o r a l a r g e breeder r e a c t o r where l a r g e r q u a n t i t i e s of 3H would be involved, some control of t h e tritium r e l e a s e would be required. If air-operated sump j e t s axe required i n other r e a c t o r s , care should

be used t o insure t h a t t h e e j e c t o r s have s u f f i c i e n t capacity.

251

14.

HELIUM LEAK DETECTOR SYSTEM

J. K. Franzreb

14.1

Description and Method of Operation

A leak detector system was used .to monitor a l l flanges i n t h e MSRE , system which could.permit.the-escape of radioactive materials.

I n addition,

a l l flanges which had t o be maintained by remotely-operated t o o l i n g were provided with leak-detected type j o i n t s t o serve as an indication of s a t i s factory reassembly.

The 119 leak-detected flanges were pressurized with

helium by 65 l i n e s from 8 headers.

As shown on Fig. 14-1, some leak-

detector l i n e s were used f o r more than one flange.

The principles of con-

s t r u c t i o n of a t y p i c a l leak-detected flange closure i s shown on Fig. 14-2. I n normal operation, t h e e n t i r e system w a s interconnected and t h e helium supply was valved off.

Periodic repressurization w a s necessary t o

keep t h e pressure within the l i m i t s of 90 t o 100 psig. leak, helium flowed i n t o t h

I n t h e event of a

affected systems and t h e r e s u l t a n t l o s s i n

pressure i n t h e leak detector headers actuated an alarm i n t h e control room. If t h e records indicated t h a t t h e r a t e of pressure decrease was greater

than t h e 0.66-psi/hr

a r b i t r a r y l i m i t , t h e headers were i s o l a t e d t o narrow

down t h e location of t h e leaking flange.

The leak-detector l i n e s were then

valved o f f individually or i n groups u n t i l t h e location of t h e leak was de-

A reference volume and a d i f f e r e n t i a l pressure t r a n s m i t t e r was provided to aid i n i s o l a t i n g leaks and for determining leak rates. The altermined.

lowable leak r a t e from each flange o r group of flanges was set a t l x

The volumes i n

were determined by pressuriz-

ious p a r t s

unknown q u a n t i t i e s of gas with

an accurately

calculated by using t h e calculated using t sections of t h e system.

unknown system volumes were

Cali

given i n Table 14-1.

1 ~ + 1 ~

=~P3(v1 2

~ 2 ) . Leak

rates were

of pressure decay from these The volumes of various p a r t s of t h e system a r e

252

ANY LINE REQUIRING LEAK DETECTOR SERVICE COMPONENT REMOTE

OISCONNECT FLANGE PERMANENTLY INSTALLED PIPE

LEAK DETECTOR LINE

I ,

Fig. 14.1

FROM LEAK -C-~OETECTOR

STAT ION

Method of U t i l i z i n g One Leak Detector Line t o Serve Two Flanges i n Series

ORNL DWG 64-8833

cn w

\ t Fig. 14.2

Schematic Diagram of Leak-Detected Flange Closure

254 Table 1 4 , l

Volumes i n Leak Detector System

Volume of Line 400 = 184 cc Volume of High Pressure s i d e of DP c e l l = 730 cc Volume of Header Line Line Line Line Line Line Line Line Line Line

401 = 156 cc 410 = 244 411 = 216 412 = 224 413 = 144

414 415 416 417 418 419

= OPEN = OPEN

88 OPEN OPEN

1 4 1 cc

Header 403 = Line 430 = Line 431 = Line 432 = Line 433 = Line 434 = Line 435 = Line 436 = Line 437 = Line 438 = Line 439 = Header 404 Line-440 = Line 441'= Line 442 = Line 4 4 3 = Line 444. = Line 445 = Line 446 = Line 447 = Line 448 = Line 449 =

Line

NOTE:

Line Line . Line

149 cc

117 117

108 106

66 63

- 143

Header Line Line Line Line Line Line

130

421 = OPEN 422 = 111 4 2 3 ' ~125 424 = 120 425 = 115 426 = 115 427 = 115 428 = 138 429 = 135

123 122 122

195 193 189 178 122 108

405 = 144 450 = OPEN

451 = 113 452 = 108 453 = 106

454 = 455 = 456 = 457 = Line 458-= Line 459 =

= 65

= = = Header 402 = Line 420 = Line Line Line Line Line Line Line Line Line

Volume of Header Line Line Line Line Line Line Line Line

Header Line Line Line Line Line Line Line Line Line Line

406 = 460 =

461 = 462 = 463 =

464 465 466 467 468 469

109 104 109 102

99 98 148 154 99 148 99

= 107 = 108 = OPEN = OPEN

= OPEN = OPEN -

407'= 150 470 = OPEN 471 = 291 472 = 235

473 = 474 = 475 = 476 = 477 = 478 = 479 =

518 143 135 138 109 305

466 Header 408 = 147 Line 480 = 144 Line 481 = 159 Line 482 = . Line 483 = Line 484 = Line 485 = Line 486 =

123 123 310 210 SPARE L\ne 487 = SPARE Llne 488 = SPARE Line 489 = SPARE

The "B" valves i n t h e leak detector l i n e s remained open during all tests.

255

14.3

Operating Experience

During construction and pre-operating leak t e s t i n g , some d i f f i c u l t y was encountered i n a t t a i n i n g acceptable leak rates on some flanges, part i c u l a r l y t h e very l a r g e s a l t f'reeze flanges.

During t h i s and subsequent

maintenance periods, t h e leak detector system proved very valuable i n checking t h e progress and degree of t i g h t n e s s of a l l flanges. Operating experience showed t h a t once a flange was made t i g h t , it stayed t i g h t during service.

During t h e e n t i r e l i f e of t h e MSRE, no leak

ever developed t h a t was serious enough t o cause a shutdown of t h e reactor. Some problem was posed by a f'reeze flange (FT-201) located near t h e point where t h e coolant s a l t l e f t t h e heat exchanger i n t h e r e a c t o r c e l l .

It leaked more than t h e maximum allowable r a t e of 1 x

lom3 cc/sec

was not i n t h e piping and t h e coolant system had been cooled down.

when salt This

leak w a s measured a t about 0.2 cc/sec on Augugt 2 , 1966 while t h e system

was drained and cold.

The flange always became acceptably sealed when t h e

system w a s heated p r i o r t o

t i n g molten s a l t i n t o t h e piping and remained

so when t h e r e a c t o r was at power.

The s e a l i n g action was probably due t o

the thermal expansion of t h

t e r n a l s of t h e flange being g r e a t e r than

t h a t of t h e e x t e r n a l clamp.

This would cause an increase i n bearing pres-

sure on t h e s e a t i n g surfaces of t h e metal "0" ring. During normal operation w i t h t h e e n t i r e system interconnected, t h e pressure decreased at about rvals of one t o t

s i per day, thus requiring repressurizing n a 52-day p e r i ystem operation

e r a t u r e i n Octobe

e r 1968, t h e p r

decay rate w a s

I n a 60-d

power operat i o

ptember and Oc-

i/day.

tober 1969, t h e decay r

0.97 psi/day.

November 30, 1968, t h e 1

e did increase t o 2.1

thermal expansion combined

During

system cooldown on sifhr

mal pressure decay p

Differential

l y caused t h e

high indicated leak r When an abnormal t o perform

entire s

then each l i n e of t h e

4 t o 8 hours

red, it would onsisted of i s o l a t i n

h header and

t o be leaking.

ng t h e leak

r a t e of an individual l i n e took about an hour o r two, hcyever, a rough check t o guide maintenance personnel could be obtained i n about 10 minutes.

256

The individual l i n e s , which monitored more than one flange, gave an i n t e r p r e t a t i o n problem.

It w a s not possible t o t e l l at t h e outset i f all

t h e leak w a s through one flange o r apportioned between multiple flanges on t h e same l i n e .

When t h i s s i t u a t i o n arose, t h e p r a c t i c e became t h a t of al-

t e r n a t e l y tightening one flange, then t h e other, incrementally, and running a combined leak rate on t h e l i n e between these operations.

By observing

when improvement w a s made, the apportionment of t h e leak between flanges could be estimated. It w a s necessary t o exceed t h e maximum recommended torque i n some in-

stances.

The technique was used t o overtorque where necessary t o seat t h e

sealing faces, then t o loosen- t h e b o l t s and retorque them t o t h e nominal o r , i f necessary, t o t h e maximum value l i s t e d i n Table 14.2. ceptional cases, it was necesiary t o exceed t h i s maximum.

I n a few ex-

Some of t h e

m e t a l O-rings had t o be replaced w i t h gold-plated O-rings i n order t o ob-

t a i n acceptable leak rates.

Some o f t h e 1/2-in. b o l t s i n small flanges

were torqued as high as 150 ft-lbs and operated at about 80 ft-lbs.

The

maximum recommended torque was 70 ft-lbs.

LfiJ

The record of measurements of system and flange leak rates w a s kept

i n a "Leak Detector Log Book" at t h e LKD control cabinet.

Flange leak

rates, plus a record of all work done involving any of these flanges, was logged i n the "MSRE Flange Log" notebook. each flange on a separate page. t e c t o r Valve No.,

This l a t t e r l o g gave d e t a i l s on

Information given w a s Flange No.

, leak

de-

s i z e of b o l t s ( o r i n t h e case of freeze flanges, t h e

force exerted by t h e removable c l i p i n t o n s ) , r i n g s i z e , t h e maximum permissible and recommended torque i n ft-lbs.,

t h e location of t h e leak de-

t e c t e d j o i n t , plus t h e date of action, t h e reason f o r t h e a c t i o n , a c t u a l torque applied, and any general comments.

14.4 Discussion and Recommendations The leak detector system functioned s a t i s f a c t o r i l y as designed.

Due

t o t h e two s e a l i n g surfaces on t h e "0" r i n g s , primary containment w a s maintained as long as t h e leak detector system overpressure w a s g r e a t e r than any possible pressure that could develop i n t h e primary system.

Therefore,

leak r a t e l i m i t a t i o n s were s e t based on convenience i n maintaining an adequate overpressure.

257

Table 14.2

Bolt Diameter

1-112

Threads Per inch

MSRE Flange Bolt Torque Chart

Nominal Torque ( f t - l b .

Maximum Torque ( ft-lb. )

135

150 240

225

533

800 1200

9 7 6

900

360

Wrench Size

314" 718" 1-1/ 8" 1--5/16*~ 1-112" 2- 318"

Tightening Procedure:

1. Torque b o l t s t o nominal torque value following a c r i s s c r o s s sequence. Approach f i n a l value i n successive increments of 10 t o 20 f t - l b . 2. If leak e x i s t , torque b o l t s t o maximum torque value, back o f f and retorque t o nominal value per s t e p 1.

3. Repea't s t e p 2 as required, increasing nominal value by 10% each successive s t e p , NEVER EXCEED MAXIMUM VALUES without permission of t h e MSRE Operations o r Maintenance Chief.

258

U Remote welding o r brazing techniques should be used where maintenance i s infrequent.

This should be more economical and would reduce t h e number

of d e t a i l s t o be handled by t h e operating personnel. When opening primary system flanges, it i s desirable t o have t h e leak detector vented t o atmospheric pressure t o minimize t h e spread of a c t i v i t y . Each header should have a l i n e which i s connected t o t h e c e l l t o f a c i l i t a t e doing t h i s since t h e gas i n t h e l i n e could be contaminated. ,

259

15.

INSTRUMENT AIR SYSTEM T. L. Hudson

15.1 Description The instrument air system supplied clean, dry, compressed air f o r pneumatic instruments and other s p e c i a l uses.

Two Joy compressors, AC-1

and AC-2, were used t o compress t h e a i r , which then passed through an a f t e r cooler and entrainment separator t o a common l i n e supplying two p a r a l l e l receiving tanks.

From t h e receiving tanks, t h e a i r passed through one of

two p a r a l l e l drying s t a t i o n s containing T r i n i t y h e a t i e s s dryers.

The dry a i r was d i s t r i b u t e d through headers t o l o c a l l y mounted f i l t e r and reducing

stations.

The service a i r compressor, AC-3,

could be valved i n t o t h e in-

strument a i r system upstream of t h e receiver tanks i f required. 15.2

Operating Experience

The capacity of t h e instrument a i r system w i t h one a i r compressor i n operation w a s adequate at all times. O f t h i s amount,

Normal a i r usage was about 70 scfm.

18 scfm was used f o r purging t h e instrument a i r dryers,

12 scfm f o r supplying n o n c r i t i c a l instruments, and 40 scfm f o r supplying t h e c r i t i c a l instruments.

The c r i t i c a l instruments were supplied by t h e

emergency instrument a i r supply, which consisted of two banks of s i x nitrogen cylinders, when t h e normal instrument a i r supply w a s l o s t . The capacity of t h e emergency instrument a i r supply w a s t e s t e d during

normal operation with f l u s h s a l t t o assure an orderly shutdown of t h e rea c t o r i n case of loss of both instrument a i r compressors (see MSRE Test Report 2.3.11.3).

With an emergency a i r usage of 40 scfm and one bank of

cylinders at t h e i r lowest operating, 1500 p s i g ( i n i t i a l low-pressure alarm point) and t h e other bank fW.1 (52000 p s i g ) a t t h e time both a i r compressors stop would.provi.de t h e following times: (1) A i r i n t h e compress

surge tank would l a s t 5 4 min.

( 2 ) F i r s t bank of nitrogen cylinder ( i f used t o TOO p s i g - a t

this

pressure t h e control valve stopped controlling) would l a s t 513 min,

( 3 ) Subsequent f u l l banks of nitrogen cylinder ( i f used t o 700 p s i g ) would l a s t 520 min.

260

From t h e test it w a s concluded t h a t t h e emergency instrument a i r system

was adequate.

It would have been necessary t o i n s t a l l a l a r g e r control

valve t o more f u l l y u t i l i z e t h e nitrogen cylinders. Leaks i n t h e emergency nitrogen headers were repaired several times and t h e control valve w a s repaired once.

On t h e average one bank of nitro-

gen w a s changed out every two weeks t o keep t h e cylinders above t h e i n i t i a l

alarm pressure of 1500 psig.

15.3, Conclusions The operation of t h e normal instrument a i r system was very satisfactory.

A t no time w a s it necessary t o use t h e emergency nitrogen system

during operation of t h e reactor.

One instrument air compressor was oper-

ated with t h e other i n standby except f o r during r e p a i r s o r maintenance. Even w i t h a complete power outage, it was possible t o s t & t one air corn-

pressor on diesel power before t h e receiver tank pressure dropped enough t o cause a t r a n s f e r t o the emergency system.

Although some extensive main-

tenance was done on t h e a i r compressors, only one was ever inoperative at a time.

261

16.

ELECTRICAL SYSTEM-*

T. L. Hudson

16.1 Description .

The MSRE e l e c t r i c a l power was supplied from t h e ORNL substation by e i t h e r of two 13.8-kV TVA l i n e s , a preferred l i n e o r an a l t e r n a t e .

These

were separated by two interlocked. pole-mounted motbr-operated l i n e switches. .-Automatic t r a n s f e r w a s porvided from t h e preferred l i n e t o t h e a l t e r n a t e Both t r a n s f e r switches could be manually operated remotely from t h e

line.

a u x i l i a r y control room.

The ac power entered the MSRE building through two transformer subA 1500-kVa, 480-V, %phase substation served t h e process equip-

stations.

ment, and a 750-kVa, 480-V, >phase a u x i l i a r y s t a t i o n w a s i n i t i a l l y f o r building services, such as l i g h t i n g , v e n t i l a t i o n , e t c .

After t h e process

substation became overloaded, bus 5 which supplies e l e c t r i c a l heaters w a s connected t o t h e a u x i l i a r y system.

Three diesel-generators were i n s t a l l e d

f o r emergency ac use. There were two separate area dc systems, a 48-v system and a 250-V system.

These were normally operated by ac-dc motor-generator sets and

had b a t t e r y supplies f o r emergency use. The 48-v dc system provided power f o r e l e c t r i c a l controls and one channel of t h e r e a c t o r s a f e t y systems.

The system c o n s i s t s of two 3-kW motor-generator sets (each r a t e d at 53 ampere at 56 V ) and a bank of 48-V b a t t e r i e s (600 amp-hr).

The two generators normally supplied t h e dc power

and charged t h e b a t t e r i e s .

I n an emergency, when no ac power was available

t o run t h e motor-generators

8-V dc power was supplied d i r e c t l y from t h e

batteries.

The 250-V dc system provided power t o t h e r e l i a b l e power system, process breakers t r i p power, 13.8-kV t r a n s f e r control power, and emergency lighthg.

The system consisted

f one 125-kW motor-generator s e t and a

bank of 240-V b a t t e r i e s (364 amp-hr). The r e l i a b l e power sys

provided power t o c e r t a i n important process

equipment and one channel of t h e s a f e t y systems.

I n i t i a l l y , t h e power was

supplied from a 25-kW dc-ac motor-generator s e t .

The dc motor w a s supplied

262

from t h e 250-V dc system and t h e generator supplied t h e 120/240-V ac rel i a b l e power system.

The operation of t h e MG s e t was unreliable because

of failures i n e l e c t r i c a l control components and was replaced w i t h a s t a t i c i n v e r t e r (see 16.4). Thk MSRE e l e c t r i c a l i n s t a l l a t i o n made extensive use of t h e e l e c t r i c a l

f a c i l i t i e s t h a t were i n s t a l l e d f o r t h e ARE and ART operations i n t h e 7503

area.

Many modifications-were made t o t h e e x i s t i n g equipment, but only a

l i m i t e d ambunt of supplemental equipment was required. 16.2 .

Alternating Current System

The operation of t h e ac e l e c t r i c a l system w a s very s a t i s f a c t o r y .

There

were eleven important failures, t e n of which caused unscheduled interrupt i o n s - i n t h e power operation of t h e reactor:. f i v e were-caused by e l e c t r i c a l storms, one by a cable f a i l u r e at a component-cooling pump, one by an overload of t h e main process-power breaker, one by an arc between a 13.8-kV l i n e and an a c t i v a t o r rod t o t h e l i n e fuse, one by a f a i l u r e of t h e main transformer primary fuse, one by a f a i l u r e of t h e a u x i l i a r y transformer primary fuse, and one by a failure o f t h e drain-tank space cooler motor.

These

interruptions varied i n - l e n g t h from a f e w minutes t o s e v e r a l days.

L.i

In a l l

cases t h e necessary r e p a i r s o r modifications were made, and no damage t o reactor equipment was incurred.

These a r e discussed i n more d e t a i l below.

On two occasions while t h e r e a c t o r w a s operating at power, momentary e l e c t r i c a l outages during storms caused control-rod scrams by rangeswitching t h e nuclear power s a f e t y channels because of dips i n t h e fuelpump-motor current.

I n both these cases t h e nuclear power w a s quickly re-

stored t o t h e value t h a t e x i s t e d j u s t p r i o r t o t h e outage.

Since t h e

s a f e t y analysis indicated t h a t l o w range of t h e s a f e t y channels w a s needed only while f i l l i n g t h e r e a c t o r (when t h e f u e l pump was o f f ) , r a p i d response of t h e range-switching was not required.

Therefore, time-delay r e l a y s were

subsequently incorporated i n these c i r c u i t s t o prevent t h e i r a c t i v a t i o n on momentary power dips. Another storm-induced f a i l u r e occurred when l i g h t n i n g s t r u c k t h e main

ORNL power substation and parted one wire of t h e normal MSRE feeder. I n t h i s case t h e e l e c t r i c a l load was automatically t r a n s f e r r e d t o an a l t e r n a t e

263 feeder, and all e s s e n t i a l equipment was r e s t a r t e d i n time- t o prevent draining either t h e f u e l o r coolant system. However, operation at high nuclear power could not be resumed u n t i l service was restored on t h e main feeder. The e n t i r e 13.8-kV supply system was subsequently reviewed and improved t o make it l e s s susceptible t o damage by storms. During a thunderstorm i n June 1967, t h e reactor operation was i n t e r rupted by t h e loss of both feeders.

The r e a c t o r period s a f e t y system func-

tioned properly and scrammed t h e r e a c t o r but two amplifiers were damaged. Approximately 32 hours l a t e r , t h e reactor was returned t o c r i t i c a l operation a f t e r t h e period s a f e t y amplifiers had been repaired.

I n J u l y 1967, an interruption w a s caused by t h e loss of t h e preferred feeder during another thunderstorm.

Emergency power from t h e d i e s e l gener-

a t o r s was i n service within 2 min, and low-power nuclear operation was resumed i n about 13 min.

After r e p a i r s had been completed on t h e preferred

feeder, full-power operation was resumed i n approximately 6 hr. The short i n the component-cooling pump cable s e a l i s described i n Sect. 10 e n t i t l e d "Component Cooling System."

When t h i s short occurred,

t h e r e was a massive flow of current and t h e breaker supplying t h e e n t i r e bus t r i p p e d before t h e individual breaker f o r t h e motor. During t h e i n i t i a l full-power operation of t h e r e a c t o r , t h e main process-power breaker (breaker R) was loaded t o near i t s s e t p o i n t value. When the load on t h e coolant-radiator main-blower motors was increased by increasing the p i t c h on t h e blower fan blades, the.increased load on breaker R caused it t o t r i p

overcurrent.

This condition was relieved

by t r a n s f e r r i n g bus 5 (%200

) from t h e main process transformer t o t h e

a u x i l i a r y power transformer.

The load transfer decreased t h e current

through breaker R *om

about 1700 -/phase

A building power failur when an a r c developed betwee

t o about 1550.

ccurred on a foggy morning i n October 1967, 13.8-kV l i n e t o t h e main transformer f o r

t h e building and p a r a l l e l metal a c t i v a t o r rod t o t h e 13.8-kV l i n e fuse. ined.

was not unusually wet.

ult caused the, operation of t h e preferred

feeder overcurrent ground

Although t h e weather was foggy, it

The cause o f t h e a r c i s un

located a t t h e ORNL substation, which

t r i p p e d t h e feeder breaker before t h e l i n e overcurrent r e l a y located at t h e MSRE could operate t o prevent an automatic t r a n s f e r t o t h e a l t e r n a t e feeder.

264

When t h e MSRE was t r a n s f e r r e d , t h e a l t e r n a t e feeder overcurrent .ground re. l a y operated and t h e a l t e r n a t e feeder breaker a l s o tripped, (To prevent

an automatic t r a n s f e r t o t h e a l t e r n a t e feeder on a similar f a u l t , an overcurrent ground r e l a y was later i n s t a l l e d at t h e MSRE.)

A f t e r an interrup-

t i o n of 37 min, low-power nuclear operation was resumed on emergency elect r i c a l power from t h e d i e s e l generators.

The-damage w a s repaired, t h e

spacing between t h e l i n e and t h e a c t i v a t o r rod wa$ increased, and normal service w a s restored after an i n t e r r u p t i o n of 2 h r . I n April 1969, t h e r e was a failure of a fuse on t h e primary side of t h e MSFtE main power transformer.

This r e s u l t e d i n t h e loss of voltage on

one of t h e transformer windings.

The power supply was changed from a three-

phase t o a single-phase supply.

This caused a drop i n secondary voltage

across t w o windings and i n c r e a s e - i n voltage across t h e other winding.

The

motor-operated equipment continued .to operate u n t i l t h e individual breakers were t r i p p e d by t h e high current i n two of the l i n e s ,

A l l breakers did not

t r i p at t h e same time which made it d i f f i c u l t t o l o c a t e t h e trouble immediately.

The reactor w a s scrammed from f u l l power and t h e d i e s e l gener-

a t o r s were s t a r t e d , but both the fuel and coolant systems were drained .*

before cooling air could be restored t o t h e freeze valves.

No apparent

cause f o r t h e fuse failure w a s found other than possible overheating from poor e l e c t r i c a l contacts. About t h e middle of February 1966, a ground developed i n t h e drain-tank c e l l space cooler 10-hp motor af'ter abaut 2900 h r of operation,

An iso-

l a t i o n transformer w a s i n s t a l l e d t o i s o l a t e t h e ground from t h e e l e c t r i c a l A f t e r a f e w hours of operation, t h e motor stopped and couldn't be

system.

restarted. drained.

The r e a c t o r was taken s u b c r i t i c a l and t h e f u e l system w a s The drain-tank c e l l was opened t o provide c e l l cooling.

The

cooler and motor were removed, and examination showed that t h e r o t o r had slipped along t h e shaf't u n t i l t h e r o t o r fan blades rubbed, causing t h e \

s t a t o r windings t o be destroyed.

The f a i l u r e appeared t o be due t o im-

proper manufacture and not a design weakness.

Therefore, t h e damaged motor

was replaced w i t h one p r a c t i c a l l y i d e n t i c a l . On one occasion t h e a u x i l i a r y power systemwas out of s e r v i c e f o r ab9ut 2 h r s due t o a failure of t h e a u x i l i a r y power transformer primary fuse.

During t h e fuse replacement t h e operation o f , t h e r e a c t o r continued

265

at power with bus 5 supplied from diesel-generator 5.

The fuse f a i l u r e

apparently w a s caused by overheating from poor e l e c t r i c a l contact.

Due t o

low r e s i s t a n c e readings , t h e t h r e e a u x i l i a r y power t r a n s f e r s were replaced i n 1968.

16.3

250-V dc System

The operation of t h i s system has been very s a t i s f a c t o r y .

There have

been some tube f a i l u r e s i n t h e automatic voltage control system f o r t h e generator, but each time t h e u n i t was operated on manual control u n t i l t h e tubes i n t h e automatic controls could be replaced. Once a month t h e charging voltage w a s increased t o 275 V f o r 24 h r s t o give t h e b a t t e r i e s an equalizing charge. The b a t t e r i e s were given a 3-hr l i f e t e s t under full load.

tage dropped t o 215 v o l t s over t h e three-hour period.

The vol-

Extrapolating t h e

voltage curve indicated t h a t t h e lower l i m i t of 210-V would have been reached i n 'L3.5 hours which i s well above t h e 2-hr l i f e upon which t h e design was based.

16.4 Reliable Power System The 120-V ac r e l i a b l e power system was o r i g i n a l l y supplied by an MG set.

I n e a r l y operation, t h i s MG s e t ( o r i g i n a l l y i n s t a l l e d f o r t h e ARE)

proved very unreliable due t o failures i n t h e e l e c t r i c a l c o n t r o l components. To improve t h e r e l i a b i l i t y and

w a s replaced with a new 6 on t h e r e l i a b l e power sup some three-phase load, was

ncrease t h e capacity, t h e motor generator three-phase s t a t i c i n v e r t e r .

The t o t a l load

including t h e on-line computer, which has Va.

The s t a t i c i n v e r t e r offered a number

of advantages over t h e old r

i n v e r t e r , including higher e f f i c i e n c y

(increased 35$), smaller s i z

moving p a r t s , q u i e t operation, and ex-

c e l l e n t voltage and frequenc t h i s equipment, both t h e c

ation.

P r i o r t o t h e i n s t a l l a t i o n of

d voltage regulation of t h e r e l i a b l e

power supply were inadequate f o r t h e operation of t h e on-line computer. During t h e checkout and t e s t i n g o f t h e s t a t i c i n v e r t e r i n April 1966 a f t e r t h e i n s t a l l a t i o n had been completed, trouble occasionally developed

266

t h a t blew t h e load fuses.

Two times t h e manufacturer's f i e l d engineer made

exhaustive t e s t s and could not f i n d any trouble, but a l l symptoms indicated t h a t t h e trouble was i n t h e low-voltage l o g i c power supply, which supplies

24-V dc control power,

On April 25, t h e i n v e r t e r failed again while the

reactor w a s operating at l o w power.

Apparently t h e t r a n s f e r caused a tran-

- s i e n t on t h e ac system, thus a l s o dropping out t h e rod scram r e l a y supplied from t h e ac system, causing a control rod scram.

The i n v e r t e r load was

automatically t r a n s f e r r e d t o t h e normal ac power supply.

After t h i s f a i l u r e

a new power-supply module w a s i n s t a l l e d . The i n v e r t e r output voltage regulation w a s 208 f 1 / 2 V, and t h e output frequency b e t t e r than 60 cps f 0.01%.

The i n v e r t e r was tested f o r 2-1/4 h r

operating from t h e 250-V b a t t e r y system without t h e dc generator operating. Although t h e input t o t h e i n v e r t e r varied 20 V, t h e output varied only

0.5 V. There have been t h r e e other failures associated with t h e s t a t i c inv e r t e r t h a t have automatically t r a n s f e r r e d t h e normal i n v e r t e r load t o TVA The first was caused by failure of load fuses which r e s u l t e d i n an

supply.

unscheduled control-rod scram.

The second w a s caused by a f a i l u r e of a

t h y r i s t e r i n the inverter logic circuit.

The t h i r d w a s caused by a momen-

t a r y ground during maintenance on t h e computer. When a t r a n s f e r occurred, many alarms were received i n t h e control room.

Some were caused by l o s s of operating equipment, and others were These

due t o t h e momentary l o s s of control voltage t o c e r t a i n monitors. had t o be r e s e t t o stop t h e alarm and control action.

After t h e second

i n v e r t e r f a i l u r e , all operating equipment was r e s t a r t e d , but an emergency f u e l drain occurred before t h e reactor-cell a i r a c t i v i t y monitors were reset.

No drain o r other serious condition r e s u l t e d from t h e other i n v e r t e r

failures

16.5 48-V dc System The

48-V dc system w4s very r e l i a b l e .

The only maintenance on t h e

system w a s caused by excessive arcing of t h e generator brushes.

Initially

one generator was operated near r a t e d capacity t o supply t h e load,

Filters

and viewing windows were i n s t a l l e d on t h e generators and both generators

267

were operated i n p a r a l l e l t o reduce t h e individual generator current a f t e r t h e load s l i g h t l y exceeded t h e capacity of one generator.

These changes

g r e a t l y reduced t h e generator brush problems Once a month, t h e charging voltage was increased t o 54 v o l t s f o r ~8 h r t o give t h e b a t t e r i e s an equalizing charge. The b a t t e r i e s were given a 6-hr l i f e t e s t .

47.5 amps and t h e l a s t 2.5

h r at 19.5 amps.

The first 3.5 h r was at

The b a t t e r i e s expended 215

amp-hour of t h e 600 amp-hour l i f e with e s s e n t i a l l y no voltage drop.

16.6 Diesel Generators The t h r e e diesel-power generators, each with a capacity of 300 kW, supplied emergency ac power t o motors and heaters i n t h e MSRE. t o be q u i t e r e l i a b l e .

They were s t a r t e d and operated unloaded f o r about an

Once a month they were t e s t e d under load.

hour each week.

They proved

f a i l e d t o start when required during a power outage.

They never

Under emergency con-

d i t i o n s , they were usually running within 2 min a f t e r a f a i l u r e of t h e norm a l power supply.

I n February 1966, a crack w a s found i n t h e block of D E 3 a t one of Attempts a t r e p a i r s were not successful, but t h e u n i t was

t h e b o l t holes. operable.

Later D E 3 w a s replaced with a surplus d i e s e l generator from

another f a c i l i t y .

The replacement unit was similar t o t h e o r i g i n a l except

t h a t it w a s s t a r t e d by compressed a i r instead of by an e l e c t r i c starter.

16.7

100-kVa Variable Frequency Motor-Gen r a t o r Set

A temporary i n s t a l l a t i o n of a used variable frequency MG s e t w a s made

at t h e MSRE t o supply t h e f'uel-pump motor during s p e c i a l t e s t s .

The fuel-

pump motor was operated from 610 rpm t o 1165 rpm (maximum speed with MG

s e t ) over a period of several weeks. The automatic controls f o r t h e variable speed and generator voltage did not work properly durin load.

e i n i t i a l checkout using a dummy resistance

After replacement of some'defective components i n t h e control c i r -

c u i t s , t h e f u e l pump was operated from t h e MG set.

bi*,

The operation of t h e

'automatic controls were s t i l l not completely s a t i s f a c t o r y and some manual

268

adjustments were required.

Frequently t h e fuel-pump motor was stopped be-

cause of over adjustment of t h e manual controls.

Also during t h e lower-

speed operation, t h e speed of t h e f u e l pump w a s not constant enough t o make t h e desired t e s t s .

Therefore, t h e automatic controls were bypassed and

manual controls of speed and voltage were i n s t a l l e d ,

Satisfactory ope=?

t i o n w i t h manual controls was not obtained u n t i l t h e f u e l pump was operated

at about 35 V above rated voltage, assuming t h e fuel-pump motor voltage i s d i r e c t l y proportional t o speed.

16.8 Conclusion Y

Considering t h e age and the number of modifications made t o t h e existing e l e c t r i c a l system within t h e E R E building, t h e performance has been very satisfactory.

The main weakness of t h e e n t i r e e l e c t r i c a l system was

with t h e 13.8-kV feeders and transformers supplying t h e MSRE.

equipment w a s over 15 years old.

Most of t h e

During 1966, t h e r e were t h r e e momentary

interruptions of t h e power t o t h e MSRE caused by thunderstorms.

The e n t i r e

13-8-kV supply system was reviewed and improved t o m a k e it less susceptible t o damage by storms.

Since t h e improvements were made, there have been

only two interruptions due tâ&#x20AC;&#x2122;o storms i n t h e l a s t t h r e e years of operation. ! I

ie*s

269

17.

HEATERS

T. L. Hudson

17.1

Description

E l e c t r i c a l heaters were provided f o r a l l p a r t s of t h e salt system t o permit preheating of t h e c i r c u l a t i n g loops and t o maintain t h e temperature (In practice the of t h e salt at 1200'F during zero-power operation. heaters were kept energized during power operation as well.)

The t o t a l in-

s t a l l e d capacity was %go0 kW; t h e a c t u a l power requirement w a s less than h a l f t h i s amount.

It w a s possible t o supply a l l heaters e s s e n t i a l t o an

orderly shutdown, including enough t o keep thy s a l t molten i n t h e drain tanks, from a 300-kW diesel-driven generator.

A summary of t h e heaters

and i n s u l a t i o n on t h e salt systems i s given i n Table 17.1. The type of i n s t a l l a t i o n was determined by t h e location.

I n t h e re-

actor and drain-tank c e l l s , radiation from long-lived f i s s i o n products pre-

cluded d i r e c t maintenance a f t e r start of power operation.

There, t h e heat-

ers were designed for removal and replacement using long-handled t o o l s inserted through t h e c e l l roof.

I n t h e coolant c e l l , t h e piping could be

approached within a few minutes a f t e r a shutdown, and t h e heaters and insul a t i o n were t h e conventional type used i n component t e s t f a c i l i t i e s . Removable heater u n i t s i n the reactor and drain-tank c e l l s were of two types.

On t h e piping and heat exchanger, t h e heaters were nichrome ele-

me5ts i n ceramic p l a t e s , located on t h e inner surfaces of t h e removable u n i t s t h a t a l s o insulated t h e s i d e s and t o p of t h e equipment.

On t h e re-

a c t o r vessel, f u e l pump, and d r a i n tanks, t h e removable heaters were ins e r t e d i n t o penetrations i n permanently mounted insulation. All e l e c t r i c a l i n s u l a t i o n w a s ceramic type,

ecause of t h e high gamma-radiation f i e l d s (up

t o lo5 R/h i n t h e reactor c e l l during operation). Thermal i n s u l a t i o n i n t h e removable heater-insulation u n i t s was of t h e metallic, multiple-layer, without dusting.

r e f l e c t i v e type, which could withstand handling

The permanently mounted i n s u l a t i o n w a s low-conductivity,

high temperature low sullfur content,

ceramic fiber of expanded s i l i c a ,

with a t h i n metal sheet between t h e i n s u l a t i o n and t h e pipe surfaces (see

ORNL Drawing E-MM-â&#x201A;Ź3-51676).

_ . _" _

- - ."

....

. . . . . . . . . . . . . I

.. .

.. .~ -.. .. .

...

.- .

.. .

-. . . .. ..

- .

Table 17.1

Insulation

Number Heater Units

Elements

Removable

Permanent'

1135 ft of 3/8-in. by 0.03Fin. w a l l Inconel tubing. Each removable unit consists of 7 U-tubes.

Remewsble

Permanent'

3/b-in.diam, stainless s t e e l sheath 5-it long with ceramic insulation and nichrome elements.

Removable

177

Ceramic plate with nichrome elements.

Permanent

Permanent'

Tubular 0 . 3 1 5 4 ~OD Inconel sheath With nichrome elements.

Heat Exchanger

Removable

Removableb

Ceramic plate with nichrome elements.

Freeze Valve

Removable

Removable'

Tubular 0.3l5-in. OD Inconel sheath With nichrame elements.

Reactor Beck

Permanent

Permanentu

Tubular 0.3l5-in. OD Inconel sheath With nichrome elements.

Drain Tanks

Remvable

Permanent'

156

Ceramic plates with nlchromc elements

l-l/P-in.-diam Piping ( F i l l Lines )

Removable

Ceramic plates with nichrome elements

R&E

m e

Location Reactor C e l l

Heaters and Thermal Insulation

System Reactor

-1

pump

fi-in.-dIsm Piping

Typc

Thermal

Heaters

Number Heater

Number controls

Type Heater Elementsd

Powere Supply

Draln Cell

'Ceramic fiber of expanded silica. bMetallic, multiple-layer. d&!eramlc plates maximum i n s t d l w a t t a g e 30 watts/in2 of surface. H' normal parer supply; E emergency power supply.

TubuSar 0.315-in.

-J

- normal rating 500 watts/ft of heater. ,

._.

...

Table 17.1

Heaters

Type Thermal Insulation

Freeze Valves (Fill Lines)

Removable

Removableb

F i l l Line

See N o t e c

Permanent'

Type

Location

Drain C e l l

Coolant C e l l

System

Number Heater Units

(Continued)

Number Heater Elements 27 1

Type Beater Elements

N&E

Direct resistance heating of lc1/2-in. by 0.195-in. wd.1 INOR-8 pipe 65 rt long.

H&E

E&E

Ceramic p l a t e s with nichrome elements

H&E

Ceramic plates with nichrome elements

Ceramic plates and tubular 0.315-in. OD Inconel sheath with nichrome elements.

Permanent'

162

Ceramic plates and tubular 0 . 3 1 F i n . OD Inconel sheath with nichrome element 8 .

N&E

a Permanent

Ceramic plates and tubular 0.315-in. OD Inconel sheath with nichrome elements

N&E

Permanent'

Tubular 0.315-in. OD Inconel sheath with nichrone elements

Freeze V a l v e s (Transfer l i n e s )

Permanent

Permanent'

Coolant Pump

Permanent

a Permanent

Permanent

Permanent'

Radiator

Permanent

Drain Tank

Permanent

Piping

Powere Supply

Ceramic p l a t e s with nichrame elements

Permanent 1/2-in.-diam iping (Transfer ines )

>in.-diam

Hmber Controls

'Ceramic

f i b e r of expanded s i l i c a .

bMetallic, multiple-layer. 'The fill l i n e was heated by passing current through t h e wall of t h e pipe. The pipe was removable. normal r a t i n g 500 w a t t s / r t of heater. dCeranric plates maximum i n s t a l l wattage 30 watts/in2 of surface. Tubular 0.315-in. OD

N' -normal power supply; E

emergency pmer supply.

.......

..........

. . . . . . . . . . . . . . . . . . .

Table 17.1

Type

Location ~-

System ~~

Coolant C e l l (continued)

Heaters

Type Thermal Insulation

Number Heater Units

..j

......................

(Continued)

Number Heater Elements

Number Controls

Powere Supply

Ceramic p l a t e s with nichrome elements.

Ceramic plates ana tubular 0.35-in.

A m

Type Heater Elementsd ~

Freeze Valves

Permanent

Permanent'

F i l l Lines

Permanent

Permanent'

~ _ _ _

OD Inconel sheath With nichromc elements.

Flow Transmitters

Permanent

Permanent'

Tubular 0,315411. OD Inconel sheath with nichrome elements.

Level Element

Permanent

Permanenta

Tubular 0.315-in. OD Inconel sheath with nichrome elements

n h)

'Ceramic 'Ceramic e A

fiber of expanded. s i l i c a . plates d m u m i n s t a l l wattage

30 watts/inz of surface. Tubular 0.315-in.

- normal power supply; E - emergency power supply.

normal r a t i n g 500 watts/f't

of heater.

273

Maintenance requirements were minimized by designing t h e heaters w i t h excess capacity and operating them a t reduced voltage.

If some of t h e

h e a t e r elements f a i l e d , t h e other elements could be operated a t a higher voltage o r t h e power could be increased t o t h e adjacent heaters.

Heater

u n i t s t h a t were less accessible had spare heater elements i n s t a l l e d and connected, ready f o r use w i t h only minor, out-of-cell wiring changes. Control of t h e e l e c t r i c a l input t o t h e heaters w a s e n t i r e l y manual, i n response t o system temperatures displayed -on e i t h e r a multipoint tempera-

ture recorder o r a cathode-ray tube by a multipoint scanner.

A t l e a s t one

thermocouple was attached t o the piping o r equipment under each h e a t e r u n i t , with additional couples at p o t e n t i a l cold spots.

17.2 Preoperational Checkout To assure t h a t t h e heaters would perform properly and t o provide reference data f o r future trouble-shooting, and after i n s t acl l a t i o n .

a number of t e s t s were done before

A v i s u a l examination of t h e heaters and t h e wiring

i n t h e junction boxes was made and all obvious e r r o r s were corrected.

Con-

t i n u i t y checks were completed and measurements of t h e h e a t e r c i r c u i t res i s t a n c e s and resistances t o ground were recorded. *

Each h e a t e r c i r c u i t was

then energized individually t o assure t h a t t h e proper thermocouple o r thermocouples responded. The e f f e c t i v e resistance of each h e a t e r c i r c u i t w a s calculated and

these were compared with measured r e s i s t a n c e of t h e heaters before and aft e r being i n s t a l l e d i n t h e c e l l .

Several of t h e r e s i s t a n c e measurements

were made using a volt-ohm-meter but t h e accuracy of these readings was usua l l y not good enough.

The f i n a l r e s i s t a n c e measurements were made with a

Wheatstone bridge and were recorded t o t h e nearest one-hundredth ohm. Table 17.2 i s a summa

f r e p a i r s made t o t h e heaters during t h e pre-

operational checkout.

17.3 System Heatup The heatup of t h e coolant, f u e l , and drain tank systems was s t a r t e d as

soon as construction and checkout of t h e systems were completed.

During

t h e first heatup, a l l heaters operated s a t i s f a c t o r i l y and t h e r e were no

274

Table l7,2

Summary of Preoperational Heater Checkout

Repair o r Replacement O f Heaters

Reactor Cell

Corrected wiring Replaced open heater element

Drain Cell Coolant C e l l

Replaced broken ceramic beads on removable unit I

yo&â&#x201A;Źf i e d removable heater box to eliminate possible short c i r c u i t i n ceramic bead insulated leads

Modified heat box t o permit remote maintenance

Replaced broke M I cable end seal

Replaced 1 MI cable

Does not apply t o coolant c e l l ,

275

heater failures, but t h e r e were a few areas where t h e temperature w a s low because of a i r leakage and lack of thermal insulation.

These areas were

t h e i n l e t t o t h e fuel-pump furnace, north end of heat exchanger, t r a n s f e r l i n e entrance t o t h e drain-tank furnace, and t h e coolant r a d i a t o r .

After

t h e a i r leakage was reduced and thermal i n s u l a t i o n i n s t a l l e d , s a t i s f a c t o r y temperatures were obtained with t h e heaters as i n s t a l l e d .

The voltage s e t -

t i n g required f o r each heater w a s determined during t h e s t a r t u p t e s t i n g and stops were s e t t o l i m i t t h e controls t o hold t h e system temperatures t o 1300°F o r below.

The general heatup o f t h e systems and t h e d i f f i c u l t i e s

a r e discussed i n t h e following t h r e e sections,

17.3.1

Coolant and Coolant Drain.-Tank Systems Heatup..

The i n i t i a l heatup of t h e coolant system i n . t h e coolant c e l l and.coo1-

ant , d r a i n c e l l was s t a r t e d on October 13, 1964.

The coolant pump was run

during t h e heatup with normal helium purgeaan'd t h e heaters were adjusted t o l i m i t t h e heatup r a t e t o about 100°F/hr.

Heaters located i n t h e c e l l

penetration sleeves on l i n e s 200 and 201were operated at a reduced s e t -

t i n g t o provide a temperature gradient u n t i l t h e heatup of t h e r e a c t o r c e l l was s t a r t e d so as t o reduce t h e thermal s t r e s s e s at t h e anchor point, Heater and thermocouple leads on top of t h e r a d i a t o r enclosure were observed t o be overheating as t h e system was being leveled at a reference The r a d i a t o r was cooled t o ambient temperature and

temperature of 350°F,

a section of t h e t o p of the r a d i a t o r w a s found t h a t had not been insulated. The i n s u l a t i o n w a s i n s t a l l e d and a second heating of t h e r a d i a t o r was started.

Some of t h e r a d i a t o r heaters were disconnected i n an unsuccessful

e f f o r t t o achieve a s a t i s f a c t o r y temperature d i s t r i b u t i o n . w a s 1073 t o 1245OF and t h e o u t l e t tubes were 620 t o 1000°F.

The r a d i a t o r This condition

was reached nine days a f t e r

heatup w a s s t a r t e d ,

remainder of t h e coolant sy

i n t h e coolant c e l l was heated t o 1047 t o

During t h i s time t h e

59OF except on l i n e 202 between t h e r a d i a t o r and t h e coolant pump.

There

s 400°F difference among temperatures along t h i s section of l i n e under a

s i n g l e h e a t e r control.

The s p

p a r t o f t h e permanent i n age.

The coolant drain

culty with a t

was reduced t o about 175OF by replacing e a r t h e r a d i a t o r o u t l e t t o reduce inleak-

was heated . t o about 1200'F without any d i f f i -

e r a t u r e differenlce of 30°F between t h e highest and lowest

thermocouple reading.

27 6

1p The r a d i a t o r w a s cooled t o room temperature f o r inspection and t o re-

A f t e r r e p a i r s were made, t h e r a d i a t o r w a s again

p a i r t h e doors and seals. heated,

a 500'F

With t h e best temperature d i s t r i b u t i o n obtained, t h e r e was about spread on t h e r a d i a t o r o u t l e t tubes.

t i v e l y l a r g e heat leak existed,

This indicated t h a t a rela-

An inspection was made of t h e r a d i a t o r

enclosure near t h e o u t l e t h e a t e r and s i x compartments i n t h e enclosure wall were found t o be without insulation.

After t h e proper i n s u l a t i o n w a s in-

stalled and gaps i n t h e v i c i n i t y of t h e o u t l e t piping were caulked, satis-

factory temperatures were obtained with a l l h e a t e r s i n service as o r i g i n a l l y designed.

Localized overheating i n t h e areas above t h e r a d i a t o r enThe heat leaks were primarily a t openings

closure w a s again observed.

through which t h e ceramic-beaded insulated e l e c t r i c a l leads entered t h e enclosure,

Some minor i n s u l a t i o n damage w a s sustained on t h e wiring i n s i d e

t h e junction boxes mounted about 12 inches above t h e t o p of t h e radiator enclosure During t h e following shutdown, l a t e i n 1965, r a d i a t o r doors of improved design were i n s t a l l e d ,

Heat l o s s e s due t o air leakage around t h e

doors continued t o present a problem,

To obtain an acceptable temperature

d i s t r i b u t i o n , it w a s necessary t o shim t h e gaskets and disconnect some of t h e heaters, so as t o give a non-uniform heat input over'the face of t h e radiator,

When t h e main blowers were operated t o t e s t t h e doors, other

d i f f i c u l t i e s became apparent.

One of t h e s e w a s leakage of hot a i r i n t o t h e

region j u s t above t h e r a d i a t o r enclosure which caused overheating of elect r i c a l i n s u l a t i o n on t h e numerous thermocouples and h e a t e r leads i n Junction boxes located i n t h i s v i c i n i t y ,

It w a s necessary t o r e l o c a t e Qunction boxes

i n cooler locations along t h e c e l l w a l l and i n s t a l l higher temperature elect r i c a l i n s u l a t i o n on the thermocouple and h e a t e r leads over t h e enclosure, During-the approach t o f u l l r e a c t o r power, heating t h e empty r a d i a t o r t o an acceptable temperature d i s t r i b u t i o n p r i o r t o a f i l l had become increasingly d i f f i c u l t , and on t h e last f i l l , some of t h e tubes could not be heated above t h e melting point of salt u n t i l t h e r a d i a t o r was f i l l e d and circulatfon was started. door seals were modified.

After t h e shutdown i n J u l y 1966, t h e r a d i a t o r The e l e c t r i c a l heat required t o heat t h e empty

r a d i a t o r was reduced from 108 kW t o 62 kW by t h e door seal modification. I

I .

277

The four flow transmitters located i n t h e enclosure above t h e radiator required a higher heater s e t t i n g when e i t h e r o r both blowers were on because of a i r leakage past t h e transmitter. 17.3.2

Fuel System Heatup

The i n i t i a l heatup of the f u e l system and t h e coolant piping i n t h e reactor c e l l w a s s t a r t e d November 1, 1964.

A t t h i s time t h e coolant system

i n t h e coolant c e l l had already been heated. About nine days were spent on t h e first heatup and leveling of t h e temperatures.

The heatup r a t e w a s limited by t h e heatup rate of the re-

actor because there i s about 8000 pounds of INOR and 8000 pounds of graphi t e t o be heated.

To a i d i n heating up t h e graphite and t o help purge out

any remaining oxygen, t h e f u e l pump was run during t h e heatup with normal I

helium purge.

Also the f u e l system piping and heat exchanger were operated

about 150'F higher t o help heat up the reactor. The system temperatures were brought up t o an acceptable range (1000 t o l25O'F)

with a few exceptions.

Temperature a t t h e freeze flanges and

on t h e coolant l i n e s at the c e l l penetrations were, as expected, below t h e liquidus temperature of t h e salt, but only over short sections.

Thermo-

couples under the heater located between freeze flange 100 and t h e fuelpump f'urnace could be heated only t o 700 t o 900' reduced by closing gaps i n the heater base.

u n t i l heat losses were

There was about 300'F

spread

between t h e coolant i n l e t nozzle and the f u e l o u t l e t nozzle on t h e heat exchanger.

The spread was reduced by closing gaps between t h e heat-exchanger

base and adjacent heater.

The time required t o heat up t h e reactor and f u e l

system a f t e r t h e i n i t i a l heatup w a s about t h r e e days. Fuel Drain Tank System Heatup

17.3.3

The i n i t i a l heatup of t h e drain-tank c e l l was s t a r t e d November 1 2 ,

1964. This heatup included drain tanks, FD-1, FD-2, FFT, and l i n e s 103, 104, 105, 106, 107, 108, 109, and 110. heat ed

The helium purge was s

This w a s t h e l a s t of t h e systems

from t h e f u e l system through l i n e 103 t o

the drain-tank system and then exhausted a t l i n e 110. The heaters were adjusted t o limit t h e heatup r a t e t o less than 100'F/hr.

A l l of the system w a s heated t o an acceptable range (1000 t o

278

1250째F) except where l i n e s ( l i n e s 107, 108,-and 109) penetrated t h e tank On l i n e 107, t h e temperature was brought up t o 95OoF, but only

furnaces.

by operating t h e heaters i n t h e penetration at 220% of t h e i r r a t i n g ,

The

other two l i n e penetrations were somewhat b e t t e r , but s t i l l unsatisfact"0z-y. The insulation on these l i n e s near t h e penetrations was removed,

Then t h e

heaters were t i e d closer t o t h e pipes and additional i n s u l a t i o n was i n s e r t ed between t h e heaters and t h e furnace w a l l s ,

Temperatures a t these pene-

t r a t i o n s then reached 1200'F with t h e heaters a t l e s s than t h e i r r a t e d capacity.

17.4 Heater Performance The system was operated at high temperature f o r about 10 months before t h e prepower shutdown. heater elements f a i l e d .

During t h e f i r s t 1500 h r of operation, only two Both o f these were in removable heater units on

t h e main f u e l c i r c u l a t i n g l i n e s and had spare elements i n t h e ceramic heat-

er p l a t e s . changes.

"he spare elements were put i n t o s e r v i c e by out-of-cell wiring A summary of t h e h e a t e r failures and repairs, a r e included i n

Tables 17.3, 17.4, and 17.5.

Six a d d i t i o n a l heater-element f a i l u r e $ and

m e l e c t r i c a l ground at a disconnect were discovered during t h e chebkout

f o r Run 2.

I n four of t h e heaters, t h e f a i l u r e s occurred at t h e Junction

of t h e extension l e a d and t h e h e a t e r element i n s i d e t h e ceramic p l a t e . These elements and t h e e l e c t r i c a l ground were repaired before s t a r t u p , and t h e other two elements were l e f t out of s e r v i c e f o r Runs 2 and 3* During subsequent operation, one i n - c e l l h e a t e r f a i l u r e was noted, and one f a i l u r e occurred at a h e a t e r power supply,

(The l a t t e r was repaired

It may be noted t h a t a number of h e a t e r f a i l u r e s could occur

immediately.)

without s i g n i f i c a n t l y a f f e c t i n g t h e operation of t h e r e a c t o r system, Three more heater-element f a i l u r e s were found during t h e prepower shutdown,

I n addition, a number of minor defects (grounds, damaged connectors)

were found.

All t h e defects, including those l e f t from e a r l i e r operations,

were corrected. An important cause of ceramic heater-element f a i l u r e has been separa-

t i o n of t h e extension lead from t h e h e a t e r element.

This i s apparently

due t o a combination of a design weakness and excessive flexing during

c Table 17.3

Date

System

Summary of Heater Failures i n the Reactor Cell

Heater

Type Failure

Repairs Made

CRITICAL EXPERIMENT PERIOD

11/9/64 12/ 31/64

Reactor

R-1-2

5-in. piping

H-100-1

One U-tube grounded Failure of top element

1/5/65 3/22/65

5-in. piping

H-102-5

Failure of top element

5-in. piping

H-100-1

Connected up spares t o continue Run 1. Replaced ceramic elements before s t a r t of Run 2.

3/22/65

%in. piping

H-102-5

Replaced ceramic elements before start of Run 2.

H-201-4B

Lead grounded Lead grounded

4/30/65 7/20/65

Heat exchanger

Hx-2

U-tube shortened 3 i n . Connected up spares t o continue Run 1.

Realigned disconnect. Removed sharp edge i n lead. R>

PREPOWER SHUTDOWN

Two fuel-pump heater units resistance increase from 12 t o 17 ohms since the preoperational checkout. These w e r e removed from t h e reactor c e l l and t h e mechanical j o i n t between the s o l i d and flexible nickel lead w a s highly oxidized. A l l five fuel-pump heater units were removed from the reactor c e l l and a welded connection was made at each lead extension.

7/30/65

Fuel pump

FP1&2

7/30/65

5-in. piping

H-102-1

Broken disconnect replaced.

7/30/65 7/30/65

5-in. piping 5-in. piping

H-200-10

Broken disconnect replaced.

H-200-11

Broken disconnect replaced.

7/30/65

5-in. piping

H-201-6

Broken disconnect replaced. POWER OPERATION

12/29/65

Heat Exchanger

HX-1

Heater leads t o 2 out of 8 elements f a i l e d

Operation continued. 4/9/68 (see below)

Repaired

9/15/66

Heat Exchanger

HX-1

Heater leads of 2 more elements failed.

Operation continued. Repaired 4/9/68 (see below).

Table 17.3 (continued)

Date

System

Heater

Type Failure

Repairs Made

POWER OPERATION

(continued) 10/13/66

Heat Exchanger

HX-1

Heater leads of 2 more elements failed.

Operation continued. Repaired 4/9/68 (see below).

Heat Exchanger

HX-1

A l l elements out of' service due t o heater leads.

Operation continued. Repaired 4/9/68 (see below).

5-in. piping

H-102-1s

1 of 3 i n s t a l l e d spares

No repairs required. Adequate capacity i n others.

failed

9/13/67

Freeze valve

FV-103

Complete f a i l u r e

12/12/67

Heat Exchanger

HX-2

Open lead good

12/19167

He a t Exchanger

Hx-2

Heater leads of 6 out of 8 elements f a i l e d

4/9/68

Heat Exchanger

HX-1 HX-2

Heaters were removed from reactor c e l l and were repaired by welding the heater leads t o the terminal s t r i p s . A l l heater elements were good.

10/1/68

5-in. piping

H-100-1

Failure of 1 side element

Connected up spares.

10/21/6 8

Heat Exchanger

HX-1

Heater grounded

Taken out of service.

8/2/69

5-in. piping

H-10 2- 5

Failure of 1 side element

Connected up spares.

No repairs required. Adequate temperature without heater.

- a l l elements

Operation continued. Repaired 4/9/68 (see below). .

Operation continued. Repaired 4/9/68 (see below).

c Table 17.4

Date

Summary of Heater Failures i n the Fuel Drain Tank C e l l

Heater

System

Type Failure

Repairs Made

CRITICAL EXPERIMENT PERIOD

W65 5/65

F i l l Line

H-106-1

Failure 1 side element

Replaced element before Run 2.

F i l l Line

H-104-1

Failure 1 side element

Replaced a l l elements -new design during prepower shutdown.

Bad M I cable

Replaced, M I cable during prepower shut down.

- loose

Repaired 8/6/65 (see below).

Top element connectidn

5/65

Freeze Valves

FV-105-3

Top and 1 side element loose connections

5/65

Freeze Valves

FV-106-1

Top elements -loose connection

Hepaired 8/6/65 (see below). Repaired 8/6/65 (see below).

PREPOWER SHUTDOWN

8/5/65

Drain Tank

8/6/65

Freeze Valves

Replaced a l l heater elements (new design) i n freeze valve heaters FV-104-1, FV-104-3, FV-105-1, FV-105-3, FV-106-1, FV-106-3, and H-104-5.

10/10/65

F i l l Line

H-106-1

â&#x20AC;&#x2122;

FD2-1

Female disconnect grounded

Failure 2 elements due t o bad thermocouple

Added 3 in. of ceramic beads t o t h e 3 leads Q,

Replaced a l l elements -new design.

POWER OPERATION 1/21/67

F i l l Line

H-106-4

Failure 1 side element

Continued operation using two good elements and additional heat from adjacent heaters.

Table 17.5

Date

System

Heater

Summary

of Heater Failures i n the Coolant Cell

Type Failure

Repairs Made

PREPOWER SHUTDOWN CHECKOUT

8/12/65

5-in. penetration H-200-14s

5-in. penetration H-201-11 5-in. penetration H-201-11s Radiator

CR-2

Radiator

CR- 3

5-in. piping 5-in. piping

H-201-12 H-202-2

out of 8 elements failed loose connection

out of 8 elements failed loose connection out of 8 elements failed loose connection element failed loose connection element failed loose connection Element grounded Failure 1 element

Replaced a l l heater elements (new design) i n heaters H-200-14, H-200-15, H-201-10, H-201-11, and spares.

Replaced element.

Replaced element. Replaced element, Replaced element.

POWER OPERATION

5-in. piping Checkout after Run 7 8/17/66 Radiator 8/17/66 5-in. piping 8/17/66 Radiator inlet header 8/17/66 Coolant Pump

H-202-2

Broken lead

Reconnected.

CR-3 H-202-2 CR-7

Broken extension wire Failure 1 element Failure 1 element

CP-2

Failure 1 element

8/17/66

H-206-1s

Failure 1 spare element

Replaced element. Replaced element. Disconnected element capacity in others. Disconnected element capacity in others. No repairs necessary. capacity i n others.

1/17/66

F i l l line

- adequate - adequate Adequate

Table 17.5 (continued)

Date

System

Heater

Type Failure

Repairs Made

Repairs af’ter R u n 11

3/30/67

Radiator

CR-6

Failure 1 element

Replaced element.

Broken ceramic 2 broken lead wires. Nickel lead wire had crystallized.

Replaced element. Removed defective sections and welded leads and dye-checked.

1 element grounded

Disconnected element. capacity i n others.

Repairs af’ter Run 13

512

”

Adequate

284

i n s t a l l a t i o n of t h e elements.

The j o i n t w a s redesigned and new elements,

which incorporate t h e change, were i n s t a l l e d when such failures occurred. Two fuel-pump h e a t e r u n i t s resistance increased from 12 t o 17 ohms. These were removed from t h e r e a c t o r c e l l and t h e mechanical j o i n t between t h e h e a t e r s o l i d and flexible n i c k e l extension lead was found t o be highly oxidized,

A l l f i v e fuel-pump h e a t e r u n i t s were removed from t h e reactor

c e l l and a welded connection was made a t each l e a d extension.

No other

trouble w a s encountered with these heaters. Before and during t h e prepower shutdown, d i r e c t maintenance w a s performed on t h e heaters i n the r e a c t o r and drain-tank c e l l . 17.4.2

During Power Operation

A l l heaters performed properly during t h e heatup of t h e f u e l and

coolant loops i n December 1965 f o r t h e s t a r t of power operation, with one minor exception.

There was a broken lead on a h e a t e r between t h e r a d i a t o r

and t h e coolant pump.

It w a s simply reconnected.

A f t e r t h e s t a r t u p , one

of t h r e e groups of elements i n one of t h e heat-exchanger h e a t e r u n i t s failed.

No action was taken a t t h i s t i m e because t h e other elements were

enough t o produce t h e desired temperature.

A summary of t h e h e a t e r f a i l u r e s

LlJ

and r e p a i r s made during power operation i s included i n Tables 17.3, 17.4, and 17.5. During t h e shutdown i n September 1966, a r e s i s t a n c e check w a s made on a l l i n - c e l l h e a t e r s at junction boxes located outside t h e r e a c t o r and

drain-tank c e l l and compared with i n i t i a l resistances.

Other than t h e

heat-exchanger h e a t e r , only one h e a t e r w a s found that had f a i l e d .

This was

one of t h r e e i n s t a l l e d spare heaters on t h e v e r t i c a l section of pipe under t h e heat exchanger.

No r e p a i r s were required, since t h e r e i s adequate ca-

pacity i n t h e other heaters on t h i s section of pipe. The resistance of t h e coolant-system h e a t e r s located outside of t h e r e a c t o r c e l l was a l s o checked.

Five heaters were found t h a t had f a i l e d .

One w a s on t h e r a d i a t o r , one on t h e coolant system 5-in. piping, one on t h e cdolant-pump furnace, one i n t h e r a d i a t o r o u t l e t header, and t h e other one, a spare heater, was on t h e f i l l - l i n e piping.

The h e a t e r s on t h e 5-in.

piping and on t h e r a d i a t o r were replaced, t h e r e w a s adequate h e a t e r capacity at t h e other locations.

285

6.i

Aside from heater-element f a i l u r e s , a remote disconnect f o r one of t h e fuel-pump heaters was damaged during remote operation associated w i t h t h e thawing of a salt plug i n t h e fuel-pump offgas l i n e .

The h e a t e r was

plugged i n t o a space disconnect t o r e s t o r e it t o service. . By t h e last of October

1966, all of t h e h e a t e r elements i n one of t h e

heat-exchanger h e a t e r u n i t s had f a i l e d .

Satisfactory heat-exchanger tem-

peratures were maintained without t h i s h e a t e r , even with t h e f u e l loop empty.

However, continuous c i r c u l a t i o n of t h e coolant salt was maintained

u n t i l after Run 11, so t h e full e f f e c t of t h e h e a t e r failure could not be determined.

Tests were performed w i t h both t h e f u e l and coolant loops

empty after Run 11 t o determine t h e need f o r t h i s h e a t e r i n preheating t h e system from a cold condition.

When helium c i r c u l a t i o n was stopped i n t h e

f u e l system, t h e temperature d i s t r i b u t i o n was s a t i s f a c t o r y f o r a f i l l w i t h out t h e h e a t e r operating.

With helium c i r c u l a t i o n i n t h e f u e l system, one

s a t i s f a c t o r y temperatures temperature decreased t o below 8 0 0 ~ ~ Since . could be achieved without it, the f a i l e d h e a t e r was not repaired at t h i s time.

While t h e r e a c t o r system was being heated up f o r Run 13, Heater FV-103 was l o s t by an open c i r c u i t .

This 1.5-kW bent-tubular-type

h e a t e r normally

supplies about 200 W of heat t o a b i n . section of t h e f u e l drain l i n e within t h e r e a c t o r vessel furnace, between t h e freeze valve and t h e resistance-heated section of t h e l i n e .

The purpose of t h e h e a t e r w a s t o

help control the temperature p r o f i l e through t h e freeze valve.

After t h e

heater f a i l e d , tests with f l u s h s a l t shoued t h a t by proper adjustment of t h e cooling a i r controls, t w i t h thaw times i n an acce

eeze valve could be maintained r e l i a b l y range.

(Thaw time w a s around 1 2 min when

t h e r e a c t o r vessel was at 1 1 8 0 ~ and ~ t h e center of t h e freeze valve was at

495OF.)

Therefore, t h i s h e a t e r was not replaced.

I n December 1967, a l e a d t o t h e adjacent f a i l e d heat-exchanger h e a t e r opened, thereby reducing t h e

utput of t h i s h e a t e r by 50%. A week l a t e r a

second p a r t i a l f a i l u r e f u r t h e r reduced.the output t o only 33%of normal. Although these f a i l u r e s d i d not a f f e c t operattons as long as s a l t w a s kept c i r c u l a t i n g , t h e lack of heat i n t h e two adjacent heaters made it necessary

t o r e p a i r them during t h e shutdown af'te/ Run 14.

These two u n i t s were

286

removed from t h e reactor c e l l e a r l y i n t h e shutdown (on t h e fourth and f i f t h day a f t e r t h e end of full-power operation) t o see i f r e p a i r s were possible o r i f replacement u n i t s would be required, The trouble was found i n t h e Junction boxes mounted on t o p of t h e h e a t e r s , where t h e l e a d wires

from several of t h e h e a t e r elements had burned i n two at t h e i r screwed connections t o t h e terminal s t r i p s ,

Repair w a s complicated by t h e induced ac-

t i v i t y i n t h e assemblies, which produced a gamma r a d i a t i o n f i e l d of about

3 R/hr at 1 ft, A temporary work s h i e l d was s e t having concrete block w a l l s and a top of s t e e l p l a t e and lead block.

Direct maintenance on t h e Junc-

t i o n was done through a small opening i n t h e top.

The terminal s t r i p s and

connections were severely oxidized, apparently due t o high temperatures This s i t u a t i o n w a s corrected by i n s t a l l i n g nickel term-

during operation,

i n a l s t r i p s and welding t h e h e a t e r leads d i r e c t l y t o them.

The copper wire

from t h e terminal s t r i p t o t h e disconnect w a s a l s o replaced with No, 12 nickel wire welded t o t h e terminal s t r i p .

Aside from t h e damage i n t h e

Junction boxes, both heat-exchanger heaters were i n good condition, and a f t e r t h e r e p a i r s , both were r e i n s t a l l e d through t h e maintenance s h i e l d

without unusual d i f f i c u l t y . Af'ter t h e u n i t s were r e i n s t a l l e d i n t h e c e l l , it was found t h a t t h e

current on one phase of south end u n i t was zeroo This proved t o be due t o a fault i n t h e permanently mounted l e a d wire i n t h e c e l l .

The fault was

circumvented by i n s t a l l i n g a Jumper cable between t h e h e a t e r disconnect and

a spare disconnect.

The heaters functioned s a t i s f a c t o r i l y during t h e sub-

sequent heatup of t h e system. One of t h e elements i n t h e h e a t e r u n i t on l i n e 106 adgacent t o t h e Junction of l i n e 103 with l i n e s 104, 105, and 106 f a i l e d during t h e middle of January 1967.

This was not replaced since it was possible t o maintain

aaequate temperature using t h e other elements and adjacent heaters. There have been t h r e e failures of r a d i a t o r headers due t o broken nickel lead wires.

The leads look l i k e t h e nickel had crystallized.,

Several bro-

ken ceramic bushings have been replaced. The heater located on t h e south end of t h e heat exchanger developed a p a r t i a l ground i n t h e l a t t e r p a r t of October 1968.

I n order t o c l e a r t h e

ground detector l i g h t s on Bus 5, t h e h e a t e r w a s taken out of service.

This LivJ

287

was possible because t e s t s had been performed a f t e r Run 11which determined t h a t t h i s h e a t e r was not e s s e n t i a l f o r preheating t h e system from a cold condition.

The cause of t h e grad was suspected t o be a broken ceramic

bead on the l e a d wire between a h e a t e r element and t h e junction box mounted on t o p of t h e unit. About a 30% drop i n current on one heater on t h e f u e l l i n e between t h e reactor and t h e f u e l pump (H-100-1) occurred l a t e i n September 1968.

After

a r e s i s t a n c e check indicated a s i d e h e a t e r element had f a i l e d , t h e spare

heater elements were placed i n service. About a 30% drop i n current on t h e h e a t e r on t h e f u e l l i n e between t h e heat exchanger and t h e reactor (H-102-5) occurred i n t h e e a r l y p a r t of Augu s t 1969.

A f t e r a resistance check indicated t h a t one of t h e t h r e e h e a t e r

elements had f a i l e d , t h e spare h e a t e r elements were placed i n service. The individual ammeters f o r each h e a t e r c i r c u i t were read and recorded daily.

Since operation s t a r t e d , most h e a t e r f a i l u r e s caused a s u f f i c i e n t

change i n ammeter reading s o as t o be noticed.

Variation i n voltage sup-

p l i e d t o t h e MSRE w a s usually l e s s than 2% over a 24-hour period.

2% change caused l e s s than a 2'F/hr isothermal conditions. tings.

change of t h e system temperature under

This w a s e a s i l y adjusted by changes i n h e a t e r s e t -

After t h e system had been brought t o equilibrium at temperature it

was not usually necessary t o m a k e f u r t h e r adjustments. including power requirements f o r 1200'F Table

A total

17.6.

Typical h e a t e r data

isothermal conditions a r e given i n

The four types of f a i l u r e s experienced w i t h the h e a t e r leads a r e given i n Table 17.7 along with methods used t o make repairs.

17.5 Before nuclear power operation was started, system cooldown t e s t s were run t o determine t h e e f f e c t s on t h e system i f e l e c t r i c a l power w a s l o s t . The cooldown of t h e f u e l and coolant systems w a s done simultaneously since

these are connected at t h e heat exchanger.

These t e s t s which were done

with t h e r a d i a t o r doors closed a i d not show any d i f f i c u l t i e s .

A much more

serious condition occurred on two occasions a f t e r nuclear power operation was s t a r t e d when t h e cold r a d i a t o r doors were scrammed and t h e coolant pump

Table 17.6

Heater Numbers R1, R2, & R3 FT-1 & FT-2 FFT-1 & m - 2 FD1-1 & FD1-2 FD2-1 & FD2-2

Equipment

Typical Heater Data

1200'F Operating Data Average Heat Loss Heat Loss Per f t 2 of Surfacb % of Max k W Installed Watts

Maximum Installed kW

Insulation' Thickuess in.

Reactor Vessel Fuel Pump

69.6 18.2

45 47 45

FD-1

FD-2

29.9 12 12.8

16.1 14.9

112

28.5

34.3

74.2

33.1

68.7

47 ' 57 57

43 66

30' of 5" piping 23.5' of 5" piping

H202-2

29' of 5" piping

RCH-1

7' of 5" piping

RCH-2

5' of 5" piping

7.5' of 5" piping

RCH-3

4 4 4 Metallic MultipleWYer Metallic MultipleLwer Metallic MultipleLayer ~

'Ceramic

Per f t of

Per f t of Pipe H200-13 H201-22

Calculated Heat Loss Per ft2 of Surface Watts

15.3

150 140

Pipe 126 126

4.7

19.8

162

126

2.5

19.2

357

2.0

400

23.75 21.6

5.4

22.8

3.3

23.75

03 03

15 ~~

~~~

3.4

22.7 ~

453 ~

fiber of expanded s i l i c a .

'Design based on heat loss of 600 w a t t / f t of pipe on t e s t unit.

289

Table 17.7

Heater

Type of Heater Lead Failure and Repairs

Failure

Correction Made

(1)Ceramic

Separation of extension lead from heater element

Designed leads with a cross bar

( 2 ) Fuel pump tubular

Excess oxidation at mechani c a l j o i n t between s o l i d and f l e x i b l e head

Made welded j o i n t s and dye-checked

( 3 ) Heat exchanger

Excess oxidation at mechani c a l j o i n t s a t terminal s t r i p i n junction box

Made welded j o i n t s and dye-checked

(4) Radiator

Nickel extension lead c r y st a l l i zed

Removed defective sect i o n , welded and dyechecked

290

stopped.

S a l t w a s frozen i n some of t h e r a d i a t o r tubes but was succesfully

melted out with no apparent damage t o t h e radiator.

These coolant tests

and radiator d i f f i c u l t i e s are discussed i n t h e following four sections,

17.5.1

Fuel and Coolant Systems Cooldown Rate, Radiator Door Closed

Three cooldown t e s t s were run.

In the f i r s t t e s t , t h e cooldown rate

was determined f o r t h e f u e l and coolant systems with s a l t circulat'ing i n both systems and with all t h e reactor and coolant c e l l heaters o f f except \

the heaters on t h e f i l l l i n e s ,

The f u e l and coolant system reference tem-

perature dropped from 1160 and ll69'F one hour and 52 min.

t o 1105 and 1115'F respectively i n

This i s a cooldown rate of 0,5'F/min.

In t h e second t e s t , t h e same heaters were o f f as i n t h e f i r s t t e s t and the f u e l and coolant pumps were turned off.

A f t e r t e n minutes, t h e radi-

a t o r heaters and t h e f u e l and coolant pumps were turned back on.

This s i m -

ulated a power f a i l u r e followed by t h e s t a r t u p of t h e d i e s e l generators, The f i e 1 and coolant system reference temperatures dropped from 1156 and

1185'~ t o 1089 and 1112'F respectively i n 2 hours and 50 minutes.

This i s

a cooldown r a t e of 0.4'F/min, In t h e t h i r d t e s t , all t h e heaters and t h e f u e l and coolant pumps were turned off and remained o f f f o r t h e duration of t h e test,,

The reference

temperatures dropped from 1150 and 1170'F t o 1120 and 1141'F respectively f o r t h e f u e l and coolant system i n 38 minutes.

This i s a cooldown r a t e of

0.,8~~/min. The reference temperature on t h e r a d i a t o r dropped from 1170 t o 1090'F i n the same period,

This is a cooldown r a t e of 2,1째F/min,

In all three tests t h e temperatures of t h e coolant f l o a t l e v e l indicators dropped rapidly and t h e i r heaters were turned back on. ~

17,5,2

Fuel Drain Tank Cooldown Rate

With flush salt i n FD-2, a test w a s made of t h e cooldown r a t e by turning off all FD-2 heaters.

w a s 1085'~.

The average temperature a t t h e s t a r t of t h e test

Ten and one-half hours later, t h e average temperature was 1020'F.

From t h i s data, it is estimated t h a t it wouid take at l e a s t 56 hours f o r FD-2 t o cool down from 1200'F t o 850'F upon l o s s of e l e c t r i c a l power t o the tank heaters.

Decay heating would'extend t h i s time.

291

17.5.3

Coolant Drain Tank Cooldown Rate

With coolant s a l t i n CDT, a t e s t was made of t h e cooldown r a t e by turning o f f a l l CDT heaters.

The average temperature of f i v e thermocouples

t o~ 1 0 8 5 ' ~ i n sixteen hours. dropped from 1 1 6 0 ~

From t h i s data it i s e s t i -

mated t h a t it would take at l e a s t 70 hours f o r CDT t o cool down from 1200'F t o 850'3'

upon l o s s of e l e c t r i c a l power t o t h e tank heaters.

17.6 Discussion The f u e l loop operated above 9OO'F loop above 9OO'F

f o r 27,438 hours.

f o r 30,848 hours and t h e coolant

There were 1 3 heating and cooling cy-

c l e s of f u e l loop and 12 f o r t h e coolant loop.

The four drain tanks, in-

cluding t h e drain and fill l i n e s , were operated near 1200'F continuously since t h e s a l t was added t o t h e tanks except f o r two times when FD-2 was cooled down f o r i n s t a l l i n g and removing equipment f o r t h e 233U addition. There w a s not an interruption of t h e power operation of t h e reactor due t o a heater failure.

k-,

The main weakness of t h e ceramic heaters was t h e separation of t h e extension leads from t h e heater elements.

20 correct t h i s , a t o t a l of 70

(30 i n t h e r e a c t o r and f u e l drain tank c e l l and 40 i n t h e coolant c e l l ) ceramic elements of an approved design were i n s t a l l e d during t h e prepower shutdown.

Performance since then has been excellent.

There were only

5 failures which could be a t t r i b u t e d t o t h e ceramic elements out of 221 installed units.

292

18.

SAMPLERS , ,

? .

- "

A. I. Krakoviak

I n general, . . . . of both t h e f u e l and coolan ystem w a s accom. . sampling plished by a power-driven c i b l e which'lowered a s m a l l capsule i n t o t h e c i r culating l i q u i d of t h e p G p bowl.

The capsule w a s then .raised and t h e s a l t

was permitted t o s o l i d i f y before f i n a l r e t r i e v a l and shipment t o t h e Anal y t i c a l Laboratory f o r .analysis.

Lead shielding (%I-0 inches) w a s i n s t a l l e d

around t h e major components of t h e - f u e l sampler and a stainless-steel-clad Double containment

lead shipping cask was used t o transport t h e s q l e .

was accomplished by moving t h e sample ( a f t e r i s o l a t i o n from t h e pump bowl) through a s e r i e s of containment areas while maintaining at l e a s t two valves o r membranes between t h e f u e l system o r sample and .the environment, Since t h e induced a c t i v i t y i n t h e coolant s a l t i s very short-lived, no shielding w a s required during sampling and only s i n g l e containment w a s necessary during sample transport. A t o t a l of 81 coolant salt samples were taken and a t a t a l . o f 745 f u e l

sampling cycles were completed during t h e f i v e years of MSRE operation. Included i n t h e f u e l sampling t o t a l were 152 f u e l enrichments of either uranium o r plutonium,

I n addition t o sampling and enriching, t h e f u e l sam-

p l e r was a l s o used t o sample t h e gas space above t h e l i q u i d i n t h e pump bowl and t o make chemical additions such as beryllium, niobium, and FeF2 t o t h e fuel,

18,i Fuel Sampler-Enricher Since t h e fuel-sampler usage was more frequent and t h e complexity of i t s components were g r e a t e r than those of t h e coolant sampler, t h e major p a r t of t h i s report describes t h e operating experience with t h e samplerenricher,

A b r i e f description of t h e coolant sampler and a discussion of

i t s operation are included at t h e end of t h i s r e p o r t ,

18.1.~Description of t h e Fuel Sampler-Enricher The component arrangement of t h e f u e l system sampler-enricherS0 i s

shown i n Fig. 1 8 , ~The vacuum pumps and helium purge gas system required f o r pressure and atmosphere control a r e not shown.

The drive motor f o r

293

Fig. 18.1

Compo

for Sampler-Enricher

294

the cable, the l a t c h onto which capsules were hung were located i n primary

containment area 1 C which w a s approximately 17-1/2 f e e t from t h e pump bowl. C r i t i c a l closures such as t h e operational and maintenance valves (motordriven gate valves)

access port t o area 1 C (doubly-sealed door), and t h e

removal valve ( b a l l valve) were buffered w i t h helium t o assure t h a t i f any sealing surfaces leaked, t h e leakage would be i n e r t gas i n t o t h e system. Metered quantities of helium were supplied t o these seals and t h e leakage

rates across t h e seals were measured by t h e f i n a l equilibrium buffer s e a l pressure.

Excess s e a l leakage i n t o t h e system o r t o t h e containment area

could be detected by a decrease i n buffer seal pressure and a pressure increase i n t h e adjacent volume.

To prevent t h e accidental opening o r closure

of compartment doors or valves i n improper sequence during sample t r a n s f e r , a system of e l e c t r i c a l interlocks based on buffer seal pressures and-me-

chanical signals along with a detailed check l i s t (6A-3) ( R e f . 51) w a s used. I n general, a sample capsule, similar t o t h a t shown i n Fig. 18.2, w a s loaded i n t o a doubly sealed transport tube which w a s lowered through t h e removal valve i n t o area 3A.

The lower p a r t of t h e transport tube containing

the capsule w a s unscrewed and l e f t i n area

3A whereas t h e upper p a r t was

withdrawn above t h e removal valve and t h e valve closed.

The access port

t o area 1 C was then opened and the capsule w a s hung on t h e l a t c h w i t h t h e manipulators (the l a t c h and l a t c h pin were so designed t h a t t h e pin could not be disengaged during sampling except i n t h i s area).

The access port

w a s closed and area 1 C w a s first evacuated then pressurized t o equal t h a t i n t h e pump bowl before the maintenance ( M V ) and operational (OV) valves were opened.

The cable drive motor w a s actuated u n t i l t h e capsule w a s im-

mersed i n the salt i n the guide cage of t h e pump bowl (lower p a r t of Fig.

18.1).

The sample w a s withdrawn 18 i n . where t h e salt w a s allowed t o freeze

i n the v e r t i c a l section of unheated 1-1/2 in. Sched-40 pipe before f i n a l r e t r i e v a l i n t o area 1 C .

As t h e capsule o r sample was moved.from one com-

partment t o another there were evacuation and pressurization s t e p s which removed adsorbed oxygen during capsule i n s e r t i o n and removed gaseous fission products during sample removal.

Capsule t r a n s f e r s inside t h e sampler

were viewed with a periscope. Enriching w a s accomplished by lowering an enriching capsule (Fig. 18.2) i n t o t h e pump bowl and allowing t h e frozen s a l t i n the capsule t o m e l t and

h.d

Fig. 1 8 . 2

Sample Capsule and Enriching Capsule

be washed out before r e t r i e v a l of t h e empty capsule ( t h e w a l l s of t h e enriching capsules w e r e d r i l l e d through t o t h e frozen salt i n several places

j u s t p r i o r t o loading t h e capsule i n t o t h e sampler). The sampler-enricher w a s d e s i w d t o make 1000 sampling or-enriching cycles f o r a period of one year;52 however, t h e MSRE and sampler were operated f o r 20 runs over a period of 5 years. 18.1.2

Experience with t h e Fuel Sampler-Enricher During Pre-Critical and Low-Power Operation

The sampler-enricher w a s i n s t a l l e d and put i n operation i n May 1965.

Sampler t e s t i n g , shakedown, and operator t r a i n i n g sessions were conducted concurrently with t h e sampling and enriching operations of Run 2.

During

t h i s period, 53 f u e l samples were taken, 87 enrichments made, and 20 opera t o r s t r a i n e d i n t h e use of t h e sampler. â&#x20AC;&#x2122;

Some of t h e problems encountered

during t h i s period were: 1.

B u f f e r gas leaks at t h e Operational (OV), maintenance (MV), and removal (RV) valves.

Failure of a solenoid valve on t h e removal valve actuator.

Failure of t h e removal valve t o close completely requiring manual closureo.

Failure of t h e access port (AP) t o close completely.

5. 6.

Momentary failure of cable drive motor on capsule r e t r i e v a l . Rupture of t h e flexible containment membrane (boot) a t t h e manipulator

7. 8.

Deformation of t h e manipulator a r m and fingers. Accidental drop of an empty capsule onto t h e operational valve.

These and additional problems encountered with t h e subsequent 86 sampling cycles made during t h e approach t o power runs a r e discussed i n more d e t a i l

i n t h e following paragraphs. B u f f e r G a s Leaks at t h e Operational and Maintenance Valves -During

t h e sampling operations of Run 2, leaks developed through t h e upper m e t a l to-metal

seats of both t h e operational and maintenance valves.

The oper-

a t i o n a l valve w a s subsequently removed; examination showed t h a t a t h i n black ring, which was e a s i l y removed, had formed at t h e upper s e a l i n g surface of t h e valve gate and a small quantity o f salt spheres ( < 1 gm) had

297

collected between t h e s e a t s of t h e gate.

Apparently s a l t p a r t i c l e s were

dislodged from t h e capsules during t h e operation where t h e capsules ( a f t e r having been dipped i n t h e pump bowl) were disengaged from t h e l a t c h above t h e valve.

When t h e stem and gate were lubricated, t h e valve sealed almost

completely; however on removal of t h e l u b r i c a n t , t h e buffer gas leak r a t e increased t o 2 cc/min through t h e upper gate s e a l and remained a t zero through t h e lower s e a l . A few sampling cycles a f t e r the valve w a s r e i n s t a l l e d , t h e leak r a t e

again increased t o 20 cc/min through t h e upper s e a l .

Apparently more s a l t

p a r t i c l e s had lodged between t h e buffer-gas s e a l i n g surfaces.

Repeated ef-

f o r t s t o blow t h e p a r t i c l e s from the sealing surfaces f a i l e d .

Since t h e r e

were t h r e e other s e a l i n g surfaces between t h e pump bowl and t h e sample access area and since t h e leak did not increase with continued sampling, t h e valve was not replaced. After approximately 86 additional sampling cycles, an opportunity t o clean t h e operational valve arose when t h e r e a c t o r was drained t o r e p a i r t h e cable d r i v e motor i n April 1966 ( t o be discussed l a t e r ) .

Cleaning t h e

s e a t i n g surfaces decreased t h e leak r a t e f o r a while u n t i l an empty cap-

sule w a s accidentally dropped on t h e gate.

Subsequent valve operation

again r e s u l t e d i n a high buffer-gas leak r a t e .

The b u f f e r gas s e a l a t t h i s

surface continued t o d e t e r i o r a t e slowly throughout t h e remainder of MSRE operations; t h e only consequence was a slow pressure buildup i n t h e containment volume above t h e valve which required periodic venting.

From

February 1967 u n t i l f i n a l shutdown, t h e b u f f e r supply pressure t o t h i s s e a l

w a s used only when t h e sampler was i n operation thereby obviating t h e periodic venting of area 1 C . During t h e l a t t e r par leakage of buffer gas t h r o

966 t h e r e had been a gradual increase i n e upper s e a t of t h e maintenance valve also.

Although t h e procedure s p e c i f i e d t h a t t h e maintenance valve be opened before t h e operational valve is opened, and that t h e operational valve be closed before t h e maintenance valve is closed, on a t l e a s t one occasion t h e reverse sequence w a s followed during closure and thus foreign matter could have been dislodged f r o m t h e upper valve onto t h e lower valve.

While t h e

leak r a t e remained r e l a t i v e l y small (Q25 cc/min), it was s u f f i c i e n t t o

298 cause d i f f i c u l t y with proper operation of t h e interlocks (an e l e c t r i c a l s i g n a l which indicated t h a t t h e valve was indeed closed was based on t h e a b i l i t y of seating surfaces t o maintain a specified b u f f e r pressure),

Since

cleanup would have been d i f f i c u l t and replacement of t h e valve w a s not warranted by t h e leak rate alone, a mechanical method of assuring t h a t t h e valve w a s closed was s u b s t i t u t e d f o r t h e pneumatic method.

No additional

problems with these valves were reported, Area 1 C Access Port -The

access port door was operated by s i x clamps

which r e l i e d on pneumatic operators and a spring f o r opening and a scheduled time-delay of these pneumatic operators f o r closure ;. however, once closed, t h e mechanical linkage was such t h a t t h e clamps no longer require gas pressure t o keep t h e door t i g h t l y closed.

The a b i l i t y of two concen-

t r i c neoprene gaskets (between t h e access port door and housing) t o contain helium at 55 p s i a s a t i s f i e d an interlock and indicated t h a t a gas-tight s e a l did indeed e x i s t between areas 1 C and 3A.

On s e v e r a l occasions during

t h e p r e - c r i t i c a l run, t h e access port f a i l e d t o close properly.

Also, one

o r two of t h e s i x access port operators f a i l e d t o open and t h e manipulator

was used t o release t h e s t i c k i n g operators.

During t h e shutdown following

t h e low-power experiment, t h e clamps were readjusted and an extension was added t o t h e pin of each clamp; t h i s extension made it e a s i e r t o open any clamp manually with t h e manipulator i n t h e event of a pneumatic malfunction. Flexible Containment Membranes

- There

are two 0.020-in.

thick flexi-

b l e urethane membranes (boots) which p a r t i a l l y enclosed and provided mob i l i t y t o t h e manipulator arm used i n t r a n s f e r r i n g capsules t o and from t h e l a t c h i n area1C and t h e bottom portion of t h e t r a n s p o r t c a r r i e r i n a r e a

3. A vacuum w a s maintained between these boots t o monitor t h e i n t e g r i t y of t h e boots and assure t h a t double containment did indeed exist between t h e sampler operator and t h e sample.

A loss of vacuum and a pressure re-

duction i n area 3 A would indicate a leak i n t h e outer boot whereas a loss of vacuum with no pressure change i n a r e a 3A would implicate t h e inner boot (atmosphere s i d e ) o r possibly both boots. There were t h r e e boot f a i l u r e s during t h e first 140 sampling cycles

of Run 2.

One w a s caused by an inadvertent evacuation of area 3A while

the manipulator cover was off; t h e r e s u l t a n t pressure gradient ruptured the boot.

On another occasion, t h e boot was snagged on t h e bottom piece

LiJ

299

t-,

of t h e transport tube i n area 3A when t h e manipulator was used t o release a stuck access port clamp.

The t h i r d f a i l u r e r e s u l t e d from pinching t h e

boot between t h e manipulator arm and t h e housing. A t t h e end of t h e next run ( R u n 3 ) , a pressure switch ( w i t h an alarm

and i n t e r l o c k ) was i n s t a l l e d t o detect and prevent a negative pressure gradient g r e a t e r than 1/2 p s i between area 3A and t h e manipulator arm. S t e e l rings were i n s t a l l e d i n t h e convolutions of t h e inner boot t o hold

it f r e e of t h e manipulator a r m and possibly prevent t h e pinching-type failure.

Also t o ensure t h a t at l e a s t two b a r r i e r s e x i s t e d between primary con-

tainment and sampling personnel, a pressure switch was i n s t a l l e d on t h e manipulator cover which required a negative pressure of 4-in.

Hg i n the

cover before e i t h e r the operational o r maintenance valve could be opened. I n April 1966, t h e manipulator cover pressure was inadvertently evacuated t o -25-in.

Hg without lowering area 3A pressure simultaneously; t h e

12-psi d i f f e r e n t i a l caused a s m a l l puncture of t h e inner boot.

This s e t

of boots had been used f o r 48 sampling cycles since power operation was begun and t h e maximum power reached a t t h a t time was 2.5 MW.

The r a d i a t i o n

l e v e l of t h e manipulator assembly and boots w a s 10 R/hr at 3 inches from t h e f i n g e r s , but was reduced by a f a c t o r of 10 by scrubbing w i t h soap and water.

About 2 hours were required f o r replacement of both boots.

Deformation of t h e Manipulator A r m and Fingers

- During Run 2 ,

the

manipulator arm and fingers were bent which caused d i f f i c u l t y i n gripping the l a t c h cable and moving t h e manipulator arm.

a.rm was replaced and a

A t t h e end of Run 3, t h e

l / 4 x 114 inch projection was welded t o t h e bottom

of each finger t o aid i n grasping t h e capsule cable from thâ&#x20AC;&#x2122;e f l o o r of area

3A.

The manipulator a r m w a s replaced and t h e clearances between t h e arm

and c a s t l e j o i n t were increased t o reduce t h e force required t o operate the manipulator.

The manipulator and fingers operated s a t i s f a c t o r i l y f o r 421 additional sampling cycles until A u g u s t 1968 at which time t h e manipulator assembly

was replaced because t h e t i p

t h e fingers no longer closed t i g h t l y .

old assembly w a s decontaminated and repaired.

The

The replacement fingers op-

e r a t e d s a t i s f a c t o r i l y u n t i l September 1969 when t h e double metal bellows

300

developed a leak, The sampler on t h e f u e l processing ystem t 5 cannibal i z e d t o make t h e r e p a i r s ; t h i s u n i t operated s a t i s f a c t o r i l y f o r t h e subsequent three months of reactor operation before shutdown. Capsule Recovery from t h e Operational Valve

-On

one occasion during

t h e p r e c r i t i c a l run, an empty sampling capsule w a s accidentally knocked i n t o area 1 C before t h e l a t c h pin was completely engaged i n t h e l a t c h , and the capsule dropped onto t h e gate of t h e operational valve,

The capsule

was r e t r i e v e d by removing t h e manipulator assembly from 3, opening t h e access p o r t , and snaring t h e capsule cable w i t h a wire hook., Since t h i s type of accident could reoccur, t h e brass l a t c h pins used t o a t t a c h t h e capsule t o t h e l a t c h w e r e replaced with nickel-plated mild s t e e l pins so t h a t capsule recovery could then be effected with a magnet. After approximately a month of full-power operation i n July 1966, an empty capsule again was dropped onto t h e gate of t h e operational valve when t h e manipulator slipped during the l a t c h i n g operation.

A magnet w a s l o w -

ered i n t o t h e t r a n s f e r l i n e and the capsule assembly was recovered as t h e magnet w a s withdrawn.

The radiation l e v e l of t h e magnet a f t e r t h e recovery

operation was 10 R/hr at 2 ft.

(pi

A l l sampling capsules of t h e bucket-type assembled since t h e l a t t e r part of 1966 contained a nickel-plated mild s t e e l top so t h a t capsule re-

covery could be made with a magnet. Removal Area, Valve, and Seals

- During t h e p r e - c r i t i c a l

run, t h e

removal s e a l and valve required realignment and increased tolerances bef o r e t h e transport container and removal t o o l assembly would s l i d e through these u n i t s f r e e l y without binding.

I n addition, t h e removal valve f a i l e d

t o s e a l properly even though t h e b a l l and seals were replaced.

Therefore,

at t h e end of t h e next run, t h e valve assembly was replaced with a modified version t o improve t h e s e a l i n g c h a r a c t e r i s t i c s and improve f u t u r e access t o t h e valve.

This u n i t served s a t i s f a c t o r i l y except t h a t a b u f f e r gas

leak developed i n t h e top t e f l o n s e a l of t h e valve during t h e l a t t e r p a r t of 1967.

A s a consequence of t h e leak, t h e equilibrium b u f f e r s e a l pres-

sure reached only 30-35 p s i a instead of t h e normal 5 0 5 5 p s i a ; t h e associated pressure switch which permits t h e opening of t h e operational and maintenance valves and t h e access port was lowered from 50 t o 30 and f i n a l l y i n December 1967 t o 25 p s i a where it remained f o r t h e remainder of r e a c t o r operations

301

Latch, Capsule, and Cable Malfunctions

- Early i n Run 2, t h e cable

drive motor s t a l l e d during t h e withdrawal of an empty capsule.

The cap-

sule was r e i n s e r t e d about 12 i n . and then withdrawn successfully. reason f o r t h e malfunction,was unknown.

The

Approximately 100 sampling cycles

l a t e r i n December 1965, t h e motor s t a l l e d again w h i l e withdrawing a 10-gm sample.

Af'ter repeated attempts, t h e capsule was withdrawn completely and

t h e cable was examined.

The capsule was empty and t h e cable w a s found t o It

be backed up i n t o t h e drive u n i t box and caught i n t h e motor gears.

was assumed t h a t t h e l a t c h had hung on t h e gate of t h e operational o r maintenance valve causing t h e cable t o c o i l up i n s i d e a r e a 1 C . The cable was straightened, t h e l i m i t switches on t h e operational and maintenance valves were reset t o open t h e valves wider and t h e diameter o f t h e l a t c h was reduced.

I n April 1966, after approximately a week a t a power l e v e l of 5 MW, t h e capsule drive-unit motor s t a l l e d as t h e l a t c h w a s being r e t r i e v e d through t h e pipe bend near t h e pump bowl.

d r a w a l s , t h e l a t c h was retrieved.

After several i n s e r t s and with-

The l a t c h and p a r t of t h e cable were

then pulled through t h e access p o r t , t h e removal valve, and i n t o t h e sample transport cask which provided shielding while t h e l a t c h was replaced with one designed t o provide an additional 1/8-inch diametrical clearance, Approximately 3 months l a t e r t h e capsule stuck again temporarily f o r some unexplained reason on withdrawal and no sample w a s obtained.

Subse-

quent hangups with more serious consequences a r e discussed i n Section 18.1.3. -While

t e s t i n g t h e new l a t c h ,

t h e cable position indicator stopped when t h e l a t c h was approximately two

feet from t h e pump bowl; t h e upper limit switch a l s o actuated at t h i s time. E l e c t r i c a l continuity chec d r a w windings of t h e cab1

motor and a l s o t o t h e

All these leads penetrate

ontainment w a l l through

ptacle.

Since r e p a i r c

without v i o l a t i n g contai l a t c h were then pulled i transport container and

owed open c i r c u i t s t o t h e i n s e r t and with-

er l i m i t switch. ommon 8-pin re-

t be made while f u e l was i n t h e r e a c t o r he r e a c t o r w a s drained.

The cable and

s i n g t h e manipulator t o p u l l and t h e

rt operators t o hofd t h e cable so t h a t

t h e cable could be regrasped with t h e manipulators f o r another p u l l .

The

302

operational and maintenance valves were then closed.

The assembly includ-

ing t h e drive motor, cable, and l a t c h were removed, p a r t i a l l y decontaminated, and repaired with t h e use of shadow shielding techniques,

(Three

connector pins i n one of t h e 8-pin receptacles were burned o f f , )

A l l six

similar receptacles at t h i s location were f i l l e d with epoxy r e s i n t o provide additional i n s u l a t i o n and strength,

The cable which w a s bent i n sev-

eral places during r e t r i e v a l w a s decontaminated with soap and water t o reduce t h e radiation l e v e l t o 5 R/hr a t 3 i n . before it was straightened. A f t e r reassembly and checkout, normal sampling was resumed. Contamination of t h e Removal Seal - A n

a r e a a t t h e t o p of t h e sampler

was provided s o t h a t t h e lower p a r t of t h e t r a n s p o r t c a r r i e r and capsule could be evacuated and pressurized several times t o remove oxygen and moist u r e from t h e capsule and c a r r i e r before i n s e r t i o n i n t o t h e dry box (Area 3A) of t h e sampler.

The bottom s e a l f o r t h e removal a r e a w a s provided by

t h e removal valve and a s e a l at t h e t o p was made when t h e transport tube was i n s e r t e d through a s e t of O-rings at t h e t o p of t h e removal volume. During sampling the shipping cask w a s aligned and placed over t h e removal a r e a e A s m a l l amount of vacuum grease was smeared onto t h e lower p a r t of t h e transport tube before it w a s lowered through t h e shipping cask and i n t o the O-rings of t h e removal seal,

When t h e t r a n s p o r t tube was f u r t h e r lo-

wered i n t o Area 3A (a contaminated a r e a ) t o i n s e r t an empty capsule and again t o pick up t h e sample, t h e outer surfaces of t h e transport tube and removal t o o l became contaminated with s o l i d f i s s i o n products from t h e cap-

sule, f l o o r of Area 3, and/or manipulator,

Some of these p a r t i c l e s were

wiped o f f by t h e O-rings of t h e removal s e a l as t h e removal t o o l and transport tube were withdrawn from Area 3A; t h a t remaining on t h e removal t o o l

was wiped o f f with a damp cloth.

The shipping cask w a s gradually c o n t a d -

nated as both t h e removal t o o l and t r a n s p o r t tube were drawn i n t o it,

By t h i s process t h e removal seal becomes progressively more contambated with each sample,

During p r e - c r i t i c a l operations t h i s t r a n s f e r mechanism f o r

p a r t i c u l a t e matter was not obvious; however, after power operations had begun and especially after t h e l a t c h maintenance work, where t h e l a t c h was manually r e t r i e v e d from t h e v i c i n i t y of t h e pump and out through t h e removal area f o r r e p a i r , t h e removal area and t h e top of t h e sampler became contaminated.

303

The adoption of t h e following procedure proved e f f e c t i v e i n preventing t h e spread of contamination during sample removal and transport throughout t h e remainder of reactor operations: (1) The top of t h e sampler w a s established as a Contamination Zone. (2) The removal t o o l was wiped w i t h a damp c l o t h during w i t h d r a w a l

from t h e removal area.

( 3 ) The shipping cask was wiped with a damp cloth p r i o r t o enclosure i n a p l a s t i c bag f o r shipment t o t h e a n a l y t i c a l laboratory.

(4)

The t o p of the sampler was wiped w i t h a damp cloth a f t e r each

sample.

( 5 ) An exhaust hood was erected near t h e removal area and p a r t i a l l y enclosed t h e shipping cask t o maintain a s l i g h t l y reduced pressure i n t h i s area s o t h a t any airborne p a r t i c l e s would be drawn i n t o t h e f i l t e r and exhaust system of t h e building.

(6) The removal seal area was cleaned periodically Control C i r c u i t r y Changes -During

w i t h damp wipes.

t h i s period f i v e changes were made

t o t h e sampler control c i r c u i t s : (1) A pressure switch was added t o prevent evacuation of Area 3A t o

more than 10 i n . of water g r e a t e r than t h e manipulator cover area.

(2) An i n t e r l o c k was added t o require t h a t both t h e operational and maintenance valves be closed before the access port can be opened.

( 3 ) A permissive switch and l i g h t were i n s t a l l e d t o indicate t h a t 1 C pressure i s equal t o o r

less than Area 3A pressur (4) A fuse w a s add

sule drive motor c i r c u i t t o protect

les from excessive currents. oss t h e two motor windings

t o l i m i t any high voltage pe

i n g s t a r t i n g and stopping. -During

t h e low-power experi-

an opening near t h e t o p and capable of nto t h e pump bowl but f o t long enough t o permit Subsequent longer assembl i e

obtained s a l t except one which was believed

t o have hung on t h e l a t c h stop at t h e pump bowl and did not e n t e r t h e pump. The capsules were subsequently modified t o allow them t o hang s t r a i g h t .

304 â&#x20AC;&#x2DC;Most of t h e samples taken a f t e r the reactor exceeded 1 MW (nuclear power) were highly radioactive and exceeded 1000 R/hr at 3 i n .

Fission

products adhering t o t h e outside of t h e capsules contaminated t h e bottom cups of t h e transport containers.

For a l i m i t e d time disposable p l a s t i c

l i n e r s i n s e r t e d i n the cups were e f f e c t i v e i n reducing t h e radiation l e v e l

i n t h e bottom portion of t h e transport tubes.

However, as more f i s s i o n

products b u i l t i n t h e f u e l and t h e i n t e r i o r of t h e sampler became more contaminated, it w a s more economical t o f a b r i c a t e disposable bottom cups of mild s t e e l r a t h e r than decontaminate and reuse t h e s t a i n l e s s s t e e l cups, The new cups were s h o r t e r than t h e previous ones and t h e p l a s t i c l i n e r w a s

used t o confine t h e capsule, wire, and l a t c h pin so t h a t t h e top p a r t of t h e transport c a r r i e r could be more e a s i l y slipped over t h i s combination t o engage the threads and double O-ring s e a l s at t h e bottom of t h e mild s t e e l cups. On one occasion while the t o p of t h e transport container was being mated with t h e bottom which contained a 50-g capsule, t h e wire on t h e capsule caught i n t h e threads of t h e mating pieces and galled before t h e two

pieces were sealed completely.

Thereafter t h e l i n e r s were lengthened t o

accommodate t h e t h e 50-g capsules. I n J u l y 1966, t h e f u e l pump w a s accidentally o v e r f i l l e d w i t h flush

salt and salt w a s pushed up i n t o t h e sampler tube approximately two f e e t above t h e f u e l pump where it froze and thus prevented t h e capsule and l a t c h from being lowered i n t o the pump bowl.

The salt w a s melted and allowed t o

drain back i n t o t h e pump by energizing a set of remotely-placed heaters i

around t h e sampler tube and t h e normal heaters surrounding t h e pump bowl. Normal sampling w a s subsequently resumed. Contamination of t h e area sGrounding t h e sampler was minimized during maintenance by designating t h e area surrounding t h e sampler a Contaminat i o n Control Zone and at times-by erecting a p l a s t i c t e n t around t h e sampler.

Repair work was done inside t h e zone with appropriate protective

clothing, gloves, and shoe scuffs.

Contamination w a s found outside t h e

zone twice and on both occasions it was a t t r i b u t e d t o contaminated shoes which indicates that t h i s type of contamination i s not r e a d i l y airborne.

305

18.1.3

Subsequent Operating Experience with t h e Sampler-Enricher

The major sampler problems encountered during t h e remainder of operat i o n s were caused by t h e l a t c h and/or capsule lodging at t h e entrance t o t h e sampler tube o r at one of t h e i s o l a t i o n valves immediately below Area 1 C with subsequent unreeling and tangling of' t h e cable i n Area 1 C and i n

t h e gears of t h e drive unit. Ruptures o r pin-hole leaks i n t h e boots enclosing t h e manipulator arm were a chronic source of trouble requiring periodic replacement.

Life of

t h e boots varied widely but averaged 44 sample cycles before failure.

The

cause of most of these leaks was undetermined because t h e boots could not be examined e a s i l y due t o t h e high radiation l e v e l of t h e manipulator assembly. Gradual d e t e r i o r a t i o n of t h e buffer gas s e a l s at t h e operational, maintenance and removal valves and a t t h e access port continued and on occasions t h e time required t o take a sample was lengthened because a longer time i n t e r v a l was required f o r t h e buffer pressure t o b u i l d up suff

f i c i e n t l y t o actuate t h e proper r e l a y f o r t h e next s t e p i n t h e procedure. Loss of Capsules i n t h e Pump Bowl - I n

August 1967 during what ap-

peared t o be a normal sampling operation, t h e capsule and/or l a t c h had apparently lodged at t h e maintenance valve while being lowered, causing t h e cable t o c o i l up i n Area 1 C and i n t o t h e motor drive area where it had become snarled and kinked.

The f i r s t indication of a malfunction was during

t h e capsule r e t r i e v a l cycle when t h e motor s t a l l e d w i t h

off the

8 f t 3 i n . of cable

Repeated i n s e r t s and withdrawals recovered only 5 more in.

of cable before t h e motor s t a l l e d w i t h only an inch of t r a v e l i n e i t h e r direction ( t h i s had happene

efore i n 1965 but t h e cable and capsule

were.retrieved completely If t h e capsule and c

a$ t h e sampler tube entrance be-

low Area lC, then t h e operational and maintenance valves could be closed and sampler r e p a i r s could

roceed without a r e a c t o r d r a i n ; otherwise a

drain would be necessary.

After consideration of several p o s s i b i l i t i e s , it

was decided t o disengage t h e motor drive from t h e operational valve and close t h e valve slowly with t h e hand wheel with t h e idea t h a t i f t h e cap-

sule were below t h e valve, t h e cable would o f f e r enough r e s i s t a n c e t o clo-

sure t o be detected i n which case t h e valve would be reopened and t h e

306 reactor drained before repair. was p r a c t i c a l l y closed,

No resistance w a s detected u n t i l t h e valve

Therefore, both valves were closed and when Area

1C was opened, as expected, t h e cable was coiled up within; however t h e l a t c h , capsule, and about 6 i n . of cable were missing.

The r e a c t o r was

then drained and grappling t o o l s were fabricated of f l e x i b l e tubing and cables which could be i n s e r t e d through t h e sampler i n t o t h e pump bowl, Four of t h e f i v e t o o l s which were most successful at grasping a dummy l a t c h a r e shown i n Fig. 18.3. When t h e noose t o o l was lowered, it snared t h e l a t c h but t h e 1/32-inO s t e e l cable broke because t h e l a t c h was apparently glued i n place at t h e l a t c h stop,

with salt

The dislodging t o o l w a s then used but apparently t o no

a v a i l because a second noose grasped t h e l a t c h but could not budge it with a p u l l of 25 l b .

The pump w a s then heated u n t i l t h e l a t c h was at a tem-

perature of 700째F.

It was then l i f t e d several f e e t without d i f f i c u l t y be-

fore it lodged i n t h e tube, apparently at t h e expansion j o i n t between t h e f u e l pump and t h e sampler.

A corkscrew t o o l w a s next i n s e r t e d but t h e

l a t c h again hung on r e t r i e v a l , t h i s time a t ' t h e first bend near t h e f u e l pump. The l a t c h r e t r i e v a l t o o l which w a s designed t o f i t over t h e severed cable and s t e m and thus prevent t h e cable from jamming against t h e w a l l during r e t r i e v a l w a s successful i n p u l l i n g up t h e cable, l a t c h , and l a t c h pin,

The capsule apparently snapped i t s cable and dropped t o t h e bottom

of t h e sampler cage when t h e l a t c h came t o a sudden s t o p at t h e l a t c h stop. The dotted o u t l i n e of a capsule. i n Fig. 18.4 shows how a capsule, once detached from i t s cable, could s l i p out of t h e sampler cage and then be trapped by t h e mist s h i e l d i n t h e pump bowl. Since t h e capsule t o p i s made of n i c k e l p l a t e d mild S t e e l , a magnet i n combination with a go-gage ( F i g . 18.3) was lowered i n t o t h e pump bowl t o veri* sule.

t h a t t h e sampling tube w a s unobstructed and t o pick up t h e cap-

Capsule recovery w a s unsuccessful and since no adverse chemical o r

mechanical e f f e c t s were envisioned should t h e capsule be allowed t o remain i n t h e mist s h i e l d , f u r t h e r r e t r i e v a l attempts Were abandoned. While r e t r i e v a l e f f o r t s were i n progress, a replacement 1 - C assembly which was on hand, w a s equipped with a l a t c h made of 430 s t a i n l e s s s t e e l and with a sleeve t o prevent f'uture tangling of t h e cable i n t h e gears of the drive unit.

However, subsequent examination o f t h e o l d drive u n i t i n

a;l

307

ORNL-OWG

6?-44?08

1 3 2 -in. CABLE

CABLE SHEATH

CABLE BEING PUSHED OUT

NOOSE READY FOR FISHING DISLODGING TOOL

NOOSE TOOL

#'/46-in.

CABLE

412 in. SCHED 40 PIPE

1.375 in. DlAM

LATCH RETRIEVAL

Fig. 18.3

Tools De

GO- GAGE AND CAPSULE RETRIEVAL TOOL

Retrieval of Latch and Capsule from

308

ORNL- DWG 67-10766

OFF

' i,

LATCH ASSEMBLY NORMAL POSITION

LATCH STOP,

TOP OF FUEL PUMP

AMPLE CAPSULE SAMPLE CAPSULE

SAMPLE CAPSULE

SAMPLE CAPSULE LOST SAMPLE CAPSULE

INCHES

Fig. 18.4

Location of Sampler Latch and Capsules i n Fuel Pump Bowl

309

hii

a hot c e l l revealed t h a t t h e cable was not tangled i n t h e gears as had been assumed; t h e f a i l u r e t o withdraw o r i n s e r t was due t o a sharp k i n k i n t h e cable which became lodged i n t h e 17/32-in.

diameter channel of t h e

l a t c h positioner. When normal operations and sampling were resumed, a commercially available device was t e s t e d which was capable of sensing t h e l a t c h (now made of magnetic material) inside t h e sampler tube.

A design was prepared t o mount

t h i s proximity switch about 10 i n . below t h e maintenance valve t o detect

the l a t c h as it i s lowered during normal sampling; however, before it could be i n s t a l l e d , t h e l a t c h again lodged i n t h e sampler tube without detection

i n March 1968.

The usual 17 f t 5 in. of cable was reeled o f f and w a s tan-

gled i n Area 1C. the reel.

A s it was rewound, t h e motor s t a l l e d w i t h 13 f t 5 i n . o f f

When repeated withdrawal attempts f a i l e d , t h e r e a c t o r was drained

and the capsule access port was opened.

The capsule was v i s i b l e through t h e

lower corner of t h e opening and t h e l a t c h could be seen i n t h e back of t h e chamber w i t h many loops and c o i l s , some of which appeared t o extend down i n t o t h e sampler tube.

Rather than r i s k c u t t i n g t h e cable, t h e i s o l a t i o n

valves were l e f t open; however, when an attempt was made t o lift t h e caps u l e out of t h e chamber, t h e manipulator brushed some of t h e c o i l s causing them t o spring out through t h e port and p u l l t h e capsule wire from t h e manipulator fingers.

The capsule dropped i n t o t h e sampler tube and t h e l a t c h

tipped over, thus releasing t h e key; t h e capsule and key then disappeared down t h e sampler tube i n what appeared t o be a bottom-up position. A f t e r several unsuccessful attempts were made t o r e t r i e v e t h e capsule

by lowering magnets of various s i z e s i n t o t h e pump bowl, t h e reactor was f i l l e d w i t h f l u s h salt.

A 50-g capsule ( 5 in. long w i t h an opening 4-1/8

i n . from t h e bottom of t h e

apsule) w a s lowered i n t o the. pump but came up

empty; salt droplets clinging t o t h e outside indicated t h a t t h e capsule had been only h a l f submerged.

Later a 10-g capsule d i d c o l l e c t a sample.

I n t h e meantime a f u l tube, mist s h i e l d , and s

e p l a s t i c and metal mock-up of t h e sampler age was constructed and used t o s e l e c t t h e

b e s t t o o l s and magnets which could be lowered v i a t h e sampler i n another attempt t o r e t r i e v e t h e capsule.54 proved t o be 1 / 2 and 3/4-in.

The simplest and most e f f e c t i v e t o o l

diameter Alnico-5 magnets.

When t h e f l u s h

salt was drained and r e t r i e v a l e f f o r t s resumed, sounds from a contact

- I

310 microbhone on t h e f i e 1 pump indicated t h a t t h e 1/2-in.

an object a few inches and then dropped it.

d i a , magnet l i f t e d

A f t e r mhny attempts, t h e mag-

net came up with an object which l a t e r proved t o be t h e corroded top of t h e old capsule.

Further e f f o r t s with e i t h e r magnet were unsuccessful and t h e

second capsule w a s abandoned. During t h i s shutdown, t h e proximity switch was i n s t a l l e d about 4 i n . below t h e maintenance valve and t h e sampling procedure w a s modified t o s t i p u l a t e t h a t if t h e switch is not actuated when t h e capsule position indicator shows t h a t s u f f i c i e n t cable has been r e e l e d o f f t o reach t h e proximity switch, t h e drive w i l l be stopped.

Also a vakiac was added t o t h e

drive motor c i r c u i t so that t h e operating voltage could be changed t o 80 v o l t s t o lessen t h e p o s s i b i l i t y of seriously kinking t h e cable, During t h e next s t a r t u p , with f l u s h salt i n t h e r e a c t o r , salt w a s trapped i n a 10-g capsule but none trapped i n a 50-g capsule which requires

an immersion depth of 4-1/8 i n . t o t r a p a sample.

However, a 30-g capsule

requiring only 2-1/2 i n . of immersion did t r a p a sample.

These results

agree with t h e measurements taken during capsule r e t r i e v a l which indicated that t h e abandoned capsule w a s 1-3/4 in. above t h e bottom of t h e cage.

However, with f u e l i n the r e a c t o r , t h e a c t u a l l e v e l i n t h e pump bowl was somewhat higher than t h a t indicated because of t h e high void f r a c t i o n of t h e l i q u i d i n t h e pump bowl.

Consequently no problems were encountered

taking 50-g f u e l samples throughout t h e remainder of operations except during the l a t t e r p a r t of 1969 when t h e f u e l pump speed was lowered t o

980 rpm.

A t t h i s speed the f u e l salt void f r a c t i o n i s approximately equiv-

a l e n t t o that f o r f l u s h salt at normal pump speed.

The failure t o t r a p a

50-g sample under these conditions indicated t h a t t h e position of t h e abandoned capsules had not changed after approximately one year of operation, I n October 1969, a serious cable tangling problem w a s averted by t h e use of t h e proximity switcho When t h e switch f a i l e d t o actuate a f t e r 3 f t 10 i n , of cable was paid out during an otherwise normal capsule i n s e r t , t h e cable w a s r e t r i e v e d u n t i l t h e position i n d i c a t o r indicated t h a t a l l of t h e cable was back on t h e takeup reel and t h e capsule was f u l l y withdrawn i n t o i s o l a t i o n chamber 1 C .

The operational and maintenance valves were then

closed and t h e access port was opened.

The cable was indeed fully re-

t r i e v e d as indicated and t h e capsule was f u l l y v i s i b l e but was lodged

311

diagonally between t h e ledges of t h e doorway t o t h e chamber.

Apparently

t h e capsule had lodged at t h e sampler tube entrance o r a t one of t h e isolat i o n valves and as t h e cable w a s paid out, it somehow encircled t h e capsule because t h e capsule w a s l i f t e d and relodged a t a higher position (during cable r e t r i e v a l ) than it w a s at t h e s t a r t of t h e sampling procedure.

The

capsule was then lowered t o i t s normal position and a l l subsequent cable operations proceeded without interruptions. Wiring Fault

- I n November of 1967 as a sample w a s being withdrawn,

the 0.3-amp fuse i n t h e drive motor f a i l e d .

Subsequent t e s t s showed t h a t

t h e i n s e r t mode was normal at 0.2 amp but during w i t h d r a w a l , t h e current increased t o 1.0 o r 1.5 amp, indicating leakage'to ground.

During f u r t h e r

t e s t s , t h r e e c i r c u i t s opened, disabling t h e i n s e r t and w i t h d r a w c i r c u i t s and t h e upper l i m i t switch.

After t h e erection of a t e n t and other meas-

ures t o prevent t h e spread of contamination, a 3-in. dim hole w a s sawed

i n t h e cover p l a t e of Area 3A d i r e c t l y above t h e cable penetration i n t o the inner box (Area 1C).

The f a i l u r e was found where t h e wires were bent

back against t h e s i d e of t h e plug on t h e lower end of t h e cable between t h e inner (1C) and outer (3)containment boxes.

When a temporary con-

nection indicated t h a t everything within Area 1C was operable, t h e sample was r e t r i e v e d and t h e i s o l a t i o n valves were closed.

The damaged section

was abandoned and a new cable was i n s t a l l e d having a penetration through a 4-in. -

pipe cap welded over t h e sawed hole i n t h e t o p cover.

While t h e pipe cap w a s being welded i n place, a heavy current evi-

dently went t o ground through an adjacent penetration, destroying t h e plug and receptacle. tration.

Repairs we

As a result o f t h

was added t o each of t h e t h

made by c u t t i n g out and replacing t h i s peneexperience, an i s o l a t i o n transformer and fuse drive u n i t cables which allows one ground

without interference with operation.

After t h e r e p a i r s , a l l c i r c u i t s were

operable except t h e upper l i m i t switch which stops t h e drive motor when t h e l a t c h reaches t h e l a t c h stop.

Since t h e motor and c i r c u i t can e a s i l y w i t h -

stand blocked-rotor current, t h e upper l i m i t switch w a s bypassed and therea f t e r t h e motor w a s turned o f f when t h e position indicator indicated t h a t the l a t c h was f'ully withdrawn. ..

'RepalP 'df Cable Drive Gear

- I n December 1968 during an

attempt t o

remove a capsule from t h e l a t c h , it w a s found t h a t t h e cable could be

312 pulled o f f t h e r e e l while t h e drive motor w a s stationary.

The t e n t a t i v e

diagnosis was t h a t one of t h e p a i r of drive gears was s l i p p i n g on i t s shaft.

I n order t o gain access t o t h e drive u n i t , it w a s necessary t o re-

move t h e s h i e l d blocks over t h e reactor c e l l which, i n t u r n , required t h a t t h e f u e l be drained.

A f t e r a temporary containment enclosure and contami-

nation zone had been s e t up around t h e sampler, t h e containment box, lC, containing t h e drive u n i t was disconnected, l i f t e d , and a 3-in. hole sawed through t h e s i d e adjacent t o t h e gears.

One gear w a s found t o be loose be-

cause i t s two setscrews had come loose.

The gear was repositioned and tack-

welded on i t s shaft and t h e other gear w a s fastened with jam screws on t o p of i t s setscrews. stalled,

A patch w a s then welded on t h e box and t h e u n i t rein-

The work was done without excessive exposure of personnel despite

t h e high radiation l e v e l s (10 R/hr at 12 i n . from t h e box) by use of shielding and extended t o o l s designed e s p e c i a l l y f o r t h e job. Since other parts of t h e sampler were e a s i l y accessible during t h i s shutdown, t h e manipulator hand was repaired, t h e illuminator port and viewing window l e n s were replaced because of discoloration by r a d i a t i o n , and an imperfect s e a l i n the removal valve was replaced. Vacuum F+mp Problems

- The sampler was

equipped with two vacuum pumps.

Vacuum pump No. l w a s used t o evacuate contaminated areas such as containment areas 1C.and 3A and discharged i n t o t h e a u x i l i a r y charcoal bed; vacuum pump No. 2 was used t o evacuate non-contaminated areas such as t h e manipul a t o r cover and t h e plenum between t h e manipulator boots and discharged int o t h e containment air system f i l t e r s and t o t h e stack, During t h e July 1966 shutdown, t h e o i l l e v e l s i n t h e pumps were checked and a small amount of o i l w a s added. I n January 1969 t h e r e w a s an a c t i v i t y r e l e a s e t o t h e stack (~0.08mc) which was a t t r i b u t e d t o a gas leak around t h e s h a f t of vacuum pump No. 1. The pump w a s replaced because t h e o i l s e a l at t h e s h a f t could not be repaired without extensive decontamination procedures,

Four months l a t e r ,

repeated vacuum pump motor outages (due t o overload) were apparently caused by low o i l level.

On one inspection, t h e pump w a s p r a c t i c a l l y empty of o i l

with no evidence of external o i l leaks and was consequently r e f i l l e d t o t h e proper level.

On t h e next inspection, t h e pump contained approximately

crii

313

2 q t s more than t h e normal inventory.

drained back i n t o t h e pump.

Apparently t h e previously l o s t o i l

High pressure (>13p s i a ) i n Area 1C w a s nor-

mally vented i n t o t h e auxiliary charcoal bed through a l i n e bypassing t h e pump.

Apparently high pressure w a s vented through t h e vacuum pump instead,

thus carrying over a l a r g e portion of t h e o i l i n t o t h e holdup tank and offgas l i n e above t h e pump discharge.

After t h e o i l l e v e l w a s readjusted,

pump operation returned t o normal and t h e l a s t o i l check w a s normal. Heated Shipping Cask

-A t

temperatures l e s s than 400째F, radiation

from f i s s i o n products can produce f r e e f l u o r i n e i n frozen f u e l s a l t .

minimize f l u o r i n e production and thus avoid t h e e f f e c t s of fluorine on t h e r e s u l t s of oxide and t r i v a l e n t uranium analyses, a shipping cask w a s fabricated which maintained t h e f u e l sample at approximately 500'F

from t h e time

it was removed from t h e sampler u n t i l it was unloaded a t t h e a n a l y t i c a l laboratory.

The cask used molten b a b b i t t both as shielding and as a heat

reservoir.

Built-in e l e c t r i c heaters melted t h e babbitt before t h e sample

was loaded and t h e heat of fusion maintained t h e sample a t a r e l a t i v e l y constant temperature f o r about 9 hours during shipment and storage.

For

t h e elevated temperature, O-rings made of Viton A were s u b s t i t u t e d f o r t h e standard neoprene O-rings i n t h e transport tube.

I n addition t o t h e double

O-ring seal provided by t h e transport tube, t h e transport tube w a s capped inside t h e cask during shipment. A l l samples were unavoidably cooled t o room temperature i n t h e sampler

where they were sealed i n t h e transport tube a t atmospheric pressure.

When

t h e transport tube was i n s e r t e d i n t o t h e heated cask, t h e pressure i n t h e tube increased t o approximately 10 p s i g because of t h e temperature increase. On one occasion when t h e cap w a s removed from t h e cask, airborne a c t i v i t y was released from t h e cask cavity i n t o t h e v e n t i l a t i o n system of t h e unloading s t a t i o n at t h e a n a l y t i c a l laboratory.

Subsequently t h e s e a l i n g

surfaces of t h e t o p p a r t of t h e transport tube were found t o have been eroded by repeated decontamination procedures t o such an extent t h a t t h e double O-ring s e a l s were i n e f f e c t i v e .

After examination of t h e sealing

surfaces of t h e remaining f i v e transport tube tops and reevaluation of t h e decontamination c o s t s , it w a s decided t o discard t h e reusable s t a i n l e s s s t e e l transport tubes and use "throw-away" transport tubes made of mild

tyj

314

s t e e l f o r all subsequent samples.

A valve was l a t e r added t o t h e cap of

t h e heated cask so t h a t any future pressure buildup i n t h e cask cavity could %e released t o a charcoal f i l t e r before cap removal. Other Capsules and Samples -The

s i z e of t h e capsules t h a t could be

accommodated by t h e sampler was l i m i t e d t o 0.75 inches i n diameter and

6-1/4 inches i n length.

Approximately 1 / 4 inch g r e a t e r diameter could be

accommodated by t h e sampler but not by t h e transport tube.

Four of t h e

various types of capsules which were lowered i n t o t h e pump bawl a r e shown i n Figures 1 8 . 5 ~through ~ 18.5d, e i t h e r salt 'or gas,

The capsule i n Fig. 18,5a was used t o t r a p

The capsule was first evacuated and sealed with a mea-

sured amount of salt similar t o t h e fuel.

On entry i n t o t h e f u e l pump, t h e

s e a l i n g salt melted and salt o r gas (depending on t h e capsule elevation i n t h e pump bowl J u s t p r i o r t o salt l i q u e f a c t i o n ) was drawn i n t o t h e capsule, When t h e beryllium addition capsules were withdrawn, they were encrusted with s o l i d s as sham i n Figure 18.5b.

A s a r e s u l t , t h e capsules

were d i f f i c u l t t o load i n t o t h e bottom p a r t of t h e t r a n s p o r t tubes even though t h e p l a s t i c l i n e r was removed.

Also t h e enriching capsules were

sometimes d i f f i c u l t t o load because of t h e i r length (6-1/4 i n * ) * The ca-

b l e s of these long capsules were occasionally caught i n t h e threads of t h e mating parts of t h e transport container and were d i f f i c u l t t o unload at t h e hot c e l l o The r a d i a t i o n levels of t h e beryllium capsules and t h e empty enriching capsules were more than four times higher than t h e standard 10-g sample <

when delivered t o t h e a n a l y t i c a l laboratory, .-

Miscellaneous Sampler Problems

- During t h e l a t t e r p a r t

of 1966, a

c o t t e r key had come out of t h e t o p of t h e lower hinge p i n on t h e access port door.

The hinge pin had then worked out of t h e t o p of t h e hinge,

Using only t h e manipulator, t h e pin w a s repositioned i n t h e hinge, a new c o t t e r key was i n s e r t e d i n t h e pin and t h e key w a s spread t o lock it i n place,'

During t h i s r e p a i r , t h e lower key w a s dislodged and was a l s o re-

paired. Operation o f t h e pheumatic clamps on t h e access port door occasiona l l y r e s u l t e d i n gaseous f i s s i o n products being vented through t h e operator discharge l i n e s .

Therefore a small charcoal f i l t e r was added t o t h e op-

e r a t o r discharge l i n e s t o prevent t h i s a c t i v i t y from being discharged t o

315

ONE qe-in. HOLI THREE '4in. HOi THREE (/e-in. HO THREE '/3*-in. H(

'4in.

ORNL-DWG 69-!3!05

BETWEEN WINDOWS

SOLID NICKEL SPACER

34-in. OD x '/=-in. WALL NICKEL TUBING

ORNL-OW0 61-4184

-WINDOWS

STAINLESS STEEL CABLE

ORNL-rn 67-4785

5/a-in. OD I3-in. TALL GRAPHITE CYLINDER

-5/4-111.

TOPOF

OD NICKEL

VOLUME x)CC---,

h 4

WELU MIST SHlELDs, SHOULDER FOR GRA CYLiNDER TO SIT

CuO IN STAINLESS STEEL CLOTH BASKET

34-in.LOD Ni TUBE

'/a-in.-DIAM PERFORATIONS ABOUT EVERY '4in, 3 Be RODS '&-in. OD x

1 INLET I

L'2seF4\

NICKEL CAPILLARY

P Fig. 18.5a Freeze Valve Capsule

SECTION 6 - B

Fig. 8.5b Nickel Cage Beryllium Rod Assembly a f t e r Exposure t o MSRE Fuel S a l t for 10.5 hr

Fig. 1 8 . 5 ~Hydrocarbon Analyzer ,

c ,

c .g. 18.5d Surface Tension Capsule

316

t h e atmosphere via the stack.

Also t h e vent valves from t h e gas operators

were kept closed except during openings and closures of t h e access door. The neoprene seal at t h e access door continued t o d e t e r i o r a t e .

April of 1967, t h e increased b u f f e r gas leakage across t h e inner seal was countered by lowering t h e b u f f e r pressure requirement i n t h e s a f e t y i n t e r lock system from 50 t o 40 psia.

I n J u l y 1967, t h e s a f e t y switch s e t t i n g

w a s lowered t o 35 p s i a and t h e helium flow rate t o t h e b u f f e r seal was a l s o

increased so t h a t a s a t i s f a c t o r y b u f f e r pressure could be maintained. During t h e intensive sampling of t h e l a s t run, t h e access p o r t door frequently required repositioning w i t h t h e manipulators before t h e second set of clamps were actuated t o lock t h e door and obtain an adequate seal. The cause of t h e d e t e r i o r a t i o n of t h i s seal has not been determined.

The r a d i a t i o n l e v e l of some of t h e samples taken w a s i n excess of 1000 R/hr

at 3 inches and during boot replacements when t h e sampler was empty, t h e radiation l e v e l i n s i d e Area 3A was approximately 100 R/hr. Although some long-term d e t e r i o r a t i o n may be a t t r i b u t e d t o r a d i a t i o n damage, more l i k e l y t h e reduction i n buffer pressure during t h e l a s t two runs can be a t t r i b u t e d t o t h e f a c t t h a t t h e l e f t center h u - v i s e clamp w a s loose and thus ineffec-

t i v e i n i t s closed position. The f l e x i b l e containment membranes (boots) on t h e manipulator continued t o be a chronic source of trouble.

Generally, when a leak developed,

t h e sampling cycle would be completed before r e p a i r s were made; however, during t h e remainder of t h e sampling cycle, containment was abetted by t h r o t t l i n g t h e boot evacuation valve t o assure t h a t any leakage would be i n t o t h e plenum between t h e boots and through t h e vacuum pump t o t h e s t a c k r a t h e r than t o t h e work area.

After f u l l power w a s achieved, replacement

of t h e boots became somewhat more involved because t h e manipulator assembly was quite contaminated.

During a boot replacement i n e a r l y 1967, t h e radi-

a t i o n l e v e l of t h e manipulator arm was 300 R/hr a t 3 i n .

It was l a t e r de-

contaminated and saved f o r possible f u t u r e use. Also while sampling during e a r l y full-power operations, t h e r a d i a t i o n l e v e l at t h e operating area increased t o approximately 30 mR/hr. The radia t i o n l e v e l was reduced by an order of magnitude by changing t h e sampling procedure so t h a t a purge of helium from Area 1C t o t h e pump bowl w a s

317 maintained when t h e operational and maintenance valves were open.

Also on

r e t r i e v a l i n t o Area lC, t h e samples were purged f o r 1-1/2 h r t o t h e off-gas system before f u r t h e r removal operations were continued.

On two occasions during t h e routine t r a n s f e r of capsules between Areas 1C and 3,a capsule was inadvertently dropped and r o l l e d out of sight and out of manipulator range i n Area 3A.

These were l a t e r found and

recovered by t h e use of a mirror and a grappling fork which were lowered through t h e removal valve. During t h e l a t t e r p a r t of Run 19, routine sampling was suspended f o r approximately four days and t h e sampler w a s adapted t o accommodate a collimator and a germanium c r y s t a l detector atop t h e sampler where t h e c a r r i e r cask i s normally positioned.

A p l a s t i c plug i n t h e removal area permitted

t h e unobstructed view i n t o Area 3A where samples r e t r i e v e d f r o m t h e f u e l pump were positioned.

Gamma-ray spectrometry data were then collected on

short-lived fâ&#x20AC;&#x2122;uel f i s s i o n products within 50 minutes of t h e i r removal from t h e c i r c u l a t i n g salt stream.

18.1.4

Discussion and Conclusions

The main operational problems encountered w i t h t h e sampler-enricher r e s u l t e d from 1) manipulator boot f a i l u r e , 2 ) lodging of t h e capsule o r l a t c h at t h e sampler tube entrance, operational valve, o r maintenance valve with subsequent cable tangling, and 3) b u f f e r s e a l leaks at t h e operation and removal valves and a t t h e capsule access door. . The cause of some of t h e boot f a i l u r e s were not determined because t h e leaks were small and t h e actions which produced t h e ruptures were not always coincident with t h e discovery of t h e leak.

Also t h e boots were not

examined i n d e t a i l because of t h e high r a d i a t i o n l e v e l of t h e boots and manipulator assembly.

It i s

elieved t h a t t h e boot failures which could

not be a t t r i b u t e d t o any s p e c i f i c action occurred during t h e capsule loading operation where t h e manipulator i s r e t r a c t e d and pushed downward and

t o t h e l e f t i n order t o load t h e sample i n t o t h e l i n e r of t h e transport tube bottom.

A t times ( a

sampler operator had t o r e s

invariably w i t h t h e long capsules), t h e

o p u l l i n g t h e manipulator w i t h both hands

i n order t o r e t r a c t it far enough t o i n s e r t t h e capsule and l a t c h pin i n t o

the liner.

Had t h e capsule loading s t a t i o n been mounted t o t h e l e f t of

318 -

t h e manipulator c e n t e r l i n e , it could have been located c l o s e r t o Area 1 C

Another possible a l t e r n a t i v e

and thus require l e s s manipulator t r a v e l .

would be t o f a b r i c a t e t h e boots w i t h fewer convolutions; however, t h e convolution width would have t o be increased t o maintain t h e same extended length,

The r e t r a c t e d thickness of t h e boots would then be decreased per-

mitting t h e manipulator t o be pulled back f a r t h e r than previously. Why and where t h e capsule o r l a t c h occasionally lodged i n t h e sampler tube during capsule i n s e r t i o n has not been s a t i s f a c t o r i l y answered.

How-

ever, t h e use of t h e proximity switch minimized t h e p o s s i b i l i t y of serious cable tangling problems.

Perhaps two a d d i t i o n a l detection devices mounted

above and below t h e operational valve would have determined t h e exact trouble spot e The buffer seal leak at the access port door started t o increase i n

January 1967 and continued t o increase slowly f o r t h e remainder of E R E operations,

Perhaps some of t h i s d e t e r i o r a t i o n can be a t t r i b u t e d t o radiThe r a d i a t i o n l e v e l i n Area 3A was

ation damage t o t h e neoprene gaskets. 10HOO R/hr during intensive sampling.

Also, t h e access door w a s always

i n t h e closed position except during capsule t r a n s f e r ; t h e r e f o r e t h e neoi

prene w a s almost always compressed which oveP t h e years has probably caused

it t o y i e l d t o a thinner protrusion.

I n addition, t h e s t e l l i t e p l a t e s on

t h e access door appear t o be gouged s l i g h t l y by the s t e l l i t e - t i p p e d Knu-vise clamps,

ATSO the l e f t center clamp w a s discovered t o be loose i n i t s closed

position during t h e last run.

Therefore, it i s believed t h a t ordinary wear-

and-tear of these components was t h e primary cause of seal leakage.

The

beating t h a t t h e door took could possibly have been lessened by a flow res t r i c t o r t o t h e clamp actuating cylinders, The top s e a l s of t h e buffered valves developed leaks whereas t h e bottom seals apparently suffered only slight d e t e r i o r a t i o n which indicates t h a t p a r t i c u l a t e matter which occasionally drop on t h e valves w a s t h e cause

of seal deterioration.

Consequently t h e volumes above these valves showed

a gradual increase i n pressure and required constant venting.

desirable of these leaks was t h a t at t h e removal valve.

The l e a s t

On removal of a

sample, t h e transport tube and removal t o o l which a r e withdrawn from Area 3A a r e locked i n position above t h e removal valve and t h e removal valve i s then closed.

If an appreciable delay i s encountered between t h e

319

time t h a t t h e valve i s closed and t h e time t h a t t h e transport tube i s w i t h drawn f r o m t h e removal s e a l i n t o t h e shipping cask, appreciable pressure (due t o s e a l leakage) can b u i l d up i n t h e removal a r e a and blow contamination out of t h e removal area t o t h e top of t h e sampler and shipping cask when t h e transport tube c l e a r s t h e two O-rings of t h e removal s e a l . Swabbing the removal area w i t h wet wipes on a monthly basis g r e a t l y reduced t h e contamination problem encountered i n t h i s area.

During intensive sam-

pling, t h i s area should be cleaned more frequently. The s p e c i f i c a t i o n s f o r t h e e l e c t r i c a l i n s u l a t i o n on t h e lead dires near t h e penetrations a t Areas 1C and 3A should be reviewed.

A failure

a t t h e connector pins occurred i n May 1966 r e s u l t i n g i n open c i r c u i t s i n t h e i n s e r t and withdraw modes and i n t h e upper l i m i t l i g h t . During in, spection of t h e 1C assembly which was removed i n August 1967, t h e i n s u l a t i o n on .the lead wire was found t o be very b r i t t l e and most of it w a s unavoidably: stripped during removal of t h e drive unit.

I n November 1967 an

e l e c t r i c a l fault occurred i n t h e same c i r c u i t s as those i n May 1966 except t h e f a i l u r e occurred i n t h e wires between t h e two containment boxes.

i s not c l e a r whether these f a i l u r e s can be a t t r i b u t e d t o radiation damage

t o the i n s u l a t i o n o r possibly t o excess temperature caused by resistance heating i n t h e lead wire due t o blocked r o t o r current (0.3 amps) i n combination w i t h t h e heat supplied by t h e illuminator lamp. To increase t h e precision of uranium i s o t o p i c analyses, provisions

should be made t o clean Area 3A periodically, especially a f t e r enrichments a r e made.

During t h e enriching procedure, small q u a n t i t i e s of enriching

salt could e a s i l y adhere t o t h e manipulator and/or be 3arred loose from t h e capsule and f a l l t o t

oor of Area 3A where a l l capsules a r e l a i d

temporarily while swappin

s u l e s t o and from t h e l a t c h and t h e transport

tube.

A very small amount of enriched salt adhering t o a sample submitted

f o r i s o t o p i c analysis could d i s t o r t t h e i n t e r p r e t a t i o n of t h e r e s u l t i n g analyses.

320

18.2

Coolant Sampler

In principle, sampling of t h e coolant system was t h e same as t h a t f o r the f u e l system.

However, t h e induced a c t i v i t y i n t h e coolant salt was

very short-lived and there w a s no residual radioactivity i n t h e sample; therefore, no shielding was required and only single containment was necess a r y during transport.

Direct manipulation of t h e sample by one hand

through a glove p a r t simplified sampler construction and operation. 18.2.1

Description o f t h e Coolant Sampler

The coolant salt sampler, shown i n Fig. 18,6 consisted e s s e n t i a l l y

Two b a l l

of a dry box connected t o t h e coolant pump by a transfer tube.

valves i s o l a t e d t h e sampler from t h e pump bowl which has a mist s h i e l d and capsule guide similar t o t h a t i n t h e f u e l pump.

The 1-1/2 i n . Sched-40

t r a n s f e r tube was similar t o t h a t used f o r t h e f u e l system except t h a t no expansion j o i n t was required because t h e coolant pump position d i d not change with system temperature as did t h e fuel pump.

The sampler was pro-

vided with l i g h t i n g and viewing ports and t h e necessary capsule t r a n s f e r s inside t h e sampler were done with one hand through a glove port.

Sliding

gas seals at t h e top of t h e sample c a r r i e r along with a vacuum pump and a helium supply were used t o reduce oxygen contamination t o a minimum during sampling and during sample transport t o t h e a n a l y t i c a l laboratory. A check-list type procedure along w i t h a system of locks and keys f o r

valve manipulation w a s used t o assure proper sequence of operations, key was used t o unlock one valve, which could then be opened.

When t h e

valve w a s opened, it also locked the first key i n position and released a second key,

Removal of t h e second key (which must be removed t o unlock

t h e ne& valve i n sequence) assured t h a t t h e first valve was locked i n the proper position, 18.2.2

Operating Experience with t h e Coolant System Sampler

The coolant salt sampler operated r e l a t i v e l y trouble-free throughout MSRE operations.

Only 81 coolant system samples were taken as opposed t o

745 cycles f o r t h e f u e l system.

321

F i g . 18.6

Coolant Salt System Sampler

322 Early i n 1966 during one sampling operation, t h e l a t c h f a i l e d t o s l i d e

No reason f o r t h e d i f f i c u l t y could be However, the outside diameter of t h e l a t c h w a s reduced and no simi-

f r e e l y through t h e t r a n s f e r tube. found,

lar t r o u b l e w a s reported. Also i n the first p a r t of

ig66, one pin i n an i n d i c a t o r

shorted out at t h e penetration during a sampling cycle.

light circuit

The receptacle w a s

removed and a new one welded i n place. During t h e same period, t h e leak rate f r o m t h e lower seal of the removal valve increased.

A f t e r t h e valve w a s disassembled, cleaned, and re-

assembled, the s e a l w a s s a t i s f a c t o r y ; however, after approximately eight additional samples, t h e seal leakage again increased and was corrected by replacing the valve seals.

Also an extra washer w a s added t o the valve t o

permit more mechanical pressure t o be applied t o t h e seals.

Thereafter the

coolant sampler operated trouble-free f o r t h e f i n a l t h r e e years of MSRE operation.

18.2.3

Conclusions and Recommendations

From an operational viewpoint, two improvements may be worth considering: (1) The glove port cover should be hinged t o swing out horizontally.

On several occasions when t h e clamps w e r e removed, t h e glove port cover slipped f r o m t h e operator's hand and dropped f r e e l y t o i t s stopped position.

I n doing so, it could have pinched a bundle of nearby insulated

electrical wires.

(2) The sample c a r r i e r should be lengthened approximately two inches t o accommodate t h e longer capsules which were used during t h e l a s t run.

323

19.

CONTROL RODS

M. Richardson ~

19.1

Description

Three control rods were used i n the MSRE core vessel.

The rods were

not exposed d i r e c t l y t o t h e f u e l salt but were contained i n t h r e e "blind well" thimbles which were located i n three of t h e four channels surrounding t h e v e r t i c a l centerline of t h e graphite core.

The thimbles were straight

within the core vessel, but above t h e reactor access flange they were offset

- using two 16-in.

containment standpipe. tion.

radius bends t o provide space f o r t h e core sample A r o l l e r was located at each bend t o reduce f r i c -

The control rod drive mechanisms were v e r t i c a l l y positioned around

t h e standpipe.

The control rods were fabricated of flexible metal hose

w i t h hollow center c y l i n d r i c a l poison elements (38 per rod) beaded over

t h e lower 60 i n .

Rod motion w a s supplied by motor driven 1/4-in.

chain and sprocket drive units.

roller

Included i n t h e reverse locking power

t r a i n i s an overrunning clutch which permits power transmission i n the inA magnetic clutch w a s used t o release each rod from

s e r t d i r e c t i o n only. '

t h e power t r a i n gearing and permitted f r e e f a l l of t h e rod i n t o t h e core at k0.4 g. Continuous indication of each rod position w a s provided by two synchro torque transmitter-receiver p a i r s , one "fine" and one "coarse."

A potenti-

ometer provided position indication f o r s a f e t y interlocks i n t h e reactor f i l l circuits. The upper and lower limits of rod t r a v e l were controlled by four (2 .

upper and 2 lower) mechanically actuated switches. '

The poison elements were cooled by c e l l a i r (95%N2) from t h e compon-

ent coolant system.

This entered t h e upper end of t h e hollow control rod

and exhausted r a d i a l l y i n t o t h e thimble from a nozzle at t h e lower end.of the rod. Changes i .,n pressure drop as t h i s nozzle movedthrough a flow r e s t r i c ; t o r b u i l t i n t o the bottom of t h e thimble provided a rod position reference point ( f i d u c i a

ero).

Comparison of t h i s with t h e rod position as indi-

cated by t h e synchro rod position instrumentation w a s used t o monitor s t r e t c h i n g of t h e rods and other malfunctions.

324

c-)

The rod drive units were i d e n t i c a l as were t h e rods except f o r a slight variation i n lengths of the individual rods due t o t h e differences

of the thimble offsets. or trouble-shooting.

This allowed interchange of components f o r repairs

For i d e n t i f i c a t i o n of a rod assembly, it was neces-

s a r y t o specify the thimble number (T-1,

o r R-4)

R-2, R-3,

11-2, o r T-31,

and the drive rimer (V-1, V-2, V-3,

the rod number (R-19

o r V-4).

!he rod

assembly connected t o t h e servo controller was considered t o be the regul a t i n g rod and the other two rods were t h e shim rods.

The rod assemblies

could be used interchangeably as shims o r regulating rods by shi-fting t h e out-of-cell

rod drive disconnects.

Any one rod was capable of taking the

reactor subcritical. 24 The rods and drive assemblies were not designed f o r in-cell maintenance,

Repairs on t h e drive units were done a f t e r disassembling the rod

from the drive and removing the drive unit from t h e reactor c e l l by remote maintenance methods,, The control rod shock absorbers used t h e general principles of a ty-pic a l hydraulic shock absorber but differed i n t h a t t h e working "fluid" con-

sisted of 3/32-in.-diameter

steelballs.

A t t h e end of a scram, t h e bottom

face of the shock absorber cylinder struck permanently-mounted s t e e l blocks which were bolted t o t h e housing.

The shock absorber plunger, t o which the

control rod was attached, continued t o move downward and was decelerated by the forces developed by t h e buffer springs and by the flow of .the s t e e l b a l l s around a knob on the plunger.

bers were adjusted t o ~

The stroke length of t h e shock absor-

3 inches ~ 5 f o r each rod, 19.2

I n i t i a l Testinq

The rods were i n s t a l l e d i n the reactor vessel i n January of 1965.

The

l i m i t switches were adjusted t o coincide with t h e rod position indicators on the main console and e l e c t r i c a l continuity t e s t s were performed on all drives p r i o r t o assembly t o t h e control rods.

Satisfactory anibient tem-

perature performance was checked with t h e r e s u l t s given i n Table l g o l 0 /

325

Table 19.1

I n i t i a l Tests of Control Rod Assemblies Spare Components

Thimble No.

T-1

T-2

T-3

Rod No.

R-1

R-2

R- 3

Drive No.

v-3 0.6

v-2

v-1

Shock stroke, inches

0.59 2.85

0.6 0.58 3.6

0.6 0.58 3.7

Scram time, secs

0.835

0 752

0.775

3.8

3.9

---

0.53

0.52 50.9

Motor current w i t h d r a w Motor current i n s e r t

A i r flow, rod

- amps

- scfln

Rod t r a v e l speed, in./sec

0.53

N l rod t r a v e l , inches

50.91

50.95

Fiducial zero, inches

1.4

1-35 .

19.3

0.53

R-4 v-4 0.6 0.59

3.0

---

Periodic Testing

During reactor operation and especially a f t e r power operations commenced, it w a s not possible t o run extensive t e s t s on t h e rods.

However, it w a s necessary t o do s u f f i c i e n t t e s t i n g t o assure t h a t t h e rods would scram i f needed.

It was a l s o important t o know t h a t t h e rod position indicators were functioning properly as these were used i n making r e a c t i v i t y balances. To accomplish t h i d t o aid i n a n t i c i p a t i n g necessary maintenance , t h e following periodi A.

Rod Scrams -Scram

s t s were made,

t e s t s were routinely conducted t o determine

rod scram times before a f i e 1

11 and before c r i t i c a l operation.

Each

rod w a s withdrawn t o a height of 50 i n .

The magnetic clutch current was broken by t r i p p i n g t h e manual scram switch. The rod would f r e e - f a l l t o t h e fully-inserted position. The 50-in. drop w a s e l e c t r o n i c a l l y timed f r o m t h e moment of release t o t h e lower l i m i t switch.

The maximum permissible scram

time w a s 1.3 sec/50 i n . rod scram t e s t s could not be made during B. Rod Exercise -Since nuclear operation, each rod w a s exercised d a i l y t o demonstrate t h a t they

would move on command and a l s o t o f l e x t h e metallic hose.

The f l e x i b l e

hoses tended t o s t i f f e n with time i f allowed t o remain i n one position f o r Observation by t h e reactor operator of t h e rod-position

prolonged periods.

indicators during t h e exercise ascertained t h a t t h e synchros were working properly and t h a t t h e rods operated f r e e l y i n t h e thimbles.

Fiducial Zero - A s

described above, t h e f i d u c i a l zero was a fixed

reference position i n each of t h e control rod thimbles which r e l a t e d t h e actual rod position t o t h e rod position indicated by t h e position-in&cating potentiometers.

Fiducial zero t e s t s were routinely made before a f u e l f i l l

and before c r i t i c a l operation.

19 4 Operating Experience Except f o r d i f f i c u l t i e s with t h e rod assemblies i n thimble T-1,

opera-

t i o n w a s s a t i s f a c t o r y from both an o p e r a t i o n d and maintenance standpoint There w e r e no unscheduled shutdowns because of a c o n t r o l rod f a i l u r e .

The

t e s t i n g program was adequate t o allow advance maintenance planning and work

was done i n conjunction with maintenance i n other areas a t t h e time of a scheduled shutdown. During over 3,000 rod scrams from 50 inches (see Table 19.2) and numerous others from d i f f e r e n t elevations, t h e r e was only one time when a rod f a i l e d t o scram.

Except f o r t h i s and one other period, t h e scram times

were all l e s s than t h e 1.3 seconds s p e c i f i e d i n t h e s a f e t y limits and were usually less than 1 second.

known.

The number o r amount of rod movement is not

Once positioned, there w a s l i t t l e movement of t h e shim rods.

The

regulating rod, normally t h e assembly i n thimble T-1, moved l i t t l e at steady power u n t i l bubbles appeared i n t h e reactor.

From t h a t point on t h e regu-

l a t i n g rod moved frequently under servo control t o compensate f o r t h e action of bubbles on t h e r e a c t i v i t y .

Several rod jog tests were made and it is

estimated t h a t t h e regulating rod moved about 720 jogs per hour at 0.4 in. per jog f o r a t o t a l of 32,400 jogs. inches

The t o t a l movement was about 14,000

327

Table 19.2

Estimated Number of 50-inch Rod Scrams

Thimble No.

Rod No.

Drive No.

No. of Scrams

T-1

R-1

v-3

750

T-1

R-1

v-4

150

T-2

R-2

v-2

930

T-3

R-3

800

T-3

R-4

v-1 v-1

. >

430 -

Total

3060

The estimated l i f e of t h e control rod motors was 10,000 hours at a

reactor power of 10 MW (Ref.

49).

The condition of a l l drive motors, lu-

b r i c a t i o n , wiring, and gears remained s a t i s f a c t o r y throughout t h e e n t i r e operation.

These were l a s t inspected i n J u l y 1969 at which time t h e reac-

t o r had accumulated 92,805 MWhrs

The h i s t o r y of t h e MSRE control rods and drives a r e b e s t summarized i n Figs. 19.1, 19.2, and 19.3.

Control rods R-1 and R-2 remained i n ser-

vice t h e e n t i r e operating period i n thimbles 1 and 2. and R-4 were interchanged f a i l u r e s o f t h e drives a r e Heat-up and p r e - c r i t i

Control rods R-3

l e 3 as shown i n Fig. 19.3.

The component

a t e d i n Table 19.3.

t c i r c u l a t i o n commenced i n January 1965

and during t h e e a r l y perio

s operation, d i f f i c u l t y was encountered

with t h e l i m i t switches.

cations were t h a t t h e switch operators

were s t i c k i n g o r g a l l i n g randomly on all t h r e e rods. were made during t h e M vealed t h a t t h e s l i d i n g

Temporary r e p a i r s

April 1965 shutdown when investigation re-

rs of t h e switch operators had galled causing

t h e f a u l t y operation. The drives were removed from t h e r e a c t o r c e l l i n July 1965 f o r i n s t a l l a t i o n of modified l i m i t switch operators on a l l t h e u n i t s including t h e

FIDUCIAL ZERO (in.) MaintemHistory

Installedtemperatunswitch Replaced limit switch operator

AUGUST Replaced potentiomaer

Replaced ternparatun switch Replaced potentiometer

SEPTEMBER

APRIL MAY JUNE

n I D

JULY

I B r <

AUGUST SEPTEMBER OCTOBER NOVEMBER DECEMBER JANUARY FEBRUARY MARCH APRIL MAY JUNE JULY

tJANUARY

FEBRUARY

MARCH

Finc synchro failed Potentiometer noisy Rod No. 1 used aa shim rod

Replaced V-3 driw with V - 4 Rod No. 1 repulsing rod

8ZC

FIDUCIAL ZERO (in.) Maintenma Historv

SCRAM TIME

Installed temperetun switch Replaced limit switch operaor

g f

MAY JUNE JULY AUGUST SEPTEMBER OCTOBER NOVEMBER. DECEMBER JANUARY FEBRUARY

P-

MARCH APRIL

MAY

JUNE

JULY AUGUST

I MARCH APRIL MAY

Replaced fine synchro

JUNE JULY AUGUST

w 0 a

SEPTEMBER OCTOBER NOVEMBER

Replaced finr synchro

E SEPTEMBER OCTOBER

t Replaced potentiometer

6Z€ r

FIDUCIAL ZERO (in.)

SCRAM TIME (sed

â&#x20AC;&#x153;s ffi

Maintenam History

Installed temperature switch Replacedlimit switch operator

MAY

Repairedroller in T-3 Repaired R-3 upper hose

Bent limit switch omator

JUNE

DECEMBER JANUARY FEBRUARY

P= APRIL

R-3 sticking in thimble ReplacedR-3 with spare rod R-4

(D 0) CD

Bent limit switch operator

lunRccl I APRIL

I P

JULY AUGUST

SEPTEMBER OCTOBER

NOVEMBER

D rn -<

MAY JUNE

r m

DECEMBER

JANUARY FEBRUARY

50 inch scram time increase

Added 1.5 Ib weight to V-1 0.9 scram time Repairedrod R-3 No. 3 assembly used as reg rod

No. 3 failed to scram on demand ReplacedR-4 with R-3 Replaced potentiometer

331

Table 19.3

Drive Unit No.

Drive U n i t Component Failures

v-2

Service Time as Reg. Rod Service Time as Shim Rod

4 mo. 56 mo.

v-2

--60 mo.

v-4

v-3

5 1 mo. 4 mo.

(Spare Drive) 5 mo.

Temperature switches L i m i t switch operator

L i m i t switch operator push rod

Total

---

1000-R potentiometer

Fine position synchro transmitter

Coarse p o s i t i o n synchro t r a n s m i t t e r

spare

(V-4).

The new operators contained case-hardened bearing surfaces,

t h e r e were no more d i f f i c u l t i e s of t h i s type with t h e switch operators. Examination of t h e control rods during t h i s period revealed t h a t t h e upper w i r e sheathed upper hose of t h e No. 3 rod was badly worn at a point

28 i n , below t h e flange.

Examination of t h e No. 3 thimble revealed t h a t

t h e upper r o l l e r , on which t h e control rod operates, d i d not r o t a t e which caused t h e severe wear t o t h e rod.

The r o l l e r was replaced and a new hose

i n s t a l l e d o n â&#x20AC;&#x2122; t h e No. 3 control rod. During t h i s period temperature switches were a d d e d t o t h e lower drive mechanism.

These small bi-metallic switches were t o alarm should t h e tem-

perature within t h e drive u n i t cases exceed 200Â°F,

Without t h e downward

sweep of gas through t h e housing, convection of t h e heat r i s i n g from t h e thimble could cause a rapid temperature increase and r e s u l t a n t damage t o t h e drive assembly.

After t h e rod drives had been r e i n s t a l l e d and t h e re-

actor heated and f i l l e d with s a l t , t h e switches were t e s t e d by turning off

332

t h e gas flow.

Since no ala,rms occurred, another method'of measuring t h e

drive assenibly was devised which, although not precise, w a s adequate f o r operation.

Thls'was done by s h u t t i n g o f f t h e i n t e g r a l drive motor cooling

fan and measuring t h e resistance of t h e windings of t h e fan motor. " A s l i g h t l y greater than 10% increase i n r e s i s t i v i t y equaled %50째F change. winding temperature.

From t h e measured r e s i s t a n c e and* using t h e temperature . , coefficient of resistance of t h e wire, a curve w a s p l o t t e d r e l a t i n g t h e .

measured resistance t o t h e approximate motor temperature.

This value w a s

rdughly r e l a t e d t o t h e drive housing temperature. I n e a r l y operation, t h e control-rod servo system had not functioned properly due t o coasting of t h e r e m a t i n g rod motor and s h i m locating motor,

Brakes were i n s t a l l e d i n both of t h e s e (see 24.5).

19.4.1

Rod Assembly i n Thinible T-1

Figure 19.1 gives t h e rod drop t i m e s and f i d u c i a l zero values as w e l l

as t h e maintenance h i s t o r y f o r t h e rod assemblies located i n thimble T-1. This was first made up of Rod R-1 and drive u n i t V-3 and was used as t h e

regulating rod, except f o r a four-month period, f o r t h e e n t i r e r e a c t o r ope r a t i o n and consequently performed t h e bulk of t h e control rod service.

Other than replacement of f a u l t y drive components, t h e r e was l i t t l e d i f f i c u l t y with thi-s assembly..

The.JOOO-R,potentiometer position i n d i c a t i n g potentiometer f a i l e d 3 times due t o an o p e n - r e s i s t a n c e ' c o i l and 1 t i m e due t o a worn r e s i s t a n c e _.

coil.

This d i f f i c u l t y had been a n t i c i p a t e d as it i s a common failure f o r

t h i s type,component i n continuous service, One temperature switch f a i l e d due t o an e l e c t r i c a l ground a t t h e wire connections at t h e switch.

The ground w a s created by bumping t h e switch ,. .. during assembly of t h e drive i n t o t h e case. .

There w a s an a p p a r e n t . s h i f t . o f 1/2-fn. i n November of 1965.

i n t h e f i d u c i a l zero position

The change was a t t r i b u t e d t o slTppage of t h e chain

on t h e sprocket but examination. and t e s t i n g f a i l e d to' reproduce t h e s l i p o The 1/2-in0,, d e d a t i o n w a s ,recovered involuntarily

July of 1967 which

&in

would-indicate t h a t t h e exact cause of t h e s h i f t 5 s not known.

The fidu-

c i a l zero position p l o t i? Fig, lg.1 shows the gradual increase, .%1/2 inch,

.. " . I

. . .

333

rod from t h e e f f e c t s of usage.

i n t h e length of t h e contro

The change

i n length which occurred i n September 1969, i s t h e result of changing t h e drive u n i t . The rod scram times remained at 0.8 k .05 seconds from i n s t a l l a t i o n u n t i l t h e spring of 1969. 1.03 sec for a 50-in.

I n March of 1969, t h e scram time increased t o

drop and a l s o d u r i n g t h i s period, t h e r e w a s a failure

of t h e f i n e position synchro t r a n s m i t t e r and t h e 1000-R pot.

Analysis (see

19.5) of t h e rod acceleration and velocity during a scram revealed that t h e r e was an area of drag i n t h e lower 20 i n . of t h e scram. evidence, t h e V-3 drive was replaced w i t h t h e spare drive V-4.

From t h e above With drive

V-4 and rod R-1 i n thimble T-1 t h e scram time returned t o ~ 0 . 8 sec u n t i l shutdown.

A d e t a i l e d inspection and maintenance program i s planned f o r

drive V-3 but i s as yet incomplete. 19.4.2

Rod Assembly i n Thimble T-2

As shown i n Fig. 19. drive, R-2 rod.

t h e assembly i n thimble T-2 w a s made up of V-2

This assembly w a s used as a shim rod, except f o r short

t e s t periods-, f o r t h e entire.MSRE operation.

Other than t h e mechanical

f a i l u r e s shown i n Fig. 19.2, t h e r e were no operating d i f f i c u l t i e s . The mechanical f a i l u r e s included two f i n e position synchro f a i l u r e s and one potentiometer.

Examination of one of t h e synchros ( t h e radiation

level of t h e svnchro was Q l O mR/hr) revealed t h e failure t o be t h e w i o e r arm contact between t h e r o t a t i n g shaft and t h e windings. The f i d u c i a l zero h i s t o r y shown i n Fig. 19.2 c l e a r l y shows t h e progressive s t r e t c h i n g o f t h e rod t o be *1/2 i n . f r o m t h e time of i n s t a l l a t i o n t o t h e end of operations.

19.4.3

Rod Assembly i n Thimble T-3

This assembly was made s t a l l a t i o n as shown i n Fig. i n t h e figure, t h i s assemb control rods.

of V-1 drive and R-3 rod a t t h e time of inAs indicated by t h e scram h i s t o r y shown

t h e most troublesome of t h e t h r e e MSRE

The drive u n i t s were removed during t h e p r e c r i t i c a l mainte-

nance period i n t h e summer of' 1965 t o i n s t a l l t h e modified l i m i t switches described elsewhere.

The No. 3 c o n t r o l rod w a s inspected a t t h i s time and

found t o have a t o r n section i n t h e upper hose, 28 i n , below t h e flange. The damage t o t h e rod was t h e result of t h e f a i l u r e of t h e upper r o l l e r

334 i n t h e No. 3 thimble which had jammed and would not r o t a t e .

The r o l l e r

and rod were repaired and after reassembly, t h e rod scram time w a s C0.8 sec

. During t h e low power period of operation, it w a s found t h a t after a f u l l scram t h e lower l i m i t switch would not c l e a r when t h e rod w a s withdrawn.

It was necessary t o push t h e switch actuator by withdrawing t h e

It would function The bottom end of the. l i m i t

rod t o t h e upper l i m i t t o c l e a r t h e , l o w e r l i m i t switch. normally u n t i l t h e rod was again scrammed.

switch push rod w a s found t o be bent below t h e lower r e t a i n i n g bushing. The rod had been bent by s t r i k i n g t h e lower housing flange as t h e r e s u l t of a weak push rod spring.

Repairs consisted of straightening t h e push rod

and i n s t a l l i n g a stronger spring. Later t h e control rod commenced t o s t i c k i n t h e thimble a t a point %1-1/2 in. above t h e f i d u c i a l zero position.

The rod could be freed by

jogging and would hang i n no other position.

Direct examination of t h e

thimble was not possible, but t h e e x t e r i o r of t h e rod w a s examined and appeared t o be i n good condition.

The d i f f i c u l t y appeared t o be t h a t a sharp

projection such as a broken weld at t h e t h r o a t section of t h e f i d u c i a l zero position e x i s t e d on which t h e rod could snag. was replaced with t h e spare rod R-4,

However, when t h e R-3 rod

it d i d not snag.

This rod operated within limits but t h e scram time increased slowly from 0.85 sec i n May 1967, up t o 1,26 s e c i n December 1968. of both t h e drive and rod revealed no apparent defects.

Examination

Replacing t h e V-1

drive with t h e V-2 drive u n i t i n t h e T-3 p o s i t i o n f o r t e s t purposes made l i t t l e change i n the scram t i m e .

Therefore a 1.5-lb weight w a s added t o

t h e shock absorber assembly cf t h e o r i g i n a l V-1 drive u n i t and it was reinstalled.

This r e d u c e d t h e scram t i m e t o l e s s than 0.9 see.

The reactor was shut down i n June 1969 by manually scramming all t h e control rods.

The assembly i n thimble T-3 (R-4 and V-1) was a t 35.0 in.

and did not scram.59

The rod w a s i n s e r t e d 0.2 fn. and scrammed again, t h i s

time t h e rod dropped freely.

Repeated attempts t o make t h e rod s t i c k fol-

lowed t h i s incident but were unsuccessful. scram tests at 2-in. I

Fig. 19.4 shoys t h e r e s u l t s of

increments taken on June 2 before t h e r o d was exam-

ined and c l e a r l y shows t h e area .of high drag,

335

1.8

I .6

1.4

.2 c

:a:

I I-

0.6

0.4

0.2

ORNL-DWG 69-8976A

336

To correct t h i s d i f f i c u l t y , Rod R-4 w a s removed and replaced by Rod

R-3.

This w a s the rod which had earlier snagged i n t h e thimble.

Before

r e i n s t a l l i n g it, t h e air exhaust tube (a possible cause of t h e snagging) The defective potenti-

w a s f i l e d smooth t o eliminate all t h e sharp edges. ometer w a s also replaced on the drive (V-1).

The greatly improved perform-

ance of t h i s assembly (R-4 and V-1 i n T-3) suggests t h a t t h e trouble was i n t h e rod (R-4) rather than i n t h e drive.

The extreme radioactivity of

rod 3 (R-3) made it impractical t o examine. Figure 19.4 a l s o shows the results of drop t e s t s conducted on August 4, a f t e r a l l repairs had been accomplished f o r all t h r e e control rods.

This

was the condition of t h e assemblies during t h e f i n a l period of operation.

19.5

Improved Rod Scram T e s t i n g

It was obvious from t h e No. 3 rod failure t o scram that t h e method of

t e s t i n g each rod by a single 50-in.

scram was inadequate.

f a i l e d t o expose any l o c a l areas of high drag.

The method

The incremental method of

t e s t i n g , scram time vs s t a r t i n g - p o s i t i o n , as shown i n Fig. 19.4, showed no region of excessive drag f o r any of the rods.

cp,

In an e f f o r t t o provide

a quick t e s t i n g procedure t h a t would not require numerous rod drops, yet would reveal regions of.abnormal drag, a procedure developed f o r t h e EGCR

w a s renovated.

The output from t h e 1000-ohm position potentiometer was

amplified and transmitted by wire t o t h e main ORNL area, where it w a s passed through a f i l t e r w i t h a 5-Hz time constant t o remove transmission noise, d i g i t i z e d a t 2000 samples/sec, and stored on magnetic tape.

The

data was then sent t o t h e IH-360 computer f o r smoothing and analysis by a program especially developed f o r t h i s purpose.

This procedure when a c t i -

vated during a rod drop from 50 in. gave about 1800 data points during t h e drop.

From t h i s velocity and acceleration curves were generated.

Data

taken before r e p a i r of t h e s t i c k i n g rod assenibly (R-4 and V-1 i n T-3) showed a region of near zero acceleration between 26 and 40 i n . , i n good agreement

with t h e incremental r e s u l t s shown i n Fig. 19.4.

After r e i n s t a l l a t i o n , all

three assemblies were t e s t e d by both incremental drops and t h e single drop computer analyzed method.

Both methods showed reasonably constant acceler-

ations of 10ft/sec2 o r more. i s shown i n Fig. 19.5,59

A t y p i c a l p l o t of acceleration vs position

337

338 The signal f i l t e r i n g and data smoothing that a r e required because of the noise that is inherent i n t h e position s i g n a l generator and transmis-

sion system made it impossible t o detect w i t h c e r t a i n t y a zero acceleration zone less than about s i x inches long i f it i s more than about s i x inches below the scram s t a r t i n g point.

Revised c r i t e r i a were established f o r t h e monthly t e s t i n g of t h e rods. This consisted of scramming t h e rods from 40 and 50 inches. These two points w e r e just above the normal operating positions of t h e r e g u l a t i n g The c r i t e r i a were t h a t t h e 50-inch scram t i m e must be less

and shim rods.

than 1.00 sec, the 40-in.

scram time must be less than 0.90 see, and t h e

acceleration curves must show no area which indicates abnormal drag.

an area of high drag was noted, t h e rod would be dropped from a point within t h i s area.

This scram t i m e must be no more than 20% longer than normal f o r

that point.

Prior t o a f u e l salt f i l l , t h e rods would be scrammed from 24

inches, which i s t h e i r normal position f o r a f i l l .

The scram t i m e s *om

t h i s point must be less than 0.7 seconds and t h e acceleration curve show no abnormality.

19 6 Maintenance Experience Repairs performed on t h e drive u n i t s usually consisted of a simple replacement of worn components which were readily accessible a f t e r t h e u n i t had been removed from t h e reactor c e l l .

I n t h e event of major d i f f i c u l t i e s ,

both a spare drive assembly and a spare control rodwere available as replacement items.

It w a s possible t o perform d i r e c t out-of-cell maintenance t o t h e drives without excess exposure t o personnel.

The radiation l e v e l was about 600

mR/hr a t t h e drive case ( t h e drive u n i t w a s removed from i t s case during

maintenance).

The overall l e v e l of radiation at t h e drive mechanism w a s

about 200 mR/hr and, as an example, t h e l e v e l a t t h e f i n e synchro transmitter which was mounted on the gear case w a s about 30 mR/hr.

Maintenance

on t h e control rod i t s e l f after c r i t i c a l i t y w a s very limited.

Activity on

t h e upper hose w a s about 400 mR/hr at contact and the section containing t h e poison elements was 20 R / h r a t t

18 in. a f t e r a two-year storage period.

339

Since t h e four drive assemblies were i d e n t i c a l and t h e spare p a r t s w e r e common t o all t h e drives, t h e maintenance problems regarding replacement of f a u l t y components was not complicated.

However, investigation as t o t h e

reasons f o r slow scram t i m e s , even when t h e areas of high drag were located (see 19.5)

,w a s

by trial and e r r o r method.

c e r t a i n areas such as t h e 7/16-in.-diam

Past experience had shown t h a t

a i r tube, on which t h e control rod

s l i d e s , could become s l i g h t l y @led o r bent.

These areas were examined,

adjusted, and t h e u n i t operated on a t e s t stand f r o m t h e main console.

The

f i n a l t e s t was t o reassemble t h e complete assembly i n i t s normal operating position i n t h e reactor c e l l and proceed w i t h t h e scram tests as described elsewhere.

19.7

Discussion and Recommendations

The o v e r a l l operation of t h e rods and drives w a s q u i t e good.

There

was only one time when a rod failed t o scram i n t h e four years of use.

no time w e r e reactor operations terminated s p e c i f i c a l l y f o r control rod maintenance.

Construction of t h e assemblies w a s r e l a t i v e l y simple since

very rapid response w a s not a requirement.

The flexible controls rods were

adequate but t h e p o s s i b i l i t y of binding, breakage, etc., always existed. It was possible t o control t h e rod position closely, within 0.1 inch, from

t h e "fine" position indicator on t h e console but t h e exact physical locat i o n of t h e poison elements was subject t o weaknesses inherent t o a semiconfined flexible m e t a l hose of t h i s type.

The only known position was at

t h e f i d u c i a l zero position i n t h e thimble t o which all other positions were related. It required about 12 hours t o open t h e reactor c e l l access t o approach

t h e control rod assemblies , Location o f t h e drives i n a

a c t u a l removal required about one hour. r e a d i l y accessible location i s recommended.

The manner of attachin

control rods t o t h e drives and t h e drive

u n i t s t o t h e thimbles w a s h

ed by ( a ) poor l i g h t i n g a t t h e b o l t loca-

t i o n , (b) lack of space t o i n s e r t more r i g i d t o o l s t o perform t h e b o l t i n g which w a s done from a distance of about 18 f e e t , and ( c ) poor v i s i b i l i t y f r o m t h e t o o l operator position at t h e maintenance shield.

A modified

340 external, r a t h e r than i n t e r n a l , b o l t e d flange attachment between t h e rod housing and thimble which is c l e a r l y v i s i b l e i s recommended. The "fine" position cynchro t r a n s m i t t e r s and potentiometers, such as were used i n these u n i t s should be of a more durable type f o r extended service.

341

FREEZE VALVES

20.

M. Richardson 20.1

Introduction

The flow of salt i n t h e MSRE drain, f i l l , and process systems was cont r o l l e d by freezing and thawing short plugs of s a l t i n f l a t t e n e d sections of l-l/2-inch

pipe, c a l l e d "freeze valves."

This method of control was Although

adopted because of a lack of a proven r e l i a b l e mechanical valve.

mechanical type valves would have had t h e advantage of f a s t e r action and t h e a b i l i t y t o modulate flow, t h e "freeze valve" concept had a good record of s a t i s f a c t o r y performance, t h e freezing and thawing times were s a t i s f a c t o r i l y s h o r t , and t h e "off-on" type control did not impose any p a r t i c u l a r hardship.

~~ Design and development of t h e MSRE freeze valves began i n 1 9 6 ~ 1 .The basic design was established and had been t e s t e d by t h e time t h e MSRE cons t r u c t i o n had begun.

However, considerable e f f o r t was expended at t h e re-

actor s i t e before t h e valves were ready f o r routine operation.

Successful

freeze valve operation i s t o t a l l y dependent on t h e manner i n which t h e heat and/or coolant gas i s applied and controlled,

The e f f e c t of t h e heaters,

type of s e r v i c e , l o c a l environment, e t c . , made it necessary t o ' h n e " each of t h e valves f o r i t s s p e c i f i c location.

This e f f o r t continued u n t i l

criticality.

20.2

Description of t h e Design of t h e Freeze Valves

The valves were located i n t h e system as shown i n Fig. 20.1. twelve valves were similar arid were made of 1-1/2-in.

All

Sched-40 INOR-8 pipe.

The valve body, o r f l a t t e n e d section, w a s cold-formed t o m a k e a 2-in.-long f l a t , i2-in.

wide, with a 1/2-in.

i n t e r n a l thickness o r flow area.

A cool-

ing gas shroud w a s welded around t h e valve f l a t t o d i r e c t t h e gas f o r freezing t h e salt. was 1/2-in. shroud.

The annulus formed between t h e valve body and shroud

The 3/4-in.

gas i n l e t and exhaust l i n e s were welded t o t h e

See Fig. 20.2.

A p a i r of thermocouples located a t t h e center of t h e valve inside t h e

a i r shroud, and a p a i r upstream and downstream j u s t outside t h e a i r shroud

34 2

ORNL-DWG 73-605 TO COOLANT SALT DRAIN TANK

REACTOR

TO AND FROM COOLANT SYSTEM FV-204

FV-206

COOLANTSYSTEM FILL AND DRAIN LINES

'FV-104

FFT

FD-2

FD-1 V4IFFERENTIA.L CONTROLLER

FUEL SYSTEM FILL DRAIN AND TRANSFER LINES

Fig. 20.1

Location of Freeze Valves

omL4m3.

ALL NOT

THERMOCOUPLES SHOWN.

THERMOCOUPLES

Fig.

20.2

Typical

Freeze Valve

84-68$g

344 were connected t o the control modules and t o t h e cooling a i r modulating controller.

A s i n g l e thermocouple spaced

5 i n . from t h e valve c e n t e r l i n e ,

upstream and downstream on t h e top side of t h e 1-1/2-in. information f o r s e t t i n g t h e manual heater controls.

pipe, supplied

The heaters on e i t h e r

side of t h e valve were individually controlled. The thermocouples at t h e valve body were connected t o an E l e c t r a Sys-

tems modular control network ( s i x modules per valve) which operated automatically at preset temperatures t o maintain each valve within i t s specif i e d condition.

The control network of t h e c r i t i c a l valves, described

below, included an automatic cooling a i r flow c o n t r o l l e r which maintained the valve center a t a constant temperature.

The ultimate control, however,

remained with t h e modules. 20.3

Description of t h e Operation of t h e Freeze Valves

Valves were usually operated i n one of t h r e e steady-state conditions o r modes. (1) Deep-frozen

- t h e heaters were adjusted t o maintain t h e valve a t 40W0O0F without cooling a i r .

(2) Thawed

-The

heaters were adjusted t o maintain t h e valve at

'U200째F without cooling air.

(3) Frozen

-The

heaters remained as i n t h e thawed condition but

the cooling gas flow w a s automatically adjusted t o

hold t h e frozen valve i n condition f o r a rapid thaw. Operation of t h e valve required only a hand switch ( o r control a c t i o n ) t o freeze o r thaw t h e salt. 11

When t h e valve w a s switched t o "thaw" from t h e

frozen" condition, t h e air was cut o f f and t h e heat supplied by t h e shoul-

der headers would m e l t t h e salt i n t h e valve body.

Similarly, when t h e

valve was switched t o t h e "frozen" from t h e "thawed" condition, t h e a i r was turned on and t h e salt i n the valve body would freeze. Each valve had t h r e e coolant gas flow conditions: o r reduced flow :(hold). t u r e control modules.

o f f , full on ( b l a s t ) ,

The flow condition w a s determined by t h e temperaAlthough the exact module s e t p o i n t s varied, f o r t h e

purpose of explanation, t h e following conditions a r e used:

FV-103 control. )

(See 20.4.4 f o r

345

Settinq

Module No.

FV-lAl FV-lA2 FV-2A1 FV-2A2

850°Ft, 8 0 0 O ~ ~

FV-3A1

850°F+, 8 0 0 O ~ ~

Shoulder temperature range

700°F+

Low shoulder temperature

500°F

Center temperature

-- low alarm

1300°F+

Center temperature

-- high

alarm

Shoulder temperature range Low shoulder temperature

700°F+

Fv-3A2

Function

When it was switched t o freeze, Assume t h a t t h e valve was thawed at 1200'F. the b l a s t air ( ~ 2 4 scfh) came on. When both valve shoulder temperatures decreased t o 800°F, t h e "blast" a i r was cut o f f and t h e "hold" a i r automatically came on and held t h e valve shoulder temperatures between 8 0 0 ~ ~

. t h e shoulder temperatures continued downward f o r any reason, and 8 5 0 ~ ~If all a i r w a s cut o f f at 7OO0F by modules lA2 o r 3A2.

If hold a i r flow was

i n s u f f i c i e n t and e i t h e r of t h e shoulder temperatures exceeded 85OoF, t h e b l a s t a i r automatically switched on by t h e action of t h e lAl o r 3A1 module and reduced t h e temperature t o 8 0 0 ~ ~ . On those valves used most frequently, FV-103, 105, 106, 204, and 206, a temperature controlled d i f f e r e n t i a l a i r flow adjustment was provided f o r t h e hold a i r i n addition t o t h e overriding on-off module control.

These

c o n t r o l l e r s maintained the valve center a t a constant temperature. The normal valve condition during r e a c t o r operation w i t h f u e l s a l t was as follows:

(See Fig. 20.1.)

u t i n t h e event of an emergency drain, it was required t o thaw i n less than 1 5 minutes. (2) FV-105 and FV-106 were thawed and were required t o remain thawed (1) FV-103 was frozen

even i n t h e event of a power outage.

I n t h e event of an emergency f u e l

drain, t h e s a l t flowed t o both f u e l drain tanks through FV-105 and FV-106. A normal drain required freezing one of these valves p r i o r t o t h e d r a i n t o

direct the s a l t t o the s (3)

cted receiver tank.

FV-104, and FV's 107 t o 112 were deep-frozen.

(4) FV-204 and FV-206 were frozen during coolant s a l t c i r c u l a t i o n . These valves were required t o thaw rapidly t o prevent t h e coolant salt from freezing i n t h e r a d i a t o r tubes. -

... 346

During flush salt operation, FV-103 was frozen and FV-104 was thawed,

All other valves were deep-frozen.

FV's lo7 t o 112 were used only during shutdown periods f o r salt transfers and additions. The thaw and freeze time f o r these valves w a s not important.

When not i n use, they were frozen o r deep-frozen depending on t h e

operation i n progress.

20.4 ..

Operating Experience

After t h e i n i t i a l d i f f i c u l t i e s described below, operation of t h e .

freeze valves was quite satisfactory. . .

Table 20.1 i s a tabulation of t h e

fYeeze and thaw cycles f o r each of t h e valves.

Table $20.1 Freeze Valve Freeze-Thaw Cycles

Freeze Valve No.

No. of Cycles

103 104

105

46 97

106 107

108

109 110

53 14

111

112

204

206

The valves which were required t o melt rapidly, less than 1 5 minutes,

were timed whenever t h e system w a s drained.

These valves were FV-103, FV-

204, and FV-206. Periodically they were t e s t e d under complete power failure conditions t o assure t h a t t h e valves would melt within t h e maximum tlme allowed.

The normal thaw time without power f o r fi-103 was 9-41 min, f o r

FV-204 and FV-206 t h e normal thaw t i m e was 12-14 minutes.

347 During e a r l y operation, FV-104, FV-105, and FV-106, which were required t o remain thawed (>850째F) f o r %30 minutes, were a l s o t e s t e d under A t an i n i t i a l temperature of 1150째F,

t h e same power f a i l u r e conditions.

t h e s h o r t e s t period of time required t o reach t h e s a l t freezing point was

36 minutes, with no salt flowing through t h e pipe. It required approximately 8 hours t o bring a valve out of t h e deepfrozen condition (40wOO0F) i n t o t h e frozen condition.

Heat was applied

s e l e c t i v e l y during t h e heatup t o insure t h a t t h e salt w a s melted progress i v e l y from e i t h e r end towards t h e center of t h e freeze valve.

This was

t o avoid having molten salt trapped between two frozen plugs and possible The siphon pots were interlocked s o t h a t t h e pot.

danger of pipe rupture.

temperature was required t o be g r e a t e r than 95OoF before t h e associated valve could be thawed. 20.4.1

Testing of a Frozen Valve

After a valve had been frozen, it was pressure t e s t e d t o assure t h a t a s o l i d , l e a k t i g h t frozen plug had been obtained.

Early t e s t i n g revealed

t h a t due t o t h e f u e l drain l i n e geometry, it w a s possible t o have insuf-

f i c i e n t salt remaining i n t h e freeze valve a f t e r a drain t o make up a s o l i d Investigation, using a glass pipe model, demonstrated t h a t after t h e drain l i n e had been blown down through t h e head of s a l t i n t h e drain tank, t h e salt remained i n the siphon pot and would not flow back i n t o t h e valve body t o f i l l t h e void.

After blowing down l i n e 103, a 4 o r 5 p s i AP e x i s t -

ed between t h e drain tank and r e a c t o r , when attempts were made t o equalize

t h i s AP, t h e salt w a s force

f t h e pot and past t h e freeze valve back

i n t o t h e 103 l i n e . The method which was

t successfully w a s t o reduce t h e r e a c t o r

system pressure by %1/2 p s i g which usually allowed s u f f i c i e n t flQw-back t o f i l l t h e valve and make a good s e a l .

A similar procedure woz-ked s a t i s f a c -

t o r i l y for the transfer free 20.4.2

Control Modules

During t h e i n i t i a l chec sequent operations and t e s t s

of freeze valve' control c i r c u i t s and ~uboblems were encountered i n t h e operation

of t h e modules which make up t h e E l e c t r a Systems Corporation alarm

(ICJ

348 monitor system,

The setpoints d r i f t e d o f f t h e pre-se,

tempera-ure t r i p

value, double t r i p points occurrea, and failures t o respond t o alarm*signals were encountered. Failures and malfunctions of t h e modules were t r a c e d t o i n f e r i o r quali t y components and corrclsion o r oxidation a t t h c p r i n t e d c i r c u i t board conThese faults were mostly eliminated by replacement of f a u l t y com-

tacts.

ponents, gold-plating a l l module contacts, and minor c i r c u i t changes.

The

setpoints f o r the 72 modules were checked a t l e a s t once p e r year thereafter.

I n general, t h e modules remained within 15% of the p r e s e t values

after t h e i n i t i a l d i f f i c u l t i e s were.resolved. 20,4.3' Heat Control t o Freeze Valves

As i n s t a l l e d , t h e heaters on e i t h e r s i d e of t h e freeze sections were on a common control. Also t h e heat supplied t o t h e lower section of t h e siphon pots adjacent t o t h e freeze valve heaters w a s marginal due t o t h e manner of i n s t a l l a t i o n .

Therefore, a balanced temperature gradient on ei-

t h e r s i d e of t h e freeze plug was not obtained.

Since t h e s e t p o i n t s f o r

.the shoulder temperature control modules were adjusted f o r a balanced heat d i s t r i b u t i o n , it w a s d i f f i c u l t t o maintain the valves within t h e module

limits.

Separate heater controls were added which permitted separate con-

t r o l of t h e heat t o each s i d e of t h e freeze plug,

The h e a t e r wattage of

< t h e shoulder heaters was increased from 1 5 W/in2 t o 30 W/in2.

These added

features relieved most of t h e above d i f f i c u l t i e s . ?

The proximity of FV-105 t o FV-106 (35 inches a p a r t ) closely related t h e temperature e f f e c t s of one valve t o t h e other beyond the limits of good control.

This could only be corrected by a change i n t h e piping.

Since

t h i s was not p r a c t i c a l , t h e problem e x i s t e d throughout t h e r e a c t o r operation.

This was especially t r u e during a salt f i l l from FD-I o r FD-2 since 7

t h e hot salt i n l i n e 103 w a s 19-3/4 in. from FV-106 o r 15-1/4 i n . from Fy-105. 20.4.4

Comments on Operation of Individual Freeze Valves

"FV-103

- Since t h i s was t h e main f u e l salt drain freeze valve,

was required t o thaw i n less than 1 5 minutes. A coolant gas f l o w r a t e of 75 scf'm was .available f o r freezing. Heat w a s supplied t o t h e valve from t h e ambient temperature within t h e reactor furnace and therefore no.valve heaters were required.

Operation of t h i s valve w a s complicated by a dual

349

bi'

(1)Immediately following a r e a c t o r f i l l , t h e

set of operating conditions:

line-103 on t h e downstream s i d e of t h e valve remained full of s a l t f o r a period of several hours; (2) Line-103 w a s then emptied of salt leaving no s a l t on t h e drain tank side.

Figure 20.3 shows t h e temperature distribu-

t i o n f o r both conditions. I n order t o meet t h e fast thaw requirement and s t i l l meet both above conditions, t h e module control points were set t o f'unction as l i s t e d below: FV-103-lAl

-- Blast

FV-lO3-1A2

-- Low &temperaturea i r cut

off

temperature). FV-103-2A2

- (shoulder

a i r on at 10IO°Ff, no h y s t e r e s i s temperature. )

-- High alarm 6 0 0 ~ ~ "2 1 . of

- 465OF+ - (shoulder

3" alarm matrix (center

temperature).

FV-lO3-3A1

-- Blast

FV-103-3A2

-- Low temperature

a i r on at 765OFf (shoulder temperature )

.- 71g°F+, 50°F h y s t e r i s i s ,

a i r cut o f f

- 515'F+,

(shoulder

temperature).

The "hold air" d i f f e r e n t i a l a i r flow c o n t r o l l e r w a s normally set t o hold t h e center temperature (TE-FV-103-2Al)

a t 375 t o 40O0F.

The temperature adjustments required f o r t h e proper valve operation r e s u l t e d i n only one shoulder control module being e f f e c t i v e i n each cond i t i o n of t h e valve shown i n t h e figure. t h e TS-FV-103-3Al

During t h e i n i t i a l freeze cycle,

w a s t h e p r i n c i p a l c o n t r o l and after t h e line-103 w a s

drained of salt, t h e TS-FV-103 t h e r e w a s no s a l t below TS-

perature responded very sl on t h a t ' s i d e of t h e valve.

w a s t h e key temperature c o n t r o l since

3-31. During a thaw cycle, t h e IAl tence t h e mass of heat ( r e a c t o r v e s s e l ) w a s '

The 3Altemperature respo

since t h e source of heat w a s through t h e frozen plug perature.

t h e a&ient ten-

Selection of these temperature a l a r m points w a

after many tests. Once t h e valve c o n d i t i

were established, FV-103 o

cept for one unscheduled drain which occurred on April 15, 1969.

"his w a s

caused by an upward 5O0F drift of t h e s e t p o i n t of t h e FV-103-3A1 module from 75OoF t o 8 1 0 plus ~ ~ an administrative change of t h e FV-103-lAl

point from 10IO°F t o 1050°F,

set-

.ORNL-DWG

73-806

5l-

800

600

o NORMAL FREEZE, NO SALT IN LINE i o 3 ON DRAIN TANK SIDE

a K

- 9-11

min MELT -

0 VALVE FROZEN, SALT BOTH SIDES OF FREEZE PLUG

400 200

I 8- 1

A1 B

I R27

- 30

TE NUMBERS

Fig. 20.3

Temperature Distribution at FV-103

351

The upward d r i f t had occurred p r i o r t o t h e previous freezing operation and a short plug (<2 i n . ) of salt w a s formed i n t h e valve. and freeze t e s t indicated a good freeze.

The controls

A t t h e time of t h e incident, t h e

salt temperature was not at 1210'F f o r which t h e controls were o r i g i n a l l y adjusted, but at lllO'F with t h e cooling gas flow controlling t h e valve center temperature at 420'F. Figure 20.4 i s a graphic h i s t o r y of t h e incident and it can be seen t h a t , with t h e FV-103-lA1 adjusted t o 1050째F, t h e r e w a s l i t t l e i f any warning t o t h e operator p r i o r t o t h e melt. "E-FV-103-lAl

was about 1030'F1040'F

the reactor outlet.

"he valve normally melted when

when t h e f u e l salt w a s a t 1210'F a t

The lA1 module administrative change t o 1050'F w a s

based on an untested theory t h a t the valve should be adjusted on t h e rea c t o r i n l e t salt temperature r a t h e r than t h e o u t l e t . valve had a normal 2-in.

As long as t h e freeze

freeze plug, t h e cooling a i r could be adjusted t o

maintain t h e valve within t h e normal operating range.

I n t h i s instance,

due t o short plug, t h e valve melted as a r e s u l t of changing t h e cooling

a i r flow which i n i t i a t e d a heating cycxe' beyond the'rahge of t h e cooling a i r t o control.

The a i r flow was routinely adjusted 50'F upward when t h e

fuel salt was at a reduced temperature but t h e combination of r a i s i n g t h e 1Al alarm point, t h e short plug r e s u l t i n g f r o m t h e 3 A l upward d r i f t before

t h e previous freeze, and r a i s i n g t h e cooling setpoint r e s u l t e d i n t h e f u e l drain. FV-104

- This valve operated without d i f f i c u l t y since normally it

was deep-frozen o r thawed. It was s u f f i c i e n t l y spaced'away from t h e l i n e 103 so as t o - b e unaffected by a fuel drain. See Fig. 20.1. FV's 105 and 106 -FV-lOS and FV-106 were d i f f i c u l t t o control during a f i l l o r drain operation because of t h e proximity of t h e valves t o each

other and t o t h e 103 l i n e which was common t o both freeze valves. Fig. 20.1.

See

Heat conduction from t h e hot s a l t passing through line-103

strongly affected t h a t valve which was t o remain frozen.

I n an emergency

drain s i t u a t i o n , both valves were thawed and t h e f u e l would drain t o both tanks.

A scheduled drain required t h a t one of these valves be frozen t o

d i r e c t t h e salt t o t h e selected-drain tank only.

I n severaJ, instances

during a scheduled reactor drain t h e frozen valve thawed which r e s u l t e d

352

ORNL-DWG 73-607 SALT IN REACTORVESSELAT 111PF

SIMULTANEOUS ALARM AND THAW-SALT FLOWING THROUGH 1/2 in. EMERGENCY DRAIN TUBE ALVE CENTER TEMPERATURE RISING, COOLANT GAS LOW INCREASE NOT FAST ENOUGH TO CONTAIN HEAT-UP

TIME

Fig. 20.4

Temperatures During Unplanned Thaw of FV-103

353

i n an added s a l t t r a n s f e r operation t o put a l l t h e salt i n t o t h e proper The condition was most troublesome during a r e a c t o r f i l l operation,

tank.

a time-consuming function a t b e s t , when t h e frozen valve would go i n t o a l a r m and control interlocks would s t o p t h e f i l l o r t h e freeze-valve thaw with t h e system p a r t i a l l y f i l l e d and d r a i n t h e salt t o both storage tanks. By proceeding very c a r e n i l l y during t h e i n i t i a l stages of t h e f i l l t h i s

condition could, and usually was, overcome by stopping t h e f i l l w i t h t h e 103 l i n e f i l l e d t o above t h e t e e and allowing t h e frozen valve t o approach equilibrium.

However, the last f a i l u r e of t h i s s o r t occurred as l a t e as

April 11, 1969. A leak i n t h e primary system became evident a f t e r t h e f i n a l system

drain of December 12, 1969 which appears t o be i n t h e v i c i n i t y of FV-105. This i s discussed i n Section

5. were no d i f f i c u l t i e s o r unscheduled

FV's lo7 through 112 - m e r e thaws with these t r a n s f e r valves.

FV's 204 and 206 -The

CiJ

coolant valves have t h e same inherent fault

as FV-lo5 and FV-106 of being too close t o each other.

Since t h e valves

were always operated as a p a i r , e i t h e r frozen o r thawed, t h i s did not cause any d i f f i c u l t y .

There were no operating d i f f i c u l t i e s with these valves

after criticality.

1. Freeze valve

so located by distance o r

separate -lines so as 2.

interaction.

They should be s

ioned t h a t t h e adjacent piping i s posi-

t i v e l y full of salt.

l e d i n a v a r i e t y of ways,

v e r t i c a l l y , inclined,

e no void spaces i n the piping.

Those valves

r i t i c a l should be

r o l feature.

nating t h e automatic expensiv9, less time-cons

l i f i e d by elimi-

These valves would be l e s s

t o a u u s t , and e a s i e r

control by simple

manual air control.

The . aut omat i

sidered f o r a l l c r i t i c a l valves.

s c o n t r o l l e r should be con-

This f e a t u r e g r e a t l y reduced t h e amount

354

of time required t o maintain the valves.within'limits and reduced t h e post-freeze thermal cycling.

5 . The shakedown period f o r adjustment of t h e valve temperature cont r o l s at the reactor was time-consuming f o r a variety of reasonsp I n i t i a l l y the control modules were found t o d r i f t o f f setpoint and had double t r i p points.

These d i f f i c u l t i e s were e s s e n t i a l l y , but not completely, elimi-

nated by l o c a l redesign,

There were s i x control modules per valve f o r a

t o t a l of seventy-two individual modules t o be maintained.

Each module con-

t a i n s two indicating lamps, "Alarm" and "normal" which a r e wired i n s e r i e s with the control c i r c u i t .

There were frequent f a i l u r e s of these bulbs

which i n turn affected the module function,

Many of t h e operators had dif-

f i c u l t y i n understanding t h e functions of t h e modules and were at a l o s s as t o what corrective action was t o be taken when an alarm sounded.,

Al-

though the Electra modular system succeeded i n controlling t h e valves, a simpler system i s recommended,,

6. Individual control should be providedfor each of t h e heaters which a f f e c t t h e valve temperatures.

An excess of heat i s needed t o overcome heat

losses i n t h e valve areao Placement of t h e heaters should be given caref'ul consideration t o eliminate cold spots i n ad3acent piping.

70 valve.

An adequate supply of cooling gas should be available t o each A supply of 4050 scfm f o r 1-1/2-in.

pipe "blast air" cooling

would be s u f f i c i e n t f o r good control.

8. The modular control setpoints were established f o r normal 1210째F reactor operating temperature at steady s t a t e ,

The c r i t i c a l "fast thaw"

valve control setpoints were adjusted f o r t h i s temperature as was t h e cont r o l point f o r t h e automatic cooling gas control.

Deviation f r o m t h e nor-

m a l temperature f o r which the control points were adjusted (within 75OF) required ad3ustment of t h e cooling air flow only.

This raised o r lowered

the valve center temperature t o maintain t h e shoulder temperature control modules within limits,

During t r a n s i e n t conditions, such as heatup of t h e

system, t h e valves were severely cycled because of a l a r g e temperature gradient between two valve shoulders on t h e same valve.

One shoulder would

be above the module alarm point and t h a t control would t u r n on t h e b l a s t

air.

The other shoulder would be below t h e low temperature alarm point

and cut a l l the cooling a i r o f f .

The module controls would allow the

355

valve t o cycle back and forth across t h i s range u n t i l t h e lâ&#x20AC;&#x2122;ow temperature shoulder would heat enough (*70O0F) t o permit t h e d i f f e r e n t i a l controller t o control the a i r flow.

In practice all a i r w a s manually turned o f f

during heatup.unti1 t h e valve shoulder temperatures were above 70O0F.

The

valve then would remain i n t h e frozen condition but heat much more rapidly and t h e thermal cycling would be greatly reduced.

356

21.

FREEZE FLANGES

R. H, Guymon I

21,l

Description

Mechanical-type j o i n t s were provided i n t h e 5-in.

salt piping inside

the reactor c e l l t o permit major components t o be removed f o r maintenance. Figure 21.1 i s a sectional view of t h e so-called "freeze flange" type of The design was such t h a t t h e "0" r i n g was i n a r e l a t i v e l y cool

j o i n t used. location.

A frozen salt s e a l protected t h e r i n g j o i n t seating surfaces

from contact with sal-t which could cause corrosion when t h e s a l t w a s exposed t o moisture.

Each freeze flange had s i x thermocouples; four were

located on a 4-1/2-in. on a 10.2-in.

radius, 9,0째 apart, one on a 7-in.

radius and one

radius. 21.2

Operat ion

It w a s never necessary to.remove a major component and therefore one

important function of the freeze flanges was never t e s t e d at t h e E R E . E a r l i e r tests on a prototype flange indicated t h a t making and breaking

these could be done remotely without appreciable d i f f i c u l t y , The leak rates of t h e five freeze flanges during e a r l y operation are

given i n Table 21.1.

Similar leak r a t e s were observed,throughout operation

except t h a t one of t h e freeze flanges (FF-201) located near t h e point where t h e coolant salt l e f t the heat exchanger leaked more than the allowable leak r a t e o f 1 x

cm3/sec when salt was not i n t h e piping and t h e cool-

ant system hpd been cooled down., cm3/sec on August 2, 1966.

This leak was measured at about 0.2

The flange always became acceptaply sealed when

the system w a s heated. Although there w a s some variation between individual flanges and some variation w i t h time, the thermocouples located 4-1/2 ine' from t h e centerline of t h e pipe were usually between 750 and 95OoF0 Those on 7-in. r a d i i were about 550 t o 65OOF and those at 10.2-in.

were about 450 t o 55OoF.

The thermocycle history i s shown i n Table 21.2.

When it was decided

t o extend t h e operation of the MSRE longer than t h e o r i g i n a l plans, t h e

357

ORNL-LR-DWG 63248R2

)

BUFFER CONNECTION (SHOWN ROTATED) MODIFIED R-68 RINGGASKET

FROZEN SALT SEAL-

-4r, in.+

5-in. SCHED-40 PIPE

ze Flange and Clamp

358

Table 21.1

Observed Leak Rates of Buffer Gas from Freeze Flanges

Leakage .Rahe (E&

cm3/gec) ?

Freeze Flange

Initial Heatup

Circulating

salt

10-3

System H o t DraYned After Run 1 1

x 10-3

io-3

System Cold After . Run 1 x 3 ' 0 1

System Cold After

Run 3 x 10-3

0.57

0.7

1.5

2,o

10 3

0.4

0.25

2.26

1.2

102

0*5

0.3

0.30.

2.33

1.0

200

1.0

0.21

0.41

1.48

201

0.6

0.22 .

0.21

1.23

100

200

101

003

1.8

freeze flange t e s t loop whibh had cycled a prototype flange f o r 103 cycles

was restarted.

Af%er cycle 268, examination using dye-penetrant revealed

a crack i n t h e bore i n t h e ' v i c i n i t y of t h e weld attaching t h e alignment

stub t o t h e face of t h e flange.

The crack appeared about t h e same a f t e r

t h e f i n a l thermal cycle, number 540. 21.3

Conclusions I

The freeze flanges performed s a t i s f a c t ' o r i l y throughout t h e r e a c t o r operation.

The higher leak r a t e on FF-201when cold caused some concern

but did not require repairs.

359

Table 21.2.

Summary of Unscheduled Scrams a t MSRE with Reactor Criticala

Operatinp Hours

Year

Quarter

Core in

Critical

1966

1 2 ' 3 4

672 1293 554 1266

62 1070 1221

1 2 3

1861 1254 1318 2159

1 2 3 4 1 2 3 4

1967

1968

1969

Tot a1

amere

Number of Unscheduled Rod Scrams Human Power Total Error Failures L & C Otherb 0 3 2 1

1 6 0

2 2 0 1

1 2 0 0

1852 1186 1292 2144

2 2 1 2

1 1 0 0

0 1 1 1

1 0 0 1

0 0 0 0

2048 0 88 1000

2045 0

0 0 0 1

0 0 0 0

1850 1385 1076 1203

1800 1375 1054 1176

2 2 0

0 0 0 0

0 1 0 0

0 0 0 0

0 2 2

19027

17425

413

0 735

4 13 2

is no record of any unscheduled s c r a m s during 1965, w h e n f u e l e reactor was c r i t i c a l ( a t 1 kW o r less)

was i n t h e core f o r 1062 h f o r 230 hr.

% o s t l y equipment fau Fgr example, f i v e of t h e . l a s t s i x scrams due t o "other" causes occurred when t h e speed of t h e variable-frequency generator being used t t o d r i v e t h e f u e l pump sagged below a prescribed l i m i t .

360 22.

CONTAINMENT

P. H. Harley 22.1

Description and Criteria

Containment a t t h e MSRE w a s required t o be adequate t o prevent t h e escape of multicurie amounts of r a d i o a c t i v i t y o r dangerous amounts of other materials t o t h e atmosphere.

A minimum of two b a r r i e r s w a s provided,

During operation, t h e primary b a r r i e r was t h e piping and equipment which contained o r w a s connected t o t h e f u e l s a l t system and off-gas system.

Block valves were i n s t a l l e d i n a l l t h e helium supply (cover gas) Primary containment extended out t h e .. l i n e s and through t h e charcoal beds. The e n t i r e primary system

l i n e s t o prevent backup of a c t i v i t y . off-gas

was of all-welded construction with leak-detected flanged J o i n t s except at a f e w l e s s vulnerable locations where autoclave f i t t i n g s w e r e used. Ess e n t i a l l y zero leakage w a s permissible from t h e primary system. The secondary containment currounded t h e primary containment,,

It con-

s i s t e d mainly of t h e reactor and drain-tank c e l l s and appendages t o them, There are ~ 7 0 0penetrations f o r service l i n e s i n t o t h e c e l l s .

Figure 2 2 , l

is a simplified diagram showing t h e various type penetrations and methods of s e a l i n g o r blocking these. Under conditions of t h e MCA, t h e secondary containment would be limited t o a pressure of 39 p s i g by t h e vapor condensing ( s e e 22.4).

The c e l l

temperature would r i s e t o 2 6 0 ~Under ~ ~ t h e s e conditions t h e maximum a l lowable leakage rate as specified i n t h e MSRE Safety Analysis ReportSs was 1% per day of t h e secondary containment volume.

During maintenance, t h e reactor was shut down and t h e amount of radioac'tivity available f o r t h e chances of i t s r e l e a s e were considerably reduced, When t h e primary system w a s opened, t h e c e l l s became t h e primary containment with t h e high-bay area acting as secondary containment. 22.2

Methods Used t o Assure Adequate Containment and Results

Most of t h e time t h e primary and secondary containment w a s operated well below i t s design c a p a b i l i t i e s .

Various means were used t o assure t h a t

containment would be adequate -under t h e worst conditions e

ORnnr;. 8-2608

Fig. 22.1

Schematic of MSRE Secondary Containment Showing Typical Penetration Seals and Closures

362

The primary system was pressure-tested annually and t h e c e l l a i r ac-

t i v i t y w a s continuously monitored as an indication t h a t no leak had developed during operation.

The secondary system l e a k r a t e w a s checked annually

at a p o s i t i v e pressure and was continuously monitored at normal c e l l pressure (+ p s i g ) during operation.

I n addition t o t h i s , a l l primary and

secondary system block valves w e r e t e s t e d annually t o assure acceptable leak rates. 22.2.1

Primary System

Radiographs, dye checks, and other inspections were c a r e w l y reviewed A l l flanges had metal "0" rings

during construction of t h e primary system.

which were remotely leak-detected by pressurizing t h e r i n g groove t o 100 psig w i t h helium.

(See Section 1 4 of t h i s report. )

Although t h e system

was opened 33 t i m e s f o r f u e l additions, graphite sampling, and maintenance as indicated i n Table 22.1, there was no leakage of any flange above the allowable

loo3

cc/min while t h e reactor w a s operating.

I n 1965 before s t a r t i n g nuclear operation, a l l of t h e block o r check

valves were t e s t e d i n place o r removed and bench-tested.

This w a s p a r t of

t h e containment (primary and secondary) check l i s t , Section 4E of t h e MSRE Operating Procedures.22

The s p e c i f i e d maximum l e a k rate on t h e primary

system valves was 1 t o 2 cc/min at 20 psig. helium cover-gas l i n e s had excessive leakage.

A nurnber of valves i n t h e

This w a s due t o damaged "0"

rings and/or they had foreign p a r t i c l e s i n them, usually metal chips from machining.

After cleaning and i n s t a l l i n g new "0" rings they had no detect-

able leakage.

A f t e r r e i n s t a l l a t i o n , t h e l i n e s were pressurized and t h e

f i t t i n g s o r welds were checked with a helium leak detector. These tests were repeated annually.

The r e s u l t s a r e given i n Table

22.2 along w i t h results from t e s t s of t h e secondary block valves which a r e described later. I n addition t o t h e above, an annual strength t e s t was run on t h e p r i -

mary system. This w a s normally done by pressurizing t h e f u e l system, including d r a i n tanks, t o 60 p s i g with f l u s h s a l t c i r c u l a t i n g at normal opera t i n g temperature (1200OF). *No abnormalities were noted during any of these t e s t s .

(cd

363 Table 22.1

Opening of MSRE Primary Containment Flanges

No. of Times

Location

Primary Reasons

Off-gas l i n e at f u e l pump Off-gas l i n e i n vent house

6 7 4

Fuel pump vent l i n e

Overflow tank vent

Plugged valve, access t o a lower line

Drain tank No. 2 vent

Fuel addition

Drain tank No. 2 access

Fuel addition, samples, inspection

Drain tank No. 1 access

Inspection

DT t o FP equalizer

Plugged c a p i l l a r y

FP upper off-gas l i n e FP r o t a r y element FP l e v e l reference

T e s t o i l catch tank

1 1

Remote p r a c t i c e , inspection

Reactor graphite sampler

Line r e s t r i c t i o n Valve, f i l t e r , p a r t i c l e t r a p .Plugged valve, modify leak detector

Restricted a f t e r o v e r f i l l

Tot a1

Table 22.2

Water A i r or C e l l A i r

Simples, core inspection

3-5 cc/min

Tests of Block Valves

3 64 -

There were no leaks i n t h e primary system u n t i l a f t e r t h e f i n a l drain.

b) >

A t t h i s time a leak w a s indicated by an increase i n c e l l - a i r a c t i v i t y .

Subsequent investigation showed t h a t t h i s w a s i n t h e v i c i n i t y of one of t h e drain freeze valves (FV-105). 22.2.2

See Section 5 and Reference 25*

Secondary Containment

The c r i t e r i a f o r t h e secondary Containment was not as r i g i d as f o r t h e primary containment.

Small leaks could be t o l e r a t e d but it w a s s t i l l

necessary t o assure t h a t containment was adequate f o r t h e worst conditions The methods used a r e described below. Only one strength t e s t was made on t h e r e a c t o r and drain tank c e l l s , This was i n 1962, soon a f t e r construction 'of t h e c e l l s was completed,

The

tops of t h e c e l l s were closed temporarily by s t e e l membranes and t h e penet r a t i o n sleeves were temporarily blanked o f f .

The c e l l s were then hydro-

s t a t i c a l l y t e s t e d separately at 48 p s i g (measured a t t h e t o p of t h e c e l l s ) . A leak rate acceptance t e s t w a s a l s o made i n 1962.

The leak r a t e was ac-

ceptable, however, none of t h e penetrations had been i n s t a l l e d by t h i s

t i m e e 56 P r i o r t o power operation, t h e containment block and check valves were t e s t e d i n place or-removed and bench-tested as described i n t h e containment s t a r t u p check list (4E of t h e Operating Procedures) 22 The s p e c i f i e d maxi-

mum leak r a t e at t h e t e s t pressure of 20 p s i g varied according t o t h e location of t h e valve but usually was set a t 1 t o 2 scc/min f o r helium cover gas (primary containment.), 5 t o 10 scc/min f o r other gases and 3 t o 5 cc/min f o r l i q u i d s .

A considerable number of leaks were found and repaired.

Leaks which occurred i n t h e water system were mostly i n t h e check valves.

These were t h e r e s u l t of foreign p a r t i c l e s , apparently washed out

of t h e system and trapped between t h e f i x e d and moving p a r t s of t h e valves.

Two hard-seated valves .in t h e water system had t o be replaced with s o f t seated valves even though t h e seats were lapped i n an e f f o r t t o get them t o seal.

One w a s a swing check valve i n a water l i n e between t h e surge

tank and t h e condensate tank.

The o t h e r was a spring-loaded, hand-operated

vent valve on t o p of t h e surge t p k . The instrument-air block valves with few exceptions were found t o be

satisfactory.

There were n h e r o u s leaking tube f i t t i n g s i n t h e a i r l i n e s .

365

Severa1,quick disonnects on a i r l i n e s i n s i d e t h e r e a c t o r c e l l and draintank c e l l were found t o be leaking when checked wi$h leak-detector soluThese did not c o n s t i t u t e a leak i n secondary containment since each

tion,

l i n e has a block valve outside t h e c e l l ; but a leak here would a f f e c t t h e c e l l leak r a t e indication when a i r pressure w a s on t h e l i n e t o operate t h e . valve. The b u t t e r f l y valves i n t h e 30-inch l i n e used f o r v e n t i l a t i n g t h e c e l l s d w i n g maintenance operations were checked by pressurizing between them. The leakage measured by a flowmeter w a s excessive, and t h e valves had t o be removed from t h e system t o determine t h e cause. d i r t on t h e rubber s e a t s , and one was cut. repaired.

There was considerable

These seats were cleaned and

It was found t h a t t h e motor drive u n i t s would s l i p on t h e i r

mounting p l a t e s by a small amount, thus causing a slight e r r o r i n t h e indic a t e d , p o s i t i o n of t h e valve. t o prevent t h i s .

Dowels were i n s t a l l e d i n t h e mounting p l a t e s

Small leaks were a l s o found around t h e pins which f a s t e n

the b u t t e r f l y t o t h e operating .shaft.

These leaks were repaired w i t h epoxy

resin. The l i n e from t h e thermal-shield rupture d i s c s t o t h e vapor-condensing systems had numerous threaded j o i n t s t h a t leaked badly when pressurized with nitrogen.

Each j o i n t was broken, t h e threads were coated w i t h epoxy

r e s i n , and t h e j o i n t was remade.

All j o i n t s were l e a k t i g h t when rechecked

w i t h leak-detector solution.

The c e l l pressure w a s then increased i n increments and extensive soapchecking and leak-hunting and t h e c e l l s were press with leak-detector s o l u t

done. Alternate t o p blocks were i n s t a l l e d , o 1 p s i g t o leak-check all membrane welds

c e l l s at 1 psig, all pen

leaks were found i n t h e welds. With t h e 6 , pipe j o i n t s , tube f i t t i n g s , and mineral-

insulated ( M I ) e l e c t r i c a l c

s e a l s subjected t o t h i s pressure were

checked w i t h leak-detector s o f i t t i n g s and M I cable s e a l s .

Numerous leaks were found i n tube

of t h e leaks were stopped by simply

tightening t h e threaded p a r t s of t h e s e a l .

However, t h e leak r a t e w a s

s t i l l about 4500 ft3/day, indicating some major leak t h a t had not been found.

L.i

366 A l l s h i e l d blocks were i n s t a l l e d , t h e c e l l s were p r e s s u r i z e d t o

and leak-hunting continued.

Three l a r g e leaks were'located.

5 psig,

*iiJ

One w a s

through t h e sleeve t h a t surrounds t h e f u e l off-gas l i n e from t h e reactor c e l l t o a v e n t i l a t e d p i t i n t h e vent .house; t h e sleeve was 'supposed t o have beep welded t o t h e l i n e i n t h e r e a c t o r c e l l but t h i s had been overlooked. Therefore, t h e sleeve w a s closed at t h e vent house,.where it w a s accessible. Another leak was i n an instrument air l i n e t o an i n - c e l l valve (HCV-523), and t h e t h i r d w a s from t h e vapor-condensing system t o t h e drain-tank.steam domes and out t o t h e north e l e c t r i c service a r e a t h r p a g h a l i n e t h a t was temporarily open.

A l l penetrations, tube f i t t i n g s , and M I cable seals were

again checked with leak-detector solution.

Many MI. cable s e a l s which had

not leaked at 1 p s i g were found t o be leaking, and some of those which-had been tightened and sealed a t 1 p s i g now leaked.

A s a reskt, several of t h e gland nuts s p l i t and soldering

were tightened.

was required.

Again, many of t h e s e a l s

A l l l a r g e leaks were sealed o r g r e e t l y reduced, and many of

t h e small leaks were stopped. Leak-hunting and r e p a i r s were continued at 10, 20, and 30 psig.

The

MI-cable seals as a group accounted f o r a l a r g e percentage o f t h e remaining

leaks.

To s t o p small leeks which may not have been located, all MI-cable

seals were coated with epoxy r e s i n at t h e seals outside of t h e cells'. This was done with t h e c e l l pressure at -1.5 p s i g so t h a t t h e epoxy would be drawn i n t o t h e s e a l s . Teflon tape, used extensively on MI-cable seals, other threaded pipe, and tube f i t t i n g s , did not perform s a t i s f a c t o r i l y i n providing a gas s e a l . During t h e 30-psig t e s t , l a r g e leaks were located i n both componentcoolant-pump-dome audible.)

flanges using leak-detector solution.

(One of them w a s

The i n l e t and o u t l e t valves t o t h e s e domes were then closed and

the' domes were opened for work on t h e gaskets. rubber gaskets were found t o be undamaged.

The,broad, f l a t , *Viton-

The leakage w a s a t t r i b u t e d t o

inadequate loading pressure on the broad gaskets, and after they were narrowed, no f u r t h e r leaks were observed. Leak-rate data was then taken at 30, 20, 10, and -c! p s i g with t h e component-coolant-pump domes valved back i n t o the system,

These a r e plot-

t e d i n Fig, 22.2 along with curves which r e l a t e t h e allowable leakage at various pressures t o t h e allowable leakage a t '39 p s i g (MCA pressure) f o r

367

ORNL-DWG 66-4753

500

400

300

200 W 0

a Y a W

-I

IO0

-io0

-@ @

ORIFICE FLOW

TURBULENT FLOW

--@ 0

-200

ASSUMED LINEAR RELATIONSHIP

LAMINAR FLOW

EXPERIMENTAL DATA

368

various f l o w regimes.

Also shown i s t h e highly conservative l i n e a r re-

l a t i o n s h i p which has no physical b a s i s . The allowable leak rates based on t h e o r i f i c e flow curve ( t h e most conservative r e a l i s t i c curve) along with t h e measured leak r a t e s during these first tests i n 1965 are given i n Table 22.3.

Table 22.3

C e l l Leak Rates

Test Allowable Pressure Leak Rate b i g ) (scfd)

30* * 20

* The **

1965

360

130

290

125 40

195

Observed Leak Rate (scfd)

1966

1967

1968

1969

150

instrument air block valves were closed during these t e s t s .

The leak rates at p s i g are those calculated f o r t h e periods following t h e pressure tests. The instrument a i r block valves were open during these periods.

nuaJ checks o f t h e secondary containment consisted of: (1)t e s t i n g and r e p a i r i n g all block valves, ( 2 ) checliing t h e leak rate at

'some elevated pressure (usually 20 p s i g ) , and (3) checking t h e leak r a t e while operating with t h e c e l l s at -i! psig.

The number of valves found

leaking d u r i n g t h e s e tests a r e shown i n Table 22,2.

The measured leak r a t e s

a r e given i n Table 22.3. During operation of t h e r e a c t o r , t h e c e l l s were maintained at . .

+' p s i g

On t h r e e occasiops, and . t h. e leak rate was monitored somewhat continuously. the' r e a c t o r was shut down due t o indicated high c e l l leak r a t e s . Further

64 I

369

investigation showed t h a t leakage would not have been excessive during an accident on any of these occasions.

These are described below.

I n t h e l a t t e r p a r t of May 1966, a c e l l leak r a t e of 100 scf/day at 4 p s i g w a s calculated.

The r e a c t o r w a s shut down and subsequent investi-

gation, including a 20-psig pressure t e s t , disclosed a high leakage r a t e through t h e thermocouple sheaths f r o m t h e pressurized headers i n t o t h e c e l l s . Since no s i g n i f i c a n t leakage w a s â&#x20AC;&#x2DC;detected from tkie thermocouple headers outside t h e r e a c t o r c e l l , a flowmeter was i n s t a l l e d i n t h e nitrogen supply This flow w a s included i n t h e leak-rate calculations.

line.

Prior t o t h i s

tlme t h e headers had been maintained between 5 and 50 p s i g by periodic pressurization.

To decrease t h e purge t o a minimum, t h e procedures were changed

t o maintain t h e pressure at 5 psig.

A containment block valve w a s i n s t a l l e d

i n t h e supply l i n e . In November 1966, t h e indicated c a l l leak r a t e increased t o 300 s c f d at 4 2 psig.

Therefore, t h e r e a c t o r was drained on November 20.

Leaks were

found i n two air supply l i n e s and one vent l i n e used f o r in-cell air-operated valves.

One of t h e supply l i n e s was capped since it supplied a valve which

w a s only o p e r a t e d p e r i o d i c a l l y .

A rotameter w a s i n s t a l l e d i n t h e other

supply l i p e and t h e vent l i n e w a s connected t o t h e c e l l . A f t e r measuring a c e l l leak r a t e of 45 scf/day during a 10-psig pres-

sure t e s t , Run 10 w a s started.

Early i n thy run t h e new rotameter i n t h e

a i r l i n e indicated t h a t t h e i n - c e l l leak had increased.

Rotameters were

a l s o i n s t a l l e d on t h r e e more a i r l i n e s which were found t o be leaking i n The c e l l .

Later, t h e known leakage increased t o 3500 scfd.

Although t h e

c d c u l a t e d c e l l leak r a t e appeared t o remain a t about 50 scfd, Run 10 w a s terminated i n January 196

se t h e possible e r r o r i n t h e measured purge

rates i n the leaking i n s t

a i r l i n e s exceeded t h e permissible leak

rate

During t h e shutdown, t

i r - l i n e leaks were t r a c e

quick discon-

nects i n which neoprene se

ad become embrittled.

with elastomer s e a l s ; eight

connects, a l l near t h e center of t h e reactor

c e l l , were leaking.

of t h e s e were replaced w i t h s p e c i a l adaptors

Sevente

18 disconnects

sealed at one end by an alu$num gasket and a t t h e other by a standard

metal-tubing compression f i t t i n g , (One disconnect w a s not replaced because it w a s on a l i n e t h a t i s always a t c e l l pressure.) No similar

370

d i f f i c u l t i e s were encountered.

A f t e r s e a l i n g t h e c e l l , t h e leak rate

was 50 scfd. Discussion of C e l l Leak Rate Determinations

22.3

There were at least t h r e e ways of determining t h e c e l l leak rate, These were:

(1)a material balance plus compensation f o r pressure and tem-

perature using changes of t h e d i f f e r e n t i a l pressure between t h e c e l l atmos.

phere and an i n - c e l l reference volume; ( 2 ) a material balance.plus compens a t i o n f o r absolute temperature changes i n t h e c e l l s , and (3) an oxygen balance The first method w a s used f o r c a l c u l a t i n g a l l o f f i c i a l leak r a t e s at t h e MSRE.

A discussion of each method follows.

Method 1

-- Material Balance P l u s

Compensation Based on t h e D i f f e r e n t i a l Between t h e C e l l Pressure and t h e Reference Volume Pressure

I n order to,determine t h e c e l l leak r a t e i n a reasonable time, very accurate indication of c e l l pressure changes was necessary.

The instrument

used t o determine t h e pressure changes w a s e s s e n t i a l l y a "U" tube manometer (Fig. 22.3) c a l l e d a "hook gage," .

Pointed shafts attached t o micrometers

protruded through "0" r i n g s e a l s i n t h e bottom of t h e chamber.

Readings

taken by adjusting t h e point of t h e s h a f t u n t i l it w a s at t h e l i q u i d surface and then reading t h e micrometer.

Thus changes i n water l e v e l corre-

sponding t o c e l l j p r e s s u r e changes could be accurately measured. n a l gage had a range of 2 inches of .water.

The o r i g i -

A l a t e r model was obtained i n

which t h e reference chamber could be moved a measured amount which gave a range of 12 inches of water, The changes i n c e l l pressure were measured r e l a t i v e t o t h a t of an inc e l l reference volume.

The reference volume was d i s t r i b u t e d throughout t h e

reactor c e l l and drain-tank c e l l i n an attempt t o compensate f o r c e l l temperature changes.

It was found t h a t very small leaks i n t h e l i n e s t o t h e

reference volume could give l a r g e e r r o r s i n t h e leak-rate data.

Welding

these l i n e s eliminated t h i s d i f f i c u l t y . Approximately 55 of t h e contained volume was located outside of t h e reactor and drain-tank c e l l s and had no temperature compensation.

This

371

ORNL DWG. 67-2689

1 * , d

NOOK C A G E

Fig. 22.3

Schematic of Hook Gage and Piping for Determining Pressure Change of Contained V Relative to the Reference Volume

372

included t h e 30-inch-diameter

reactor c e l l v e n t i l a t i o n duct ( l i n e 930) i n

t h e coolant drain tank c e l l and t h e component-coolant-pump domes i n t h e special equipment room.

These areas were sealed o f f as well as possible

t o minimize temperature changes.

A t power t h e r e w a s s t i l l a considerable

amount of a i r leakage past these from t h e main blowers which caused daynight temperature variations up t o 20째F during winter months.

The e f f e c t

of these temperature fluctuations was magnified by t h e presence of w a t e r vapor i n t h e c e l l air.

(Since June 1967, t h e r e was a s m a l l continuous

water leak (<1 gal/day) i n t o t h e reactor c e l l . )

This water evaporated i n

t h e c e l l s and condensed i n t h e cooler sections of t h e containment ( l i n e 930 and t h e component-coolant-pump gas cooler) e from t h e system.

This was periodically drained

I n colder weather more water would condense.,

Reactor power a l s o had an e f f e c t on t h e indicated c e l l leak r a t e due t o c e l l temperature changes.

When t h e reactor w a s taken f'rom zero t o f u l l

power, gamma heating i n the thermal s h i e l d and other equipment i n t h e c e l l caused t h e average c e l l temperature t o increase 4 t o 8OF and indicated a high c e l l leak r a t e f o r approximately 24 hours.

These changes were not

adequately compensated f o r by t h e pressure reference volume.

Af'ter t h e c e l l s were closed following each i n - c e l l maintenance period,

t h e containment w a s purged with dry B2 t o remove oxygen, s o - t h e c e l l gas s t a r t e d out dry.

A s t h e water evaporated, t h e i n i t i a l c e l l leak r a t e was

usually high (75-130

scf/day has been calculated) b u t gradually decreased

t o an equilibrium value i n 5 t o 7 days at which time condensate s t a r t e d t o form and was drained from t h e system.

Duringsthis i n i t i a l period u n t i l -

condensate apped-ed, we relied on results of leak-testing a t p o s i t i v e pres-

sure before t h e C e l l w a s evacuated and purged.

Due t o t h e s c a t t e r i n t h e

data and t h e r e l a t i v e l y small change i n pressure o r temperature t h a t represents a l a r g e leak r a t e ,

considerable time was required t o obtain data

from which r e l i a b l e leak r a t e could be determined.

To minimize t h e t e m -

perature and other e f f e c t s , i n i t i a l and f i n a l data w a s usually taken at i

change i n c e l l pressure of 0.3 inches of water (0.01 p s i ) per day o r a change i n the average temperature of O b 0 k O ~w i l l change t h e indicated leak rate by about 10 scfd.

373

t h e same time of t h e day and a t t h e same operating conditions.

Calcula-

t i o n s of i n t e r v a l s of several days were more consistent than s h o r t e r periods.

Figure 22.4 i s a p l o t of t h e data used t o determine t h e leakage

r a t e at 30 p s i g i n 1965.

Typical flow r a t e s f o r t h e material balances were:

9 scfd t o t h e sump

bubblers, 13 scfd purge f o r t h e thermocouple header, and 0 t o 1000 scfd evacuation flow.

Since t h e evacuation flow was normally t h e l a r g e s t and

was changed more often, i t s s i g n a l was sent t o t h e computer which i n t e grated it and each s h i f t typed out a value f o r t h e t o t a l volume evacuated. Method 2

-- Using t h e Absolute

C e l l Pressures and Temperatures

This method w a s not used because t h e absolute c e l l pressure indication was not as accurate as t h e indication of d i f f e r e n t i a l pressure by t h e hook gage and compensating f o r temperature changes using absolute temperatures

did not give as consistent r e s u l t s as those obtained using t h e reference volume. Method 3

-- Using an Oxygen Balance

The MSRF: containment was purged with N2 t o keep t h e 02 concentration

<5% primarily t o eliminate t h e danger of an explosion i f an o i l leak should develop i n t h e fuel-pump l u b r i c a t i n g system.

This made keeping an oxygen

balance on t h e c e l l appear as an a t t r a c t i v e method of measuring t h e c e l l leak rate when t h e c e l l was at a negative pressure.

The precision t o which w e 'could read t h e 02 analyzer w a s only 50.1% which i s equivalent t o $60 f't3 of a i r i n t h e c e l l .

With a leak r a t e nor-

mally $20 f't3/day, it became apparent t h a t , t h e only accurate calculation . would be over very long time periods.

I n comparison t o Method 1 ( c e l l leak

r a t e using t h e c e l l pressure change and a flow balance), t h e oxygen balance leak rate was consistently low by $15 scf/day.

I n f a c t , when t h e leak rate

calculated by Method 1 w a s 4 5 scf/day, t h e 02 data indicated a negative leakage.

*Ten

LiJ

ambient thermocouples were located i n t h e r e a c t o r c e l l , 6 i n t h e drain-tank c e l l , and 1 i n t h e s p e c i a l equipment room.

374

422.2

QI.0 89.2

Fig. 22.4

Containment-Conditions During Leak Rate Test at 30 psig

375

Errors involved i n t h e pressure method of calculating t h e leak r a t e would probably be associated with t h e rotameters; but t h i s would imply a 25% e r r o r i n rotameter c a l i b r a t i o n .

All of t h e rotameters involved were

disconnected and bench-calibrated at l e a s t twice.

They were found t o be

accurate t o w e l l within 5%. The containment c e l l was nominally at

+ psig,

but j u s t downstream of

t h e component-cooling pumps, t h e pressure w a s +6 psig.

It was thought t h a t

i f gas were leaking out here and i n t o t h e c e l l at a proportionately higher

rate, we might account f o r t h e anomaly.

Simultaneous solution of t h e leak-

rate equations f o r t h i s s i t u a t i o n showed t h a t although a solution w a s mathematically possible, t h e in-leak would have t o be a t %3%02 and t h e out-leak a t ' 2 0 % 02 which i s not p r a c t i c a l . These r e s u l t s caused us t o t h i n k t h a t something i n t h e c e l l w a s chemic a l l y consuming some of t h e oxygen.

Removal from t h e c e l l air of C2.5

scf/day of pure 02 would account f o r t h e leak-rate discrepancy. One suggested mechanism f o r chemically removing t h e 02 w a s by o i l decomposition.

If some of t h e o i l from t h e component-cooling pumps were con-

t a c t i n g hot pipe (say at FV-103 o r i n control rod thimbles) it would probably at l e a s t p a r t i a l l y decompose according t o t h e approximate r e l a t i o n :

Cell a i r samples were taken t o determine t h e C02 content and they showed

QO.l%

C02.

This was at l e a s t t h r e e times higher than t h e C02 content i n

air, but w a s less than one would expect t o s e e i f all t h e 02 were being consumed by o i l decomposition. The c o n t a i w e n t c e l l and support s t r u c t u r e w a s composed primarily of carbon s t e e l so another prime suspected oxygen depletion mechanism w a s r u s t . If all t h e missing 02 were being consumed t o form FepOg, it would t a k e

about 120 g/day of Fe.

Since t h e r e i s abou;t 5000 f t 2 of carbon s t e e l sur-

face a r e a i n t h e c e l l , t h i s would correspond t o a corrosion r a t e of only 0.5 mils/year (neglecting t h e support s t r u c t u r e and piping of component

coolant system).

;Li

376

This analysis l e d t o t h e conclusion t h a t enough oxygen w a s being re-

moved chemically by o i l decomposition and r u s t i n g t o produce t h e discrepancy t h a t existed between t h e d i f f e r e n t methods of c e l l leak-rate calculation. 22.4

Vapor Condensing System

An accident can be conceived i n which molten s a l t and water could

simultaneously leak i n t o t h e r e a c t o r o r drain-tank c e l l s .

The quantity of

steam produced could be such t h a t t h e pressure i n t h e c e l l s w b u l d increase above design pressure.

A vapor-condensing system w a s provided t o prevent

the steam pressure from r i s i n g above t h e 39-psig design pressure and t o r e t a i n t h e non-condensable gases. vapor-condensing system.

A 12-in. l i n e connected t h e c e l l s

the

This l i n e contained two rupture d i s c s i n p a r a l l e l

(a S i n . d i s c with a bursting pressure of 15 p s i g and a 10-in. ilisc with a bursting pressure of 20 psig). The l i n e from t h e rupture d i s c s went t o t h e bottom of a 1800-ft3 v e r t i c a l tank which contained %1200 f t 3 of water t o condense t h e steam.

The non-condensable gases went from t h e t o p of t h i s

tank t o a 3900-ft3'retention tank. The tanks were pressure-tested by t h e vendor t o 45 psig.

A f t e r being

connected t o t h e reactor c e l l , t h e system w a s leak-tested a t 30 p s i g during i n i t i a l t e s t i n g of the reactor and drain-tank c e l l s i n 1965. During the annual c e l l leak t e s t s at pressure, t h e vapor-condensing system was a l s o pressurized.

This protected t h e rupture d i s c s and provided

an i n t e g r i t y t e s t of t h e vapor-condensing system. The water used i n t h e v e r t i c a l tank contained potassium n i t r i t e potassium borate as a corrosion i n h i b i t o r . i n h i b i t o r addition was made.

Based on annual s&les,

one

There was no increase i n t h e . i r o n concentra-

t i o n which indicated l i t t l e o r no corrosion, I n i t i a l l y 4 l e v e l probes were provided t o assure proper w a t e r l e v e l . However, a f t e r one of t h e l o w level probes failed, a bubbler-type l e v e l instrument w a s i n s t a l l e d f o r measuring t h e l e v e l periodically, Since t h e r e has been no unplanned increases i n t h e pressure of t h e reactor and drain-tank c e l l s , t h e vapor-condensing system has not been used.

377

22.5

Recommendations

Leak-checking t h e numerous valves required disconnecting many l i n e s , p a r t i c u l a r l y tube f i t t i n g s and autoclave connections.

Disconnecting nu-

merous f i t t i n g s g r e a t l y increases t h e probability that one o r more w i l l leak when connections are remade.

For t h i s reason, an e f f o r t should be

made i n t h e design of t h e piping t o minimize t h e number of disconnects

necessary f o r leak-checking.

For example, nitrogen l i n e s could be con-

nected as shown i n Fig. 22.5 t o f a c i l i t a t e checking both t h e s a f e t y valve and the l i n e inside t h e c e l l .

All block valves should be of high q u a l i t y and soft-seated.

Accessi-

b i l i t y t o t h e items t o be checked and t h e points used i n checking them should be considered.

F l e x i b i l i t y should be provided i n t h e l i n e s which

must be disconnected. Leaks from i n - c e l l a i r l i n e s cause e r r o r s i n leak-rate calculations. When disconnects are necessary, t h e e f f e c t of f l u x on materials on con-

s t r u c t i o n should be considered. MI-cable s e a l s of t h e type used at t h e MSRE (brass and s t a i n l e s s s t e e l ) should be modified o r another type used.

Substituting f e r r u l e s of

t e f l o n , graphite-impregnated asbestos, or other r e l a t i v e l y soft material f o r the brass f e r r u l e s may be s u f f i c i e n t . t o q r e v e n t leakage through them. However, r a d i a t i o n damage and sheath temperature must be considered.

Stan-

dard pipe-threaded connections f o r s t a i n l e s s steel. t o s t a i n l e s s s t e e l

j o i n t s should be avoided where possible. ed on threaded connections f o r a gas seal

Teflon tape should

nor on small l i n e s where fâ&#x20AC;&#x2122;ragments-can e n t e r t h e l i n e and clog it o r prei

vent check valves from s e a t i n g properly. should-be cleaned i n t e r n

All pipe, tubing, and valves .

r i o r t o installation.

MI-sheathed thermoc

are preferable t o t h e type used.

if cost o r other reasons

However,

seals at t h e ends) preclude using them, t h e

i n i t i a l design should provide f o r pressurizing t h e terminal headers and measuring t h e leakage.

Sol

ed j o i n t s i n t h e header system a r e recom-

mended where practical; Gas a n d y z e r s used t o continuously monitor t h e c e l l atmosphere should

be located i n a w a r m area t o prevent moisture from condensing i n them.

378

379

Fluids used i n manometers o r s i m i l a r gages which a r e continuously connected t o t h e contained volume such as t h e hook gage should have a low vapor pressure t o avoid a s i g n i f i c a n t l o s s during operation.

The piping

t o t h e gage should be pitched t o prevent accumulation of water and should not be connected t o t h e bottom of temperature compensating o r reference volumes. Temperature compensating volumes should represent t h e e n t i r e containment.

The 30-in.

c e l l vent

i n e and t h e component-cooling system enclosure

(5% of c e l l volume) had no reference volume.

w e temperature reference

volume should be d i s t r i b u t e d throughout t h e c e l l s .

The MSRE used 6-in.

v e r t i c a l pipes on a weighted volume b a s i s i n t h e r e a c t o r and drain-tank cells.

Smaller diameter piping with t h e same t o t a l volume might have given

a more repqesentative temperature compensation.

The reference volume system e x t e r n a l t o t h e c e l l s must have a minimum volume and should be located i n an area h e l d at as constant a temperature as possible.

S u f f i c i e n t c e l l temperature thermocouples should be located

i n all sections of t h e containment t o enable calculation of average temperature and temperature changes as accurately as possible. If a s e n s i t i v e c e l l pressure measuring device similar t o t h e hook gage

i s used i n future reactor containment, it should be designed t o meet secondary containment requirements so t h a t block valves a r e not required. However, i f block valves are needed t o i s o l a t e t h e instrument, a separate c i r c u i t from t h e other block valves should be used.

Pressure readings a r e

required above t h e pressures a t which t h e other block valves a r e closed. Safety jumpers could be used t o keep t h i s c i r c u i t open during cell-pressure testing.

23.

BIOLOGICAL SHIEISING AND RADIATION LEVELS

T. L. Hudson The c r i t e r i o n f o r t h e MSRE b i o l o g i c a l s h i e l d design w a s t h a t t h e dose

rate would not exceed 2.5 mrem/hr during normal operation at any point on t h i e s h i e l d e x t e r i o r t h a t i s located i n an unlimited access area, . Since the MSRE had t o f i t within an e x i s t i n g r e a c t o r containment c e l l and build-

ing, t h e s h i e l d design allowed f o r addition of shielding as needed t o reduce radiation .level at l o c a l i z e d hot spots o

The shielding calculations

are reported i n Reference 57. The reactor vessel was completely surrounded by a water-cooled,

and-water-filled

thermal shield.

The thermal s h i e l d and f u e l c i r c u l a t i n g

1oop.were located i n t h e reactor c e l l . layers of' concrete blocks.

steel-

The t o p of t h e r e a c t o r c e l l had two

An annular space f i l l e d w i t h magnetite sand and

w a t e r provided t h e shielding f o r t h e sides and bottom.

See Fig. 23.1.

When t h e reactor was not i n operation, t h e f u e l was drained t o one o r both drain tanks which were located i n the drain-tank c e l l .

Magnetite con-

c r e t e w a l l s faced a l l accessible areas and t h e t o p consisted of two layers of concrete blocks. 23.1

Radiation Surveys -Approach

t o Power

Extensive health-physics surveys o f t h e r e a c t o r a r e a w e r e performed during t h e i n i t i a l approach t o power. l i g h t s h i e l d inadequacies,

These surveys were made t o b r i n g t o

With t h e exception of t h e areas discussed i n

t h e following paragraphs, t h e shielding w a s found t o be adequate. 23.1.1

Coolant Drain Tank C e l l

A t 1 kW t h e s c a t t e r i n g of fast neutron and gamma rays from t h e reactor c e l l i n t o t h e coolant drain tank c e l l through t h e 30-in.

reactor c e l l ven-

t i l a t i o n l i n e 930 caused radiation readings of 8 mrem/hr gamma and 50 mrem/hr fast neutron near t h e e x i t of the l i n e .

A t 25 kW very high readings

were found again a t l i n e 930 (70 mrem/hr gamma, 600 mrem/hr fast neutrons and 30 mrem/hr thermal neutrons).

6 in. of borated polyethylene was coolant drain tank c e l l .

A w a l l of

16 i n . of concrete blocks and

b u i l d ad3acent t o t h e 930 l i n e i n t h e

I n addition, t h e reactor off-gas l i n e 522, i n t h e

LilJ

y S f l 1fl NNQ

"XX,

N01133S

11QM 1133-

snoef

NMOa-OlOH

NQld

T8€

382

coolant drain tank c e l l was found t o be giving a high background t o t h e A t t h e end of t h e 25-kW run, t h e reading w a s 100 mR/hr at 1 i n . from

area.

t h e S i n . t h i c k lead s h i e l d around t h e 522 l i n e .

Later a t higher power

levels it w a s determined t h a t part of t h e r a d i a t i o n was from t h e drain-tank vent l i n e 561. Therefore t h i s l i n e was shielded with 3 i n . of lead. Even though shielding w a s added i n s i d e t h e coolant drain tank c e l l , t h e r a d i a t i o n

at t h e door (500 mR/hr gamma, 150 mrem/hr fast neutrons, 75 mrem/hr thermal neutron) and halfway up t h e access ramp (22 mR/hr gamma, 3 mrem/hr fast neutrons, 32 mrem/hr thermal neutron (was high during full-power operation. This- area was c l e a r l y marked with r a d i a t i o n zone signs a t t h e entrance t o t h e ramp and t h e door at t h e bottom of t h e ramp w a s locked during nuclear operation t o prevent entry i n t o t h e c e l l . 23.1-2

North E l e c t r i c Service Area

When t h e reactor power w a s r a i s e d t o 1 MW i n A p r i l 1966, t h e r a d i a t i o n l e v e l i n t h e Morth E l e c t r i c Service Area (NESA) w a s .found t o be high: mR/hr on t h e balcony and 8000 mR/hr a t t h e west w a h .

Investigation showed

t h a t t h e r e w a s radioactive gas i n t h e l i n e s through which helium i s added t o t h e drain tanks.

Two check valves i n each l i n e prevented t h e

LlJ

gas from

g e t t i n g beyond t h e secondary containment enclosure, but t h e enclosure, of 1/2-in.

s t e e l , provided l i t t l e gama shielding.

The pressure i n t h e f u e l

system at t h a t ‘time w a s controlled by t h e newly i n s t a l l e d pressure control .

valve with r a t h e r coarse trim, and t h e pressure fluctuated around t h e cont r o l point (normally 5 psig) by k %2%. These pressure f l u c t u a t i o n s caused f i s s i o n product gases t o diffuse more rapidly i n t o t h e drain tanks and back through t h e 1/4-in.

l i n e s through t h e s h i e l d i n t o t h e MESA.

The r a d i a t i o n

level was lessened by i n s t a l l i n g a temporary means of supplying an i n t e r -

mittent purge t o t h e gas-addition l i n e s t o sweep t h e f i s s i o n product gases back i n t o t h e drain tanks.

During t h e June shutdown, a permanent purge

system was i n s t a l l e d t o supply a continuous helium purge of 70 cc/min t o each of t h e t h r e e gas-addition l i n e s .

This was proved successful by sub-

sequent full-power operation i n which t h e general background i n t h e MESA

w a s <1mR/hr. Transmitter Room

- w i n g t h e r a d i a t i o n survey a t

5 MW a hot spot of

”‘

30 mR/hr gamma w a s found on t h e southwest corner about 4 f t above t h e floor. _-

383

bd A 2-ft by

4-ft by 1-in. t h i c k l e a d sheet was attached t o t h e w a l l over t h e

hot spot and a radiation zone was established. t h e i n i t i a l approach t o full power, stacked con-

Vent House -During

c r e t e blocks were added t o t h e f l o o r area of t h e vent house, over t h e charcoal beds and between t h e vent house and t h e r e a c t o r building t o keep dose

rates low. bricks.

Very narrow beams coming from cracks were shielded with l e a d

Even so, t h e background radiation l e v e l i n t h e vent house w a s

%7 mR/hr at f'ull power.

The vent house was established as a radiation

zone area.,

Water Room

- Induced

a c t i v i t y i n t h e t r e a t e d water rose t o an unex-

pectedly high level during Run

The a c t i v i t y proved t o be 12.4 h r 42K

produced i n t h e corrosion i n h i b i t o r .

A survey of possible replacements

f o r potassium l e d t o t h e choice of lithium, highly enriched i n t h e 'Li tope t o minimize tritium production.

iso-

See Water System (Section 12) of t h i s

report f o r additional d e t a i l s Top of Reactor C e l l

- It was expected t h a t additional shielding would

be required d i r e c t l y ab0 between t h e shield block

he reactor, where t h e r e a r e cracks (Q/2 i n . ) w i n g a full-power run i n J u l y 1966, two very

narrow beams of y with neutron radiation were found between c e l l blocks f o r t h e first time. tron.

These read up t o 10 mR/hr gamma and 60 mrem/hr fast neu-

They were properly marked. 23.2

Radiation Levels During Operation

, the

During operation at

narrow beams were noted from time t o

t i o n a l areas were acceptabl time.

r a d i a t i o n l e v e l s i n a l l opera-

These were mainly

house.

Due t o induced a c t i v i t y i n t h e

t r e a t e d water, areas ne

pment such as t h e surge tank and heat

exchanger were t r e a t e d as

zones.

Periodic r a d i a t i o n surveys have

not disclosed any appreciable changes i n r a d i a t i o n l e v e l s . Some t y p i c a l r a d i a t Table 23.1.

Changes i n

s i n s i d e t h e shielding a r e given i n evels following a shutdown from power are

shown i n Figs. 23.2, 23.3, and 23.4.

384

Table 23.1 Radiation Dose Rates i n Various Areas .During and Following Full-Power Operation

Gamma Dose Rate (R/hr) At

LOCATION

Reactor Cell

Drain Tank C e l l Coolant C e l l

d Fuel Sampler-Enricher d Off-gas Sampler

7 MW

x 104 4.2 x 103

b React o r Drained

10 k@

5.4 x

103

4.2 x 103

2.4

103

2.6 x

104

React o r Drained andc Flushed

103

100

1000 500

‘After 5-hOUrS operation a t 10 kW which followed sustained operation at f’ull power. bImmediately following t h e ‘above 5-hr operation a t 10 kW. C

Two days af’ter t h e above f u e l drain.

%nside t h e sampler shielding during sampling.

385

ORNL-DWG 73-608

0 EAST WALL REACTOR CELL - RM 6000-1

DAYS DECAY

ion Level in Reactor Cell

386 ORNL-DWG 73-609

d 0

1 2

3 4

5 6

7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 DAYS DECAY

Fig. 23.3

Gamma Radiation Level i n Drain Tank C e l l (Rm 6000-6 between FD-1 and FD-2)

387 ORNL-DWG 73-610

Id 5

101

lo0 5

DAYS DECAY

Fig. 23.4

Gamma Radiation Level i n Coolant C e l l (Rm-6010)

388

23.3 Conclusions The b i o l o g i c a l shielding was adequate as designed except for the few , I

l o c a l i z e d areas discussed previously.

Periodic radiation surveys have not

indicated any shif't or deterioration i n any o f t h e shielding.

389

24.

INSTRUMENTATION R. H. Guymon

24.1

Introduction

It i s not wi,--in t h e scope of t h i s report t o cover i n d e t a i l t h e performances of all instrumentation,--An

attempt has been made t o review it

from an operational viewpoint and'report s i g n i f i c a n t items. on instrumentation which

en given i n previous sections on t h e perforndividual systems and components w i l l not be

mance o f t h e e n t i r e plant repeated here.

Information

The per

of t h e on-site computer i s covered i n

Reference 58. 24.2

Description

Most of t h e MSRE instrumentation w a s of conventional type found i n Emphasis w a s placed on assuring ade-

other reactors o r chemical plants.

quate containment and reducing r a d i a t i o n damage. signed t o f a i l safe. 24.2.1

A l l c i r c u i t r y w a s de-

A very b r i e f description follows.

Nuclear Instruments

a l l of t h e nuclear instruments were located i n a 36-in.-dia.

water-filled thimble which extended from t h e high bay

through t h e r e a c t o r c e l l t o t h e v i c i n i t y of t h e reactor Fig. 24.1.) Three uncompensated i o n

d t h e i r associated f a s t - t r i p com-

parators provided s a f e t y i (High r e a c t o r o u t l e t tempe i n a two-out-of-three

f o r scramming t h e c o n t r o l rods.

so caused a rod scram.)

configuration. ovided.

These chambers

and t h e e f f e c t of t t o r e d i n t o a s i g n a l whic recorder.

They wer

.'control interlocks.

These were used d automatic PO-

r position were

orded on a 10-decade logarithmic power rod i n h i b i t , rod reverse, and other

390

Q(NL-UUG 644-625

OUTER SLEEVE ASSEMBLY (WITH EXMNSION JOINT)

1\48-in00

-EXf! JOINT

Fig. 24.1

&-CONTAINMENT

PENETRATION EL. awn

in.

Elevation View of Nuclear Instrument Penetration

6.’

391

Two compensated ion chambers were used f o r servo control of t h e regul a t i n g rod and f o r some interlocks. The power w a s recorded on a l i n e a r recorder.

Seven decades we

covered by means of manual range s e l e c t o r

switches. One high s e n s i t i v i t y BF3 chamber was provided f o r power indication during f i l l i n g of t h e reactor. 24.2.2

Process Radiation Instruments

Two kinds of detectors, ion chambers and Geiger Mueller tubes, were used t o i n d i c a t e t h e l e v e l of a c t i v i t y i n various process streams and t o provide necessary interlocks. 24.2.3

Health physics Monitoring

Gamma r a d i a t i o n w a s monitored by seven monitrons located throughout

t h e building.

The a i r contamination w a s monitored f o r beta-gamma emitting

p a r t i c l e s by seven constant a i r monitors.

The building evacuation system

operated when two o r more monitrons o r two o r more constant a i r monitors from a s p e c i f i c group of instruments detected a high l e v e l of r a d i a t i o n o r

air contamination.

24.2.4

Stack Activity Monitoring

The containment s t a c k a i r w a s checked f o r beta gamma p a r t i c u l a t e s by passing a s i d e stream through a f i l t e r paper and monitoring t h e a c t i v i t y with a GM tube.

A f t e r passing through t h e f i l t e r paper, t h e gaseous sample

passed through a charcoal t r a p which w a s monitored by another GM tube t o detect iodine.

A second side stream passed through another f i l t e r paper.

This was monitored f o r al

a t e s using a thallium-actuated zinc

sulfide screen detector,

1-alumel thermocouples (some of which were i n s t a l l e d spares) wer t h e salt thermocouples exc

A l l of even were located on t h e outside of t h e l i n e s f o r temperature indication.

The seven

ermocouple wells were located as follows:

one i n t h e r e a c t o r neck;

d two spares i n t h e r a d i a t o r i n l e t l i n e ;

and, one and two spare

radiator outlet line.

o r equipment.

The more important

o r multipoint recorders o r indicators.

Readout of others w a s accomplished by means of a scanner system which allowed t h e signals from up t o 100 thermocouples t o be s e n t t o a r o t a t i n g

392

mercury switch and subsequently t o an oscilloscope f o r display.

A switch

allowed s e l e c t i o n of any one of f i v e groups of 100 signals. 24.2.6

Pressure Indicators

Most of t h e remote indicating pressure and dp instruments were pneumatic o r e l e c t r i c force elements.

When containment w a s required, t h e vents

were referenced t o atmospheric pressure through r o l l i n g diaphram seals. S t r a i n gage and Bourdon pressure gages were used extensively.

Small changes

i n reactor c e l l pressure were determined using a "hook gage".

This was es-

s e n t i a l l y a water manometer with a micrometer f o r accurately reading changes i n water l e v e l . 24.2.7

Level and Weight Indicators

Bubbler type l e v e l instruhents were .used f o r measuring t h e s a l t l e v e l s i n the f u e l pump, coolant pump, and overflow tank. strument was a l s o i n s t a l l e d i n t h e coolant pump.

A float-type l e v e l in-

The drain tanks were sus-

pended by pneumatic weigh c e l l s t o determine t h e amount of salt t h a t they contained.

I n addition t o t h i s , two resistant-type single-point l e v e l

probes were provided i n each drain tapk.. Sight glasses, f l o a t s , bubblers, dp c e l l s , e t c . , were used i n t h e a u x i l i a r y systems. 24.2 8 Flow MeasurementNo flow instrument wm-provided i n t h e f u e l s a l t loop, however a ven-

turi meter w a s i n s t a l l e d i n t h e coolant salt loop. This w a s a standard venturi with a NaK-filled dp c e l l . Flows i n a u x i l i a r y systems were measured by o r i f i c e s , c a p i l l a r i e s p i t o t tubes, rotameters and matrix type flow elements 24.2.9 .

Miscellaneous

A semi-continuous mass spectrometer was used t o monitor t h e coolant

air stack f o r beryllium.

Other instrumentation included t h e following:

pulse-type speed elements, ammeters, voltmeters , and wattmeters f o r t h e salt pump motors; potentiometer and synchro position i n d i c a t o r s f o r t h e

control rods, f i s s i o n chambers, and r a d i a t o r doors; o q g e n and moisture analyzers f o r t h e helium cover-gas system; and an oxygen analyzer f o r t h e c e l l atmosphere.

393

24.3

I n i t i a l Checkout and Startup Tests

A comprehensive functional checkout of t h e control, s a f e t y , and alarm instrument w a s made by Instrument and Controls.personne1 p r i o r t o operation of each system.

Although some design and wiring e r r o r s were found, these

were of minor nature and were e a s i l y corrected. i n s t a l l a t i o n w a s excellent.

I n general, t h e q u a l i t y of

The following were included i n t h i s checkout:

t h e s e t p o i n t s of all switches were adjusted t o t h e proper values; continu i t y and r e s i s t a n c e checks of all thermocouples were made; t h e location of each thermocouple w a s determined by heating t h e thermocouple and measuring t h e voltage at t h e patch panel; and t h e continuity of all c i r c u i t s w a s checked and all recorders were put i n t o operation.

Most of t h i s w a s done

as construction was completed o r as t h e instruments were put i n t o service. P r i o r t o t h e first c i r c u l a t i o n of flush and coolant salts, a complete operational check was made of a l l ' instruments Bcheduled t o be used. This w a s done following t h e instrument s t a r t u p check l i s t . 2 2 This involved: (1)a complete checkout of a l l c i r c u i t s .

6.r

This was done by changing t h e

variable o r i n s e r t i n g a f a l s e s i g n a l at t h e prilsary element and assuring t h a t a l l c i r c u i t r y functioned at t h e proper setpoints; ( 2 ) a complete check of t h e patch panel t o assur

at all thermocouples - were . connected t o t h e at t h e operational records were up t o date;

proper readout instrument (3) a complete test of a1.1

equipment t o assure t h a t it would start

if needed and would ftmctio

t h a t no switches were i n a c t

r l y ; (4) a complete inspection t o assure o r jumpers i n s t a l l e d ; and ( 5 ) t e s t s t o

assure t h a t c r i t i c a l equipmen

ctioned properly, such as rod drop times.

This instrument start year.

Early t e s t s revealed

and o t h e r instrumen k

cated s e t p o i n t s which

l i s t w a s repeated approximately every r r o r s i n wiring, s e t p o i n t s which had drif'ted, t i o n properly.

Later tests indi-

and occasionally a malfunctioning

instrument. Most of t h e t r o u b l s t a r t u p check l i s t s adequate f o r s t a t i s

covered by doing t h e instrumentation iately.

Therefore records a r e in-

394

24.4

Periodic Testin&

Due t o t h e importance of some instruments o r c i r c u i t s , they were a l s o tested periodically during operation.

A s indicated by t h e sections which

follow, these tests did not reveal many serious troubles w h i c h needed correction.

They did provide assurance t h a t t h e c i r c u i t s should function i f

needed,

I n determining t h e amount and frequency of t e s t i n g , c a r e f u l con-

sideration should be given t o t h e above weighed against t h e harm done t o equipment due t o r e p e t i t i v e t e s t i n g and t h e p o s s i b i l i t y of interruption of operation.

Rod scrams, load scrams, and r e a c t o r drains occurred at t h e

MSRE due primarily t o t e s t i n g of equipment o r instrumentation. The periodic t e s t s made at t h e MSRE and t h e r e s u l t s of these t e s t s a r e given below. â&#x20AC;&#x2122;

24.4.1

Nuclear Instruments

A complete check of a l l nuclear instruments w a s made each month during

nuclear operation.

These were done by Instrument and Controls personnel

using Section 8A of t h e Operating Procedures.22

All t e s t s were s a t i s f a c -

t o r y except those l i s t e d i n Table 24.1. 24 4 2

Process Radiation Monitors

The process r a d i a t i o n monitors were tested each week during operation -by i n s e r t i n g a source near t h e primary detector and noting that each cont r o l i n t e r l o c k functioned properly and t h a t all annunciations occurred. A l l tests were s a t i s f a c t o r y except those l i s t e d i n Table 24.2,

Some e a r l y

d i f f i c u l t y was encountered i n t h a t t h e hole i n l e a d shielding f o r i n s e r t i n g t h e source was not properly placed. 24.4 e 3 Personnel Radiation and Stack Activity Monitorinq Periodic t e s t s of t h e h e a l t h physics monitors and stack a c t i v i t y instruments were o r i g i n a l l y done by MSRE personnel per Operating Procedure

8C,22 This was l a t e r taken over by o t h e r ORNL groups.

The o r i g i n a l sche-

dule f o r t h e h e a l t h physics monitors w a s t o m a k e a source check of each

instrument weekly, a matrix check monthly, and a complete evacuation t e s t semiannually.

As confidence i n t h e instruments w a s established, t h e fre-

quency of these tests was reduced t o monthly, quarterly, and semiannually.

395

Table 24.1

Date

3/22/66

Results of Periodic Tests of Nuclear Instruments

Repairs Necessary Replaced condenser i n t h e power supply of l i n e a r power channel No. 1. Replaced period balance module of nuclear safety channel No. 1.

6/16/67

Changed out high voltage supply of nuclear s a f e t y channel No. 3.

7/14/67

Repaired s c a l e r of wide-range counting channel No. 1.

8/12/68

Replaced relay of nuclear s a f e t y channel No. 3.

1/7/69

Replaced fast t r i p comparator of wide-range counting channel No. 2.

2/3/69

Replaced chamber of wide-range counting channel No. 1.

9/11/69

Replaced s c a l e r of wide-range counting channel No. 2.

10/21/69

Replaced operational amplifier of wide-range counting channel No. 1.

10/20/69

Replaced s c a l e r of wide-range counting channel No. 1.

396

Table ,24.2 Results of Periodic Tests ' O f Radiation Monitors

Troubles Detected

Date

4/13/66 5/10/66 1/9/68 1/26/68 7/21/68

3/3/69 10/10/69

arm

did

not occur, RM-596.

Instrument would not calibrate, RM-557. Two i n d i c a t o r lights burned out, RM-565. Indicator l i g h t burned out, RM-675. Indicator l i g h t burned out, RM-827. Indicator l i g h t burned out, RM-565. Indicator l i g h t burned out, RM-565.

397

cli

The stack monitors were o r i g i n a l l y t e s t e d with a source each week. Due t o t h e i r excellent performance record, t h i s w a s l a t e r reduce monthly tests. Troubles encountered were repaired immediately.

24.4.4

Safety C i r c u i t s

Periodic checks were made of all c i r c u i t s and instruments designated by t h e Instrument and Controls group as being safety.

These ipcluded rod

scram c i r c u i t s , f u e l pump and overflow tank pressures and l e v e l s , helium supply pressures, emergency f u e l drain c i r c u i t s , r e a c t o r c e l l pressures, coolant pump speeds and flows, r a d i a t o r temperatures and sampler-enricher interlocks.

These tests involved simulating a f a i l u r e apd checking as much

of t h e c i r c u i t r y as possible without i n t e r r u p t i n g operations.

A t first

these were performed weekly, then a l l except t h e rod scram checks were changed t o a monthly b a s i s .

A l l instruments functioned s a t i s f a c t o r i l y ex-

cept those indicated i n Table 24.3.

24.5

.Performance' of

t h e Nuclear Safety Instrumentation

These instruments proved t o be very r e l i a b l e . an abnormal number of spurious t r i p s occurred.

There were periods when

Many of these were believed

t o have origfnated i n f a u l t y , vibration-sensitive reltays- fn commercial elec~

t r o n i c switches which provided t h e high temperature t r i p . signals.

Another

possible source w a s the chattering of t h e r e l a y s which-change t h e sensit i v i t y of t h e f l u x amplifi c h a t t e r i n g and elimination tem reduced t h e fiequen The fast t r i p comparat e n t l y l a r g e s i g n a l w a s applie t r i p comparator modules t o e Operation of t h e r e l a y

n t h e s a f e t y c i r c u i t s . .Correction of t h e producing components elsewhere i n t h e . . e t o a very t o l e r a b l e - l e v e x .

found t o be inoperative i f a s u f f i c i e input.

A diode w a s added t o t h e fast

e t h i s difficulty. x i n t h e nuclear system generated con-

s i d e r a b l e noise, making it d i f f i c u l t t o reset t h e safety-system channels. The resistor-diode combination f o r damping t h e voltage induced by r e l a y operation was replaced by a Zener diode and a diode combination t h a t was

more s a t i s f a c t o r y .

398

Table 24.3

Results of Periodic Tests of Safety C i r c u i t s

Date

Troubles Detected

5124165

Noted t h a t PR-522 and PI-522 d i d not agree.

6/9/65

Two s a f e t y channels t r i p p e d when t e s t i t i g one channel, scram setpoint was at 120% i n s t e a d of 150% and low current t e s t did not function properly.

1/15/66

Fuel pump pressur? switch s e t p o i n t needed r e s e t t i n g .

4/10/66

Reactor c e l l pressure switch s e t p o i n t needed r e s e t t i n g .

5/20/66

Two sampler-enricher pressure switches s e t p o i n t s needed resetting.

6/12/66

Sampler-enricher pressure switch needed r e s e t t i n g .

6/15/66

Rod motion was jerky.

9 119/66

Two sampler-enricher access door l a t c h e s d i d not function properly.

11/2/66

A s a f e t y channel would not r e s e t .

12/16/66

Reactor c e l l pressure switch s e t p o i n t needed r e s e t t i n g .

11/27/67

Burned-out i n d i c a t o r l i g h t i n s a f e t y channel.

5/13/69

Defective solenoid c o i l on r e a c t o r c e l l b l o c k valve header.

9/8/69

Reactor o u t l e t s a f e t y i n t e r l o c k would not c l e a r .

11/24/69

Sampler-enricher pressure switch s e t p o i n t needed r e s e t t i n g .

399 I n t h e summer of 1967, t h e l-kW 48-V-dc

t o 120-V-ac i n v e r t e r , which

supplies ac power t o one of t h e t h r e e s a f e t y channels, f a i l e d during switching of t h e 48-V dc supply. power t r a n s i s t o r s .

It was repaired by replacement of two

During t h i s same period t h e output of t h e ion chamber

i n s a f e t y channel-2 decreased d r a s t i c a l l y during a nonoperating period, and t h e chamber w a s replaced p r i o r t o resumption of operations.

The trouble

proved t o be a f a i l u r e i n a glass s e a l t h a t allowed water t o e n t e r .the mag, nesia i n s u l a t i o n i n t h e cable, which i s & i n t e g r a l p a r t of t h e chamber. .A period s a f e t y amplifier f a i l e d when l i g h t n i n g struck t h e power l i n e t o

t h e r e a c t o r s i t e , and a replacement amplifier f a i l e d as it w a s being installed.

The f i e l d - e f f e c t t r a n s i s t o r i n t h i s type of'amplifier was sus-

c e p t i b l e t o damage by t r a n s i e n t voltages, and it was found t h a t under some conditions, damaging t r a n s i e n t s could be produced when t h e amplifier was removed from o r i n s e r t e d i n t o t h e system. signed, t e s t e d , and installed:

A protective c i r c u i t was de-

The module replacement procedure w a s modi-

f i e d t o reduce the p o s s i b i l i t y of damage incurred on i n s t a l l a t i o n of t h e module.

Two r e l a y s i n t h e s a f e t y r e l a y matrices f a i l e d , both with open

coil circuits.

A chattering contact on t h e fuel-pump motor current r e l a y

a f e t y channel 2 t o t r i p s e v e r a l times before t h e problem w a s overcome by p a r a l l e l i n g two contacts on t h e same relay. t h e core o u t l e t temper

A defective switch on

e a l s o caused several channel t r i p s and one re-

a c t o r scram before t h e trouble was i d e n t i f i e d and t h e switch w a s replaced. A wiring e r r o r i n a s a f e t y c i r c u i t was discovered and corrected.

Interlocks had recently been added i n t h e "load scram" channels t o drop t h e load when the control rods scram.

A wiring design e r r o r r e s u l t e d i n these interlocks

being bypassed by a s a f e t y

er.

Although t h e c i r c u i t s were wired t h i s way f o r a time before being discovered, t h e scram i n t e r l o c k s were always operative during power operation, s It

Operate" mode when any s a

L a t e i n 1965, t h e pow

2623 r e l a y safety ele-

ments w a s changed from 115-V-ac t o 32-V-dc. ac pickup on t h e other mod routed.

This w a s done t o eliminate

twough which t h e r e l a y c o i l current was

The 115-V-ac relays were not changed a t this time.

These f b c tioned s a t i s f a c t o r i l y u n t i l t h e summer of 1967 when two of t h e 1 5 relays failed. By t h e end of 1967, seven had f a i l e d , a l l w i t h open c i r c u i t s o r

400

s a f e conditions. f o r 32-V-dc

I n April 1968, all 15 were replaced with relays designed

operation.

Soon afier i n s t a l l a t i o n , 3 of these f a i l e d due t o

contact welding (unsafe condition).

This was apparently due t o e a r l y

There were no more f a i l u r e s during extensive

failure of defective relays.

tests made at t h i s time o r during subsequent operation.

S t a r t i n g Septem-

ber 30,. 1968, d a i l y tests were conducted on t h e e n t i r e rod scrams r e l a y matrix t o detect s i n g l e failures.

(The matrix had been t e s t e d weekly be-

fore t h a t . )

Moise suppressors were i n s t a l l e d across some o f t h e non-

s a f e t y contacts of these relays t o alleviate t h e noise which had sometimes caused d i f f i c u l t y during in-service tests. 24.6

Performance of t h e Wide-Range Counting Channels

I n i t i a l c r i t i c a l i t y tests disclosed t h a t t h e neutron f l u x attenuation i n the instrument penetration did not follow an i d e a l exponential curve. The deviation w a s t o o l a r g e t o be adequately compensated f o r by the verni-

stats i n t h e wide-range counting channels.

This i s i l l u s t r a t e d by curve

A of Fig. 24.2 which shows f i s s i o n counter response (normalized count rate)

vs w i t h d r a w a l from t h e lower end of t h e penetration i n guide tube 6.

w a s reasonable t o conclude from t h i s , and from similar curves, t h a t t h e excess neutrons responsible f o r t h e d i s t o r t e d p a r t of t h e curve were ent e r i n g t h e penetration along i t s length.

The count r a t e s i n other guide

tubes nearer t h e upper h a l f of the penetration were even more d i s t o r t e d than curve A.

Since a flux f i e l d w i t h attenuation p e r Curve A precluded

successful operation of t h e wide-range counting instrumentation, s h i e l d s of sheet cadmium were i n s e r t e d i n guide tubes 6 and 9 t o s h i e l d t h e f i s s i o n chambers from s t r a y neutrons.

Curve B of Fig. 24.2, normalized count rate

vs distance, shows t h e improvement f o r guide tube 6. Reactor period information from t h e wide-range, counting channels inh i b i t e d rod withdrawal and caused rod reverses.

Reactor period s i g n a l s

from counting channels operating at l o w input l e v e l s a r e characterized by slow response.

This inherent delay produced a problem w i t h servo-controlled

rod withdrawal during start.

I n a servo-controlled start, t h e demand sig-

n a l caused t h e regulating rod t o w i t h d r a w u n t i l t h e period-controlled "with-

d r a w i n h i b i t " i n t e r l o c k operated.

If t h e period continued t o decrease, t h e

c . .

REACTOR SHIELD-.

MIDPLANE OF =RE-

REACTQR VESSEL

CADMIUM SHIELDING SUBASSEMBLY-TYPICAL IN GUIDE TUBES 6 AND 9 SECTION A-A LOCATION OF GUIDE TUBES OUTER SHELL REMOVED FOR CLARITY- SHADED WEDGES ARE CADMIUM -UNSHADED WEDGES ARE ALUMINUM

Fig. 24.2

Guide Tube Shield in the MSRE Instrument Penetration

402 I1

reverse" interlock acted t o i n s e r t the rods i n d i r e c t opposition t o the

servo demand.

These "withdraw inhibit" and "reverse" t r i p points were

originally established at periods of +20 and +10 sec respectively.

The

delayed low-level response of t h e "inhibit" interlock allowed s u f f i c i e n t incremental rod withdrawal t o produce a 10-sec period and thus cause a reverse.

The s i t u a t i o n was aggravated by coasting of t h e shim-locating motor

i n the regulating rod l i m i t switch assenibly,. . To correct t h i s , t h e "withdraw inhibit" and "reverse" period t r i p points were changed t o +25 and +5 sec, respectively , an electro-mechanical clutch-brake was inserted i n t h e s h i m locating motor-drive t r a i n and dynamic braking c i r c u i t r y w a s i n s t a l l e d f o r the regulating rod drive motor. Throughout operation, d i f f i c u l t y w a s experienced with moisture penet r a t i o n i n t o t h e f i s s i o n chamb-ers.

The cause w a s diagnosed as excessive

s t r a i n and flexing of t h e tygon tubing sheath on the e l e c t r i c a l cables. The average l i f e t i m e was about 6 months p r i o r t o July 1968. A t t h i s time the type of tygon tubing used t o cover the cable and t h e method of sealing - - - ..

the chamber connections were changed.

This seemed t o improve t h e moisture

resistance. I n early 1967, a failure occurred due t o a short i n t h e cable t o the preamplifier.

6iiJ

The drive tube u n i t was modified t o provide f o r controlled

cable bends and there were no reoccurrences of t h i s type of d i f f i c u l t y . 24.7

Performance of the Linear Power Channels

The l i n e a r power channels and rod servo instrumentation performed very well throughout the operation.

When the f u e l was changed t o 233U, t h e rod-

control servo was modified t o allow an increase i n t h e servo dead band t o compensate f o r t h e more rapid flux response of t h e reactor. The compensated ion chambers use a small e l e c t r i c motor t o change compensation.

Early i n 1966 one of these motors had t o be replaced.

A water leak occurred i n one. of t h e compensated ion chambers i n the summer of 1969. When it w a s examined, numerous leaks were found along t h e seam weld i n t h e 321 s t a i n l e s s s t e e l bellows sheathing t h e cable, and the aluminum can at the outer support r i n g was honeyconibed by corrosion.

(The

water i n the instrument shaf't contained lithium n i t r i t e - b u f f e r e d with boric acid f o r inhibition of corrosion. )

403 24.8 Because o

B F a Nuclear Instrumentation

very unfavorable geometry, t h e strongest pract,a

-e neu-

t r o n source would not produce 2 counts/sec * f r o m t h e f i s s i o n counters i n t h e wide-range counting channels u n t i l t h e core vessel was approximately h a l f full of f u e l s a l t ; neither would it produce 2 counts/sec with f l u s h s a l t i n t h e core a t any level.

This was t h e minimum count r a t e required

t o obtain t h e permissive "confidence" i n t e r l o c k which allows f i l l i n g t h e core vessel and withdrawing t h e rods.

Therefore a counting channel using

a s e n s i t i v e B F 3 counter was added t o e s t a b l i s h "confidence" when $he core

vessel was l e s s than h a l f - f i l l e d with f u e l salt. i n 1966.

This was i n s t a l l e d e a r l y

The chamber had t o be replaced i n t h e summer of 1967 due t o mois-

ture leakage i n t o t h e cable. 24.9

No other troubles were encountered.

Nuclear Instrument Penetration

The nuclear power produced at the MSRE was determined by an overall

system heat balance.

All nuclear power instruments were c a l i b r a t e d t o agree

with t h i s primary standard.

During extended runs at higher powers, t h e

nuclear instruments indicated 15 t o 20% higher than t h e heat balance.

This

w a s found t o be due t o a r i s e i n water temperature i n t h e nuclear i n s t r u ment penetration which apparently changed t h e attenuation c h a r a c t e r i s t i c s of t h e water.

I n June 1966, a heat exchanger system was placed i n operation

to cool the water which reduced t h e temperature change f r o m zero t o full

power t o % 1 8 째 ~ ~ f r o 72 m 'F

and reduced t h e difference i n t h e two power meas-

urements t o about 5%.

elements had t o be replace p a i r s were necessary on t h e e l

ing t h e e n t i r e operation. onics.

Occasional re-

One rod and load scram r e s u l t e d

f r o m a f a l s e s i g n a l from one of these monitors (RE-528).

404 .

24.11 Performance of t h e Personnel Radiation . Monitoring and Building Evacuation System During e a r l y t e s t i n g it was found t h a t t h e r e were some areas where

Two additionalâ&#x20AC;&#x2122;horns

the building evacuation horns could not be heard. -.

and four additional beacon a l a r m lights were i n s t a l l e d .

Other than oc-

casional &nor r e p a i r s , t h e system has functioned s a t i s f a c t o r i l y . 24.12

Performance of the. Stack Monitoring System

The stack monitoring system was very reliable.

The manual range

switching made it d i f f i c u l t t o i n t e r p r e t data from t h e recorder charts and t h e Rustrak recorders were very inconvenient t o use. I

24 e 13 Performance of Thermocouples- and,the. . . Temperature IZtadout-and-f3ontrol Inshrumentation- -

A t o t a l of 1071 thermocouples were i n s t a l l e d a t t h e MSRE.

O f these,

866 were on salt systems (351 on t h e c i r c u l a t i n g loops and 515 on t h e drain tanks, drain l i n e s , and fieeze valves). i n f i v e years of service. maintenance.

O f t h e 1071, only 12 have failed

Five others were damaged during construction and

A breakdown of t h e f a i l u r e s i s given i n Table 24.4.

Table 24.4

....

Failufes Among t h e 1071 MSRE .Thermocouples

. Nature of Failure

Number

Damaged during construction

Damaged during maintenance Lead open during operation

Abnormally l o w reading (detached? )

Unknown reason

.. .. .. .. ......

LE4

Total

405 \

Only three thermocouple wells were provided i n t h e c i r c u l a t i n g salt systems, one each i n the coolant radiator i n l e t and o u t l e t pipesâ&#x20AC;&#x2122;and one

i n t h e reactor neck. o r vessel walls.

The remaining thermocouples were attached t o t h e pipe

The thermocouples on t h e radiator tubes were insulated

t o protect them from the e f f e c t s of the high-velocity a i r t h a t flows over them during power operation; the others were not insulated and thus were

subject t o e r r o r because of exposure t o heater shine and t o thermal coni

vection flow of t h e c e l l atmosphere within t h e heater insulation.

March 1965, with t h e f u e l and coolant systems c i r c u l a t i n g salt at isothermal conditions, a complete s e t of readings was taken from a l l t h e thermocouples t h a t should read the temperature of t h e c i r c u l a t i n g s a l t .

data was taken i n June 1967 at t h e start of R u n 12. s e t s of measurements are shown i n Table 24.5.

A similar s e t of

The r e s u l t s of t h e two

Comparison of t h e standard

deviations f o r t h e radiator thermocouples with those f o r t h e other thermocouples shows the e f f e c t of insulation on reducing t h e s c a t t e r .

Comparison

of t h e sets of data taken over two years apart shows very l i t t l e change, certainly no greater s c a t t e r .

Figure 24.3 shows t h a t t h e s t a t i s t i c a l dis-

t r i b u t i o n of t h e deviations on individual thermocouples from t h e mean also changed l i t t l e i n the two years. P

Table 24.5 Comparison of Readings of Thermocouples of S a l t Piping and Vessels Taken with the S a l t Isothermal

Thermocouple

Radiator tubes Other

1206.73 12.3 1102.3 f 10.6

1207.4 k 9.8

406

ORNL- DWG 67-41786

400

I-

DATA TAKEN MARCH i965

5 60 r

I-

DATA TAKEN JUNE 4967 1

$ a W

2 w

.-60

-40

-20

DEVIATION FROM AVERAGE (" F) Fig. 24.3

Comparison of MSRE Thermocouple Data fromMarch 1965 and June 1967

407

The s c a t t e r i n t h e various thermocouple readings w a s reduced t o an acceptable l e v e l by using biases t o correct each reading t o The o v e r a l l average measured while both f u e l and coolant systems were c i r c u l a t i n g s a l t These b i a s e s were entered i n t o t h e computer and

at isothermal conditions.

were axtomatically applied t o t h e thermocouple readings.

"he biases were

revised at t h e beginning of each run and were checked when isothermal cond i t i o n s e x i s t e d during t h e runs. were reliable,

Generally t h e biased thermocouple readings

However i n a few cases, t h e r e were s h i f t s which caused cal-

culational errors. During e a r l y operation t h e thermocouple scanner gave considerable d i f f i c u l t y due t o 60-cycle noise pickup, poor s t a b i l i t y , and d r i f t i n g of t h e salvaged oscilloscopes.

Refinements were needed i n design t o allow b e t t e r

i d e n t i f i c a t i o n of scanner points and provide a means by which t h e operator could c a l i b r a t e t h e instruments. operated very s a t i s f a c t o r i l y .

A f t e r t h e s e were corrected, t h e system

The r o t a t i n g mercury switches l a s t e d much

longer than t h e expected 1000-hour mean l i f e .

One switch failure occurred

when t h e nitrogen purge gas w a s inadvertently stopped.

6.,

Single point E l e c t r a Systems alarm switch modules w e r e used f o r cont r o l of freeze valves and f o r other alarm and control actions.

These gave

considerable trouble during e a r l y operation due t o d r i f t i n g o r dual setpoints and general maloperation. correct these.

A number of modifications were made t o

Printed circuit-board contacts were gold-plated t o reduce

contact r e s i s t a n c e , t h e t r i m pots used f o r h y s t e r e s i s adjustment were replaced with fixed r e s i s t o r s , and r e s i s t o r values i n modules having ambig\

uous (dual s e t p o i n t s were changed t o r e s t o r e t h e proper b i a s l e v e l s ,

These

changes , together with s t a b i l i z a t i o n , by aging, of c r i t i c a l r e s i s t o r s i n t h e switch modules and more t h e performance.

A check sh

ous periodic t e s t i n g procedures improved t h a t out of 109 switch s e t p o i n t s , 832 had

s h i f t e d less than 20째F over a six-month period.

Multiple s e t p o i n t s s t i l l

occurred and various o t h e r failures were encountered.

1969, records were kept on t h e

During 1968 and

lures of i n d i c a t o r l i g h t s on these mol

dules.

There were about 50 f a i

es per year.

This w a s important because

a burned-out l i g h t bulb could cause alarm o r control action.

408

24.14

Performance of Pressure Detectors

Most of the pressure instruments at t h e MSRE performed very well.

Dif-

f i c u l t y was encountered with t h e d i f f e r e n t i a l pressure c e l l used t o obtain the pressure drop i n the helium f l o w tlirough t h e charcoal beds (PdT-556). The span and zero s e t t i n g s s h i f i e d badly although t h e pressure capability had hever been exceeded.

Three dp c e l l s f a i l e d i n t h i s service,

Two of

-these were removed, t e s t e d , and inspected without determining t h e cause of ithe trouble.

The last replacement functioned s a t i s f a c t o r i l y at first but It i s s t i l l i n s t a l l e d .

then gave similar d i f f i c u l t i e s . 24.15

Performance of Level Indicators

The bubbler-type l e v e l instruments used i n t h e f u e l pump overflow tank More details are given in-Sections 5.8

and coolant pumps performed well.

and 6.5.

Some d i f f i c u l t y was encountered i n controlling t h e purge flow un-

t i l t h e t h r o t t l i n g valves were replaced. A high and low level resistance type l e v e l probe was provided on each

drain tank.

During early operation t h e excitation and s i g n a l cable leads

on both probes of t h e m e 1 flush tank failed. . .

These failures were caused

by excessive temperature which caused oxidation and embrittlement of t h e copper-clad,

mineral-insulated copper-wire cables.

These cables were de-

signed on t h e assumption t h a t they would be souted i n a i r .above t h e tank insulation and t h a t t h e i r operating temperature would not exceed 200'F; however, i n the actual i n s t a l l a t i o n , t h e cables were covered with insulat i o n 'and. t-..h e temperature at t h e point of attachment t o t h e probe was probably i n excess of 800'F. Repairs were accomplished by replacing t h e copperclad, mineral-insulated copper wire excitation and s i g n a l cables and port i o n s of t h e probe head assembly with a stainless-sheathed, nickel-wire cable assembly. .. f u e l drain tank No. 1 failed. wire inside t h e c e l l .

ceramic-beaded

I n t h e summer of 1968, one of t h e probes i n The failure was found t o be an open lead

The probe was restored t o service by a cross con-

nection outside t h e c e l l t o t h e equivalent lead of t h e other probe. The' i n i t i a l instrumentation provided t o assure proper water l e v e l i n -..

the' vapor-condensing tank consisted of 4 resistance type probes spaced

.. . .

409

bii

4 in. apart near t h e desired t h e instrumentation s t a r t u p

er l e v e l . I n September 1966, while doing . . . . .... k l i s t , one o f t h e two high-level switches

was found t o be defective.

ss of t h i s switch caused t h e l o s s of one chan-

n e l of information needed

stablish t h a t t h e water l e v e l i n t h e tank was

Since a second switch failure might require t h a t t h e r e a c t o r be

correct.

shut down &til the switches could be repaired, and since t h e removal of t h e switches fkom t h e tank is a d i f f i c u l t operation, a bubbler-type l e v e l measuring system, which included a containment block valve and associated s a f e t y c i r c u i t s i n t h e purge supply, w a s designed and was i n s t a l l e d i n t h e reactor-cell vapor suppression tank.

This i n s t a l l a t i o n u t i l i z e d a d i p tube

which w a s included i n t h e o r i g i n a l design i n a n t i c i p a t i o n of such need. This new l e v e l system a l s o enabled t h e operator t o check t h e water l e v e l i n t h e tanks as a routine procedure. 24.16

Performance of t h e Drain Tank Weighing, Systems

The same type pneumatic

eighing devices were used on t h e f u e l drain

tanks and t h e coolant drain tank.

The coolant drain tank weight indicators

proved t o be stable, showing no long-term drifts o r e f f e c t s of external variables.

With

5756 l b of salt i n t h e tank a t about 1200°F, t h e extreme

spread of 40 indicated weights over a period of a week was k 22 l b .

This

was only k 0.4% and was q u i t e s a t i s f a c t o r y . The indicated weights of t h e t h r e e tanks i n t h e fuel system exhibited rather large unexpl

300 pounds.

The mechani

probably was due t o chan

tures of attgched pipin sure seemed t o a l s o affe

I n some cases these amountedto 200 t o using t h i s was not d e f i n i t e l y established, but forces on t h e syspended tanks as temperatank f’urnaces changed.

e weighing s y

t h e multipositlon pne complished by s e l e c t i n g l e c t o r valves.

The

ion d r i f t , d i f f i c u l t y w a s experienced with t o r switches.

Manometer readout was ac-

i c u l a r weigh c e l l channel’with pneumatic seomposed of a stacked array of individual.

valves operated by cams on t h e operating handle s h a f t .

valves gave false weight indications. solved t h i s problem.

ms were useful

ng and d r a i n i

i n observing t r a n s f e r s of I n addition t o t h i

Reactor c e l l pres-

Leaks ‘in these

A redesign of t h e switching device

410

P r i o r t o power operation, one of t h e weigh c e l l s f a i l e d and was replaced. '

The f a i l u r e was determined t o be due t o p i t t i n g of t h e b a f f l e and

nozzle i n t h e c e l l .

This p i t t i n g w a s apparently caused by amalgamation of

mercury with t h e p l a t i n g on t h e b a f f l e and nozzle.

How t h e mercury got in-

t o the weigh c e l l s has not been determined; however, it was believed t o have come from.the manometers and t o have been p r e c i p i t a t e d on t h e b a f f l e by expansion c'ooling of t h e air leaving t h e nozzle.

Appreciable q u a n t i t i e s

of mercury were a l s o found i n t h e tare pressure regulators on t h e control panel; however, no mercury was found i n t h e interconnecting tubing o r i n other portions of t h e system. 24.17

Performance of t h e Coolant S a l t Flowmeters

Several days a f t e r t h e start of coolant salt c i r c u l a t i o n , t h e output of one of t h e two salt.flowmeter channels s t a r t e d d r i f t i n g down s c a l e , output of t h e other c h p n e l remained Steady.

The

The t r o u b l e w a s i s o l a t e d t o

a zero s h i f t and possibly a span s h i f t i n t h e NaK-filled d i f f e r e n t i a l pressure t r a n s m i t t e r i n thk drif'ting channel.

Since t h e exact cause 8could not

be determined, a spare dp c e l l was i n s t a l l e d .

Both channels f'unctioned

s a t i s f a c t o r i l y throughout t h e remainder of t h e r e a c t o r operations.

Tests

on t h e defective u n i t were inconclusive, however,- it was determined t h a t t h e s h i f t s were temperature-induced zero s h i f t s possibly caused by incomp l e t e f i l l i n g o r a leak i n t h e s i l i c o n e o i l portion of the-instrument,

An operational inconvenience persisted'throughout'operations,

When-

ever t h e r a d i a t o r air flow was increased, a.+i r leaking through t h e insulat i o n around t h e salt l e g s t o t h e dp c e l l s caused t h e temperature t o decrease rapidly.

This necessitated adjustment of t h e heaters.

Due t o t h e discrepancy between t h e r e a c t o r power l e v e l indicated by f u e l burnup, heat balance, e t c , , it is planned t o r e c a l i b r a t e t h e dp c e l l s during t h e next f i s c a l year.

24.18

411

Performance of Relays

The d i f f i c u l t i e s encountered with t h e control rod r e l a y s a r e described i n Section 24.5.

Experience with other relays i s given below.

A f t e r about 2 years of operation, t h e 48-V-dc-operated

considerable heat damage t o t h e i r b a k e l i t e frames.

relays showed

The manufacturer,

General E l e c t r i c , advised t h a t overheating of t h i s p a r t i c u l a r model was

a common problem i f t h e relays were continuously energized.

Early i n 1967,

twenty of t h e 139 relays were replaced with a l a t e r improved model. a f e w months some of these a l s o showed signs of deterioration.

Within

Therefore

i n June 1967, a l l 139 relays were field-modified by replacing t h e b u i l t - i n r e s i s t o r s with externally mounted r e s i s t o r s .

No trouble was experienced

a f t e r t h i s modification. I n September 1969, t h e load-scram c i r e d . t , which drops t h e r a d i a t o r doors and stops t h e blower, t r i p p e d several times.

Investigation showed

t h a t some of t h e r e l a y contacts had developed unusually high resistance

due t o oxide films.

Because of t h e way t h e contacts were p a r a l l e l e d i n

t h e matrix, t h e film was not burned o f f each time t h e contact closed, as i n a normal application.

These contacts were cleaned and no f u r t h e r dif-

f i c u l t y was encountered.

24.19 Two "on-site''

Training Simulation

r e a c t o r k i n e t i c s simulators w e r e developed f o r t h e pur-

pose of t r a i n i n g t h e MSRE operators i n nuclear s t a r t u p and power operation. 'Phe s t a r t u p simulator used t h e control rod p o s i t i o n s i g n a l s as inputs, and provided outputs of log count r a t e , period, l o g power, and l i n e a r power. The r e a c t o r ' s period interlocks, f l u x control system, and l i n e a r f l u x range s e l e c t o r were a l s o operational.

I n addition t o t h i s , t h e power l e v e l simu-

l a t o r used t h e r a d i a t o r door position and cooling a i r pressure drop signals

as inputs and provided readout of key system temperatures.

Both simulators

were s e t up on general purpose, portable EA1 TR-10 analog computers.

Much

of t h e a c t u a l MSRE hardware, such as control rods were used r a t h e r than simulated.

Thus t h e operators manipulated t h e a c t u a l r e a c t o r controls and

became used t o t h e instrumentation and controls system.

a very r e l i a b l e t r a i n i n g t o o l .

This proved t o be

412

24.20 . Miscellaneous The original mass spectrometer usedto monitor for beryllium in the coolant stack was replaced with an improved instrument near the start of power operation. Only occasional repairs or preventative maintenance was required since then. Folded charts were used on some recorders. These did not function properly due to the low chart speeds being used.

24.21 Conclusions and Recommendations Considering the quantity and complexity of the instrumentation, a minimal amount of difficulty was encountered. Since the MSRE was an experimental reactor, it was in m a y areas over-instrumented. This was probably due to hot-knowingwhat information might be needed and not taking enough credit for the ability-of the operators. This led to unnecessary difficulties in normal operations or in running special experiments. At the same time, there were areas where additional information'would have been beneficial

413 REFERENCES 1.

M. W. Rosenthal, P. Re Kasten, and R. B, Briggs, "Molten-Salt Reactors History, Status d Potential," NucZ. AppZ. Tech., E, 107 (1970)

MSR Program Semiannu. Progr. Rep., Jan. 31, 1964, OWL-3626.

MSR Program Semiannu.

4. MSR Program Sem&mnu.

gr. Rep,, July 31, 1967, ORNL-3708.

. Rep.,

Feb. 28, 1965, ORNL-3812.

5. MSR Program Semhanu.

. Rep.,

Aug. 31, 1965, OWL-3872.

MSR Program Semiannu.

. Rep,,

Feb. 28, 1966, ORNL-3936.

MSk Program Sembanu. Progr. Rep., Aug. 31, 1966, ORNL-4037.

8. MSR Program Sem&mnu. Progr. Rep., Feb, 28, 1967, ORNL-4119. 9.

MSR Program Semiannu.

g r . Rep., Aug. 31, 1967, OWL-4191.

10. MSR Frogram S e ~ a n n u .Progr. Rep., Feb. 29, 1968, ORNL-4254. 11. MSR Program Sem&mnu.

hi,

gr. Rep., Aug. 31, 1968, ORNL-4344-

12.

MSR Program Sermhnnu. Progr. Rep., Feb. 28, 1969, ORNL-4396.

13.

MSR Program Sem&mnu, Pr

14. MSR Program Semiannu. 15.

R. C. Robertson, MSRE

16.

R. â&#x201A;ŹI Guymon, .

Rep., Aug. 31, 1969, ORNL-4449. Rep., Feb. 28, 1970, OW-4548. and Operations Report, Part I -De-TM-728 (Jan. 1965). enreich, and J. R. Engel, MSRE Design and Program, ORNL=TM-911 (Nov. 1966). he MSRE Flush

Fuel Salts, ORNL-

Houtzeel, Spray, Mist, Bubbles 7 (June 1970) nation of the

on, Plans for 2974 (April 1970). urement and Heat-Transfer Performance

21,

J. R. Engel, MSRE Design and Operations Report, Part XI-A- Test Program for U'" Operation, ORNL-TM-2304 (Sept. 1968).

414 References (continued) 22.

R, H. Guymon, MSRE Design and Operations Report, Part VIII, Operating Procedures, Vols, I and 11, ORNL-TM-908 (Dec. 1965).

23.

R. E. Thoma, Chemical Aspects of MSRE Operations, 0-4658

24.

R. H. Guymon and P, N, Haubenreich, MSRE Design and Operations Report, Part VI, Operating Safety Limits for the Molten-Salt Reactor Experiment, ORNL-TM-733 (3rd revision), (July 25, 1969).

25.

R. H. Guymon, R. C. Steffy, Jr., and C. H. Gabbard, Preliminary Evaluation of the Leak in the MSRE Primary System Which Occurred during the Final Shutdown, ORNL-CF-70-4-24, (April 22, 1970).

26.

P. G. Smith, Development of Fuel- and Coolant-Salt Centrifugal Pumps for the MSRE, ORNL-TM-2987 (Oct. 1970).

27.

C. H. Gabbard, Inspection of MSRE Fuel Circulating Pump after the Zero-Power Experiments (Run 31, ORNL-CF-66-8-5 (Aug. 1966).

28.

R. B. Briggs, Measurement of Strains in Heat Exchanger Nozzle and Piping of MSRE, ORNL-CF-66-2-66, (Feb. 1966).

29.

C. H. Gabbard, Thermal-Stress and Strain-Fatigue Analyses of the MSRE Fuel and Coolant Pump Tanks, ORNL-TM-78 (Oct. 1962).

30.

C. H. Gabbard, R. J. Kedl, and H. B. Piper, Heat Transfer Performance of the MSRE Heat Exchanger and Radiator, ORNL-CF-67-3-38 (Mar. 1967).

31.

C. H. Gabbard, Reactor Power Measurement and Heat Transfer Performance in the Molten Salt Reactor Experiment, O R N G W 3 0 0 2 (May 1970).

32.

H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastelloy N Surveillance Specimens -First Group ORNL-TM 1997 (Nov. 1967).

33.

H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastelloy N Surveillance Specimens -Second Group ORNL-TM-2359 (Feb, 1969).

34.

H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastelloy N Surveillance Specimens - Third Group, ORNL-TM-2647 (Jan. 1970)â&#x20AC;&#x2122;.

35.

H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastelloy N Surveillance Specimens -Fourth Group, ORNL-TM-3036 (Mar. 1971).

36.

H. E. McCoy and B. McNabb, Intergranular Cracking of INOR-8 in the MSRE, ORNL-4829, (Nov. 1972).

37.

C. H. Gabbard, Design and Construction of Core Irradiation-Specimen Array for MSRE Runs 19 and 20, ORNL-%2743 (Dec. 1969).

(Dec. 1971).

415 References (continued) 38. R. B. Briggs, Effects of Irradiation of Service Life of MSRE, ORNLCF-66-5-16 (May 1966). 39. R. B. Briggs, Assessment of Service Life of MSRE, ORNL-CF-69-8-3, (Aug. 1969).

40, B. E. Prince et al., Zero-Power Physics Experiments on the Molten-Salt Experiment, ORNL-4233 (Feb. 1968).

41. J, R. Engel and B. E. Prhce, Zero-Power Experiments with 2ssU in the MSRE, ORNL-TM-3963 (Dec. 1972). 42. C. H. Gabbard and C. K. McGlothlan, A Review of Experience with the MSRE Main Blowers, ORNL-CF-67-4-1, (Abg. 1967). +

0 43. C. H. Gabbard, Bearing Failures of Main Blowers MB-1 and MB-2, CF-67-4-1, Addendum-l mar. 19683.

44. R. L. Baxter and D. L. Bernhard, "Vibration Tolerances for Industry," paper presented at ASME Plant Engineering and Maintenance Conference, Detroit, Michigan, Paper No. 67-PEM-14 (Apr. lW12, 1967).

45. E. C. Parrish and R, W. Schneider, Tests of High Efficiency Filters and Filter Installation at ORNL, ORNL-3442, (May 1963).

46. H. R. Payne, The Mechanical Design of the MSRE Control Rods and Reactor Access Nozzle, MSR-61-158 (June 1961). 47. J. L. Crowley, Temperature and Salt Levels in the MSRE Reactor Access Nozzle - A Possible Relation to MSRE Power Blips, MSR-69-13 (Feb. 1969).

48. P. N. Haubenreich, R. Blumberg, and M. Richardson, "Maintenance of the Molten-Salt Reactor Experiment" paper presented at Winter Meeting, A N S , Washington D. C. (Nov, 1970).

49. C. H. Gabbard, Radiation Dose Rates to the MSRE Fuel-Pump Motor and Control Rod Drives, MSR-65-48. 50. R. C. Robertson, MSRE pp. 244-75 (Jan. 1965).

gn and Operations Report, 0W-m-728,

51. R. H. Guymon, MSRE Design and Operations Report, Part VII, Operating Procedures, ORNL-TM-908, Vol. I1 (Dec. 1965). 52. R. B. Gallaher, MSRE Sampler-Enricher System Proposal, ORNL-CF-61-5-120, (May 1961). 53.

MSRE S t a f f , MSRE Sampler-Enricher, An Account of Recent Difficulties and Remedial Action, MSR-67-73 (Sept. 1967).

References (cpntinued) 54.

P. N. Haubenreich and MSRE Staff, An Account of Difficulties with the Sampler-Enricher that Led t o a Second Capsule Being Left in the Pump Bowl, MSR-68-125 (Sept. 1968).

55.

S. E. Beall et al., MSRE Design and Operations Report, Part V, Reactor Safety Analysis Report, ORNL-TM-732 (Aug. 1964)

56.

R. H. Guymon, Tests of the MSRE Reactor and Drain Tank Cells, MSR62-95 (Nov. 1962).

57.

P. N. Haubenreich et al., MSRE Design and Operations Report, Part 111, Nuclear Analysis, ORNL-TM-730 (Feb. 1964)

58.

Go Bo Burger, J, R. Engel, and C, De Martin, Computer Manual for MSRE Operators, ORNL-CF-67-1-28 (Jan. 1967)

59.

P. N. Haubenreich, Safety Considerations in R6sumption of MSRE

Operation, ORNL-CF-69-8-10

(Aug. 1969).

60.

P. N. Haubenreich, Responses to RORC Recommendations for MSRE, MSR69-104

61.

B. F. Hitch et al., Tests of Various Particle Filters for REmoval of Oil Mists and Hydrocarbon Vapor, ORNL-TM-1623 (Sept. 1966).

62.

W. C. Ulrich, Purge of MSRE Off-gas Line 522, MSR-66-8, (Mar. 1966).

63.

R. H. Guymon and J. R. Engel, Introduction of Oxygen in Fuel-Drain Tank, MSR-67-1 (Feb. 1967).

417 OWL-TM-3039 I n t e r n a l Distribution

1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11.

h*,

12. 13. 14-18. 19. 20. 21. 22. 23. 24. 25. 26.

S. E. Beall S. Bettis

E. R. R. W. J. F.

Blumberg B. Briggs B. C o t t r e l l L. Crowley L. Culler S. J. Ditto J. E. Engel D. E. Ferguson A. P. Fraas C. H. Gabbard W. R. Grimes R. H. Guymon P. H. Harley P. N. Haubenreich H. W. Hoffman T. L. Hudson P. R. Kasten A. I. Krakoviak K e r m i t Laughon, AEC-OSR M. I. Lundin

27. 28. 29. 30. 31. 32. 33. 34. 35. 36-37 38. 39. 40. 41. 42. 43. 44 45. 46. 47-48. 49. 50-52. 53.

R. N. Lyon R. E. MacPherson H. E. McCoy H. C. McCurdy A. J. Miller R. L. Moore L. C. Oakes A. M. Perry M. Richardson M. W. Rosenthal Dunlap Scott M. R. Sheldon M. J. Skinner I. Spiewak D. A. Sundberg J. R. Tallackson R. E. Thoma D. B. Trauger G. D. Whitman Central Research Library Y-12 Document Reference "action Laboratory Records Department Laboratory Records (RC)

External Distribution

54-56. 57-58. 59-60. 61. 62. 63. 64. 65. 82. 83.

85.

Director, Division of Reactor Licensing, USAEC, Washington D.C. 20545 Director, Division of Reactor Standards, USAEC, Washington, D.C. 20545 N. Haberman, USAEC, Washington, D.C. 20545 D. F. Cope, AEC-ORO M. Shaw, USAEC, Was ton, DOC. 20545 David E l i a s , USAEC, Washington, D.C. 20545 A. Houtzeel, TNO, 176 Second Ave., Waltham, Mass. 02154 R. C. S t e f f y , TVA, 303 Power Building Chattanooga, Tenn. 37401 Manager, Technical Information Center, AEC (For ACRS Members) Research and Technical Support Division, AEC, OR0 Technical Informatio