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LEGAL N O T I C E This 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. makes any warranty, express or implied, -or assumes any kgal liability or responsibility for the accuracy, completeness or usefulness of any information. apparatus, product or proms disclosed, or represents that its use would not infringe privately owned rights.

1 I

WASH-1 222 UC-80





[----N0 1 I C E


This 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 m y of their contractors, subcontractors. or their employees, makes any warranty, express or implied, or assumes any y, legal liability or responsibility for the a c c ~ ~ a c completeness or usefulness of m y information, apparatus product or process disclosed, or represents that its US; would not infringe privatelv owned richts


F Od~ e by the Superintendent of Documents, U.8 0 veroment Prlntfng Ofaee Washfngton, D.C., 20402




This report was prepared as input to the Office of Science and Technology's Energy Research and Development Study conducted through the Federal Council for Science and Technology.


The contents

represent the views of the panel members and not necessarily those of the Office of Science and Technology.




. 0 0 0 0 . . 0 0 . . 0 o . o 0 0 0 . . 0 0 0 0 0 0 o 0 0 0 0 0 0 0 o o o 0 o o * o o o 0 o o o o o

........ ............................................. 111. RESOURCE UTILIZATI .......................... I V o HISTORICAL DEVELOPMENT OF SALT REACTORS ................ V. SALT CONCEPT ............... VI . STATUS OF MSBR TECHNOLOW ..................................... . MSRE - The Reference Point €or Current Technolwv ........ . Continuous Fuel Processing - The t o Breeding ......... 1. pment ....................... . Fuel P u t a 1 Materials ............... . Molten S a l t R ...... 1. Fuel and Coolant S a l t s ............................. 2 . Reactor Fuel Containment Materials ................. 3. Graphite ........................................... 4 . Other S t r u c t u r a l Materials ......................... . T r i t i u m . Problem of Control ........................... II. SUMMARY















Reactor Equipment and Sktems Development



......................................... . ............................................ F . Maintenance . D i f f i c u l t Problem f o r t h e HSBR ........... f o r the .................. C . Safetv .Different H . Codes. Standard High Temperature DesiRn Methods .... 1


Components Svstems A






. ..

............. ................................. ............. .................................. A ............. ..................................

1 8

10 12 14 19 19 22 22

27 30 33 34 35 38 38 41

43 45 47 48 51



c .


25 26




A- 1



Page Tables I

I1 111


Selected Conceptual D e s i g n D a t a for a Large MSBR.


Important Dates and Statistics for the MSRE .

Camparison of Selected Parameters for the MSRE and 1000 W ( e ) MSBB


Flowsheet for Processing a Single-Fluid MSBR.


-. .



.. ,

, .

. . ..



. .



The Division of Reactor Development and Technology, USAEC,was assigned -..~.. the responsibility of~assessing the status of the technology of the. -( Molten Salt Breeder Reactor (MSBR)& part of the Federal Council of _ . : Science and Technology Research and Development Coals Study. In. conducting this review, the attractive

features and _ areas ~. problem associated wi-th the concept have been examined; but more impor-tantly, _ the assessment has -. __I-- provide a view of the Atechnologp and engineering development efforts and the associated government and i -;-



commitmentswhich tiould be required to dkvelop the MSBR ._. safe, reliable and economic power source for central station


applicatiou. The MSBR 'concept, currently '.


study at the Oak Ridge National

Laboratory (ORNL), is based on use of~'a.circulati&


reactor coupled with on-line continuous‘ fuel processing.

f&l As presently

envisioned, it would operate as a thermal spectrum reactor system ~utiliti& the

a thorium-u&i&




fuel cycle. u&izat$o+of.the .

Thus, the concept kuld nation's



~resources through operation of a b.reeder system employing &&her _ _~ _




sustaining, competitive i n d u s t r i a l c a p a b i l i t y â‚Źor producing economical

power i n a r e l i a b l e and s a f e manner.

A b a s i c ' p a r t of achievement of

t h i s objective is t o gain public acceptance of a new form of power production.

Success i n such an endeavor is required t o permit the


u t i l i t i e s and others t o consider t h e concept as a v i a b l e option f o r

generating electrical power i n t h e f u t u r e and t o consider making the heavy, long-term commitments of resources i n funds, f a c i l i t i e s and personnel needed t o provide the t r a n s i t i o n from t h e e a r l y experimental f a c i l i t i e s and demonstration plants t o f u l l scale commercial r e a c t o r power plant systems.

Consistent w i t h t h e policy established for a l l power r e a c t o r development programs, the NSBR would r e q u i r e t h e successful accompl$shment of three b a s i c research and development phases:

. An i n i t i a l research and development phase in which t h e b a s i c technical aspects of the MSBR concept are confirmed, involving exploratory development, laboratory experiment, and conceptual

i n which t h e engineering and man c a p a b i l i t i e s are developed.

This includes t h e conduct of

in-depth engineering and prooftesting of first-of-a-kind components, equipment and systems.

These would then b e

incorporated i n t o experimental i n s t a l l a t i o n s and supporting






to assure adequate xtnderst&ding

and performance experience


environmental efforts


As these research

the technohgical


and decision

is sufficiently

. A.third

~phase in.which





When the in the





make large scale commitments

by developing


in a safe,



developed and confidence

the utilities

to electric-generating

power plants


reached that would permit



manage the design,

and development

the next stage would be the construc-

-system was attained, .tion of large

economic and



development to proceed with technology

as to gain overall

with major operational,


to be resolved

as well

of design



the capability

and operation





and environmentally

f <


the Light

Water Reactor- (LWR), the high

Temperature Gas-cooled Reactor ~(HTGR) and the Liquid Metal-cooled Fast . Breeder Reac.tor (L?JFBR) has been gbined over the.past two~decades pertaining nuclear

to the 'efforts reactors

This experience


to the point

are~required of public


and commercial

and advances ~-. acceptance.

has clearly demonstrated that the phases of develop_ m&-and d&nonstration'should be simiiar'regardless of the-energy _.. _-. ~ -~~ ~_~_ _ .~ . through each of concept being exploredi that the logicai.progression a.


t.e phases is essential; and that completing t h e work through t h e t h r e e phases is an extremely d i f f i c u l t , t i m e cunsuminp: and c o s t l y undertaking, requiring the h i g h e s t level of t e c h n i c a l management, professional competence and organizational s k i l l s .

This has again

been demonstrated by t h e recent experience i n t h e expanding LWR design, corm t r u c t i o n and l i c e n s i n g activities which emphasize clearly t h e need f o r even s t r o n g e r technology and engineering e f f o r t s than .

were i n i t i a l l y provided, although i h e s e k r e

cases for t h e * f i r s t .experiments and demonstration p l a n t s .


program, which is r e l a t i v e l y w e l l advanced i n its development, tracks c l o s e l y this LWR experience and has f u r t h e r reinforced t h i s need a. i t a p p l i e s t o t h e technology, development and engineerinff a p p l i c a t i o n areas


It should also be kept in mind t h a t t h e l a r g e backlog of commitments

and t h e shortage of q u a l i f i e d engineering and t e c h n i c a l management personnel and proof test facilities 'in t h e government , i n -2ndusttv and

i n t h e u t i l i t i e s make i t even more necessary t h a t a l l t h e r e a c t o r systems be thoroughly designed and t e s t e d b e f o r e a d d i t i o n a l s i g n i f i c a n t commitment t o , and construction o f , commercial power p l a n t s are


initiated. ~

With regard t o t h e XSBR, p r e l f m h a r y r e a c t o r designs w e r e WASH-1097 ("The Use of T h o r i d I n Nuclear Pawer Reactors") based upon .., ~



the information -supplied

by &IL.



Two reactor design concepts-were . :. 1..- --considered -Y a two,_fluid and fertile .=. _-: reactor In which the __fissile i. _ ~.~ _. salts were. separated r-byL .-graphite and a single fluid concept in which ~-_<.' : i. the fissile and fertile salts 'were i completely mixed. This evaluation = problem-areas

resolution .i _.

through -conduct :of an

intensive research and developm-ent program. since the publication I .I 1: -. = .__WASH-1097, all efforts related to the two fluid system have-been discontinued


because of mechanical

of processes which would,


design problems and the development * *: ., j developed into engineering systems,


the,on-line reprocessing of fuel from single ,f_luid reactors. -.. .~ _ At present, the.*BR concept essentially ,; _ isI .-=-_ ._ _ -._-in_ the initial -- research 7~'~. and development phase, with emphasis ,on the development of basic MSBR Z~ technology. The technology program is centered at ORNLwherefair. 7. _.


essentially all research and development on molten salt reactors has --~ .z 1 .-1 ~. _ _ been performed to date. The program is currently funded at a level _', _ .; of $5 million per year. ~Expenditures to date on molten salt reactor -_ .- -_.~ ; : _ . technology both for military and civilian have -. ~, L power applications amounted to approximately $150 million of which approximately $70 million _~. -: _i _. ;.-e has been in support of central station ptier plants. These efforts date . -. -~.I. _~.: .~. ..; .-- 7.-.- .~ _ .in I ~. ~~ back to the 1940's. _ ;


_. -.~~

pi .;~~~_

In considering the MSBR for central station power it .- plant application, ...I~:. ~:- = ~~. pi : 7~- :~~ ,- is noted that this concept has several unique and features: â&#x20AC;&#x2DC;. desirable j --~ _* -at-the -same time, it is characterized b-y both complex technological and : _ ~I. - * ,rI .- ; .~ in-__


\J' p r a c t i c a l engineering problems which are s p e c i f i c t r e a c t o r s and for which s o l u t i o n s have not been developed.


this concept introduced major concerns t h a t are d i f f e r e n t in kind and

magnitude from those cammanly associated with s o l i d f u e l breeder reactors.

The development of s a t i s f a c t o r y experimental units and

f u r t h e r consideration of t h i s concept f o r use as a commercial power plant w i l l require r e s o l u t i o n of these as w e l l


o t h e r problems which

are common t o a l l reactor concepts. c

As p a r t of t h e AEC's Systems Analysis Task Force (AEC report A

WASH-1098) and the "Cost-Benefit Analysis of t h e U.S.


Reactor Program" (AEC r e p o r t s WASH-1126 and WASH-1184).


w e r e conducted on t h e cost and b e n e f i t of developing another breeder system, "parallel" t o t h e LMFBR.

The c o n s i s t e n t conclu-

s i o n reached i n these s t u d i e s is t h a t s u f f i c i e n t information is a v a i l a b l e t o i n d i c a t e t h a t t h e projected b e n e f i t s from the LMFBR program can support a p a r a l l e l breeder program.

Hawever, these

results are highly s e n s i t i v e t o t h e assumptions on p l a n t c a p i t a l c o s t s with t h e recognition, even among concepts i n which ample experience e x i s t s , t h a t c a p i t a l costs and e s p e c i a l l y small estimated differences i n costs are highly speculative f o r p l a n t s t o he b u i l t 15 or 20 years from now. analyses based upon such c

is ques t ionab l e ould c o n s t i t u t e a major b a s i s f o r


making decisions r e l a t i v e t o t h e d e s i r a b i l i t y of a p a r a l l e l breeder effort.



-_ Lid


development programs In this country and abroad

in reactor

has demonstrated -that different costs of introducing


in evaluating the pro.jected

a reactor development-program and 'carrying it forward

to the point of large scale commercial utilization, - different

would arrive at.

estimates of the methods,- scope of development and engineering


and the costs and time required toâ&#x20AC;&#x2122;bring

of successful large scale application&d






Based uponthe ARC's experience with other complex reactor development programs, it

is estimated that a total

about 2 billion


government 1nvestmektu~â&#x20AC;&#x2DC;to

in undiscounted direct

costs* could be required

to bring the molten salt breeder or any parallel

breeder to fruitfon


A magnitude of funding up to .&Is reactor. ._ it_ I level could be needed .to estab&h'the necessa&' technology and, ~._. .. . engineering bases ; obtafn the required industrial capability; and _ advance through a series of test facilities, reactor experiments, and _ : _-~, demonstratfon plants to.a commercial MSBRsafe and suitable to aeme

a viable;



, as a major energy option for central





.-~ ~,.~ w





power generation in the

_I -..



i ,-

'WASH-ilEi4- -.Updated (1970) Cost-Benefit Analysis of the U.S. Breeder Reactor Program, January 3972W





The MSBR concept is a thermal spectrum, f l u i d f u e l r e a c t o r which operates on t h e thorium-uranium

f u e l c y c l e and when coupled with

on-line f u e l processing has t h e potential f o r breeding a t a meaningful level.

The marked differences i n t h e concept as compared. I


t o s o l i d fueled r e a c t o r s , make t h e MSBR a d i s t i n c t i v e a l t e r n a t e . Although t h e concept has a t t r a c t i v e f e a t u r e s , t h e r e are a number of d i f f i c u l t development problems t h a t must b e resolved: many of t h e s e

are unique t o t h e MSBR w h i l e - o t h e r s are p e r t i n e n t t o any complex r e a c t o r ,spstem

The t e c h n i c a l e f f o r t accomplished s i n c e t h e p u b l i c a t i o n of WASH-1097 and WASH-1098 has i d e n t i f i e d and f u r t h e r defined t h e problem areas; however, t h i s work has not advanced t h e program beyond t h e i n i t i a l Although progress has been made

phase of research and development. in several areas -(e.g.,

reprocessing and improved g r a p h i t e ) , new

problems not addressed in WASH-1097 have a r i s e n which could a f f e c t t h e p r a c t i c a l i t y of designing and operating a MSBR.

Examples aâ&#x201A;Ź

major u n c e r t a i n t i e s relate t o materials of construction, methods f o r c o n t r o l of tritium, and t h e design of components and systems along with t h e i r spec;ial handling, inspection and maintenance equipment. Considerable research and development e f f o r t s are required i n order t o obtain t h e d a t a necessary t o resolve t h e u n c e r t a i n t i e s .




Assuming t h a t p r a c t i c a l s o l u t i

o these problems can be found, a

f u r t h e r assessment would have t o be made as t o t h e a d v i s a b i l i t y of proceeding t o t h e next s t a g e of t h e development program. 'In advancing t o t h e next phase, i t would be necessary t o develop a g r e a t l y expanded i n d u s t r i a l and u t i l i t y p a r t i c i p a t i o n and commitment along with


s u b s t a n t i a l increase in'govermnent support.

Such broddened involve-

ment would require an evaluation of t h e %BR

i n terns of already

e x i s t i n g commitments t o o t h e r nuclear power and'high p r i o r i t p energv development e f f o r t s .



It has long been recognized t h a t t h e importance of nuclear f u e l s f o r

power production depends i n i t i a l l y on t h e u t i l i z a t i o n of t h e n a t u r a l l y occurring fissile U-235; b u t it is t h e more qbundant f e r t i l e materials,

U-238 and Th-232, which w i l l b e t h e major source of nuclear puwer generated i n t h e future.

The b a s i c physics c h a r a c t e r i s t i c s of f i s s i l e

plutonium produced from U-238 o f f e r t h e p o t e n t i a l f o r high breeding gaiqs i n f a s t reactors, and t h e p o t e n t i a l t o expand g r e a t l y t h e u t i l i z a t i o n of uranium resources by &king f e a s i b l e t h e u t i l i z a t i o n of a d d i t i o n a l vast q u a n t i t i e s of otherwise uneconomic low grade ore.


a s i m i l a r manner, t h e b a s i c - p h y s i c s c h a r a c t e r i s t i c s of t h e thorium cycle w i l l permit f u l l u t i l i z a t i o n of t h e nation's thorium resources

while a t t h e same time o f f e r i n g t h e p o t e n t i a l f o r breeding i n thermal reactors


The estimated thorium resemes are s u f f i c i e n t t o supply t h e world's

electric energy needs f o r many hundreds of y e a r s i f the thorium is used in a high gain breeder reactor.

It ie projected t h a t if t h i s

q u a n t i t y of thorium were used in a breeder r e a c t o r , approximately

1000 Q (1 Q = 10l8 Btu) would be r e a l i z e d from t h i s f e r t i l e material. It is estimated t h a t t h e uranium reserves would a l s o supply 1000 Q*

of energy i f t h e uranium were used i n LMFBRs.

I n c o n t r a s t , only 20 Q

*Uranium recoverable a t U308 p r i c e up t o $lOO/lb.

-11would b e available i f thor

e f e r t i l e material in

an advanced converter reactor because the reactor would be dependent upon U-235 availability for f i s s

inventory make-up

(Note :

conservative estimate is that between 20 and 30 Q w i l l be

electric power generatio

n w and the year 2100.) .-






The i n v e s t i g a t i o n of molten s a l t r e a c t o r s began in t h e late 1940's as p a r t of t h e U.S. A i r c r a f t Nuclear Propulsiod (ANP) Program,


t h e A i r c r a f t Reactor Experiment (ARE) was b u i l t a t Oak Ridge and i n 1954 i t w a s operated successfully f o r nine davs a t power level@up t o

2.5 MW(th)

and f u e l o u t l e t temperatures up t o 1580.F.

mfxtute of NaF, ZrP4, and UF4.

The ARE f u e l w a s a

The moderator w a s BeO.and t h e pipinn and

vessel were constructed of Inconel.

I n 1956, ORNL began t o study molten s a l t r e a c t o r s f o r a p p l i c a t i o n as c e n t r a l s t a t i o n converters and breeders.

These s t u d i e s concluded thatl

g r a p h i t e moderated, thennal spectrum r e a c t o r s operating on a thoriumuranium cycle were most attractive f o r economic power production,


on t h e technology a t t h a t time, i t was thought t h a t a two-fluid r e a c t o r in which t h e f e r t i l e and f i s s i l e salts w e r e kept s e p a r a t e was required i n order t o have a breeder system.

The s i n g l e f l u i d r e a c t o r , while not

a breeder, appeared simpler i n design and a l s o seemed t o have t h e

p o t e n t i a l f o r l o w power costs.

Over t h e next fw years, ORNL continued t o studp both t h e two f l u i d and s i n g l e f l u i d concepts, and i n 1960 the' design of t h e s i n g l e f l u i d ' 8 MW(th) Holten S a l t Reactor Experiment (MSRE) was begun.

The MSRE was

completed i n -1965 and operated s u c c e s s f u l l y during t h e period 1965 t o 1969.

The %RE

experience is t r e a t e d in more d e t a i l i n a later s e c t i o n .

Concurrent wlth.the


and development on means for

of the MSRE;OR&performed research . proceseinp; molten salt fuels. In-1967

new discoveries were madewhich suggested that a single fluid


could be combinedwith continuous on-line fuel .. processing; to becomea I. breeder sys tern. Because of the mechkcal design prob1en-mof the two fluid


and the laboratory-&ale

would permit20n~lke


development of

it was deierdninkd that 8 shift

emphasis to the single fluih breeder concepk e_._.; system fs being studied at the present. -2




should be make;






The breeding reactions of the thorium cycle are:


2 3 2 ~ n-233~


B min.


233Pâ&#x201A;Ź4 27.4d e-


Because of t h e number of neutrons produced per neutron absorbed and t h e mall f a s t f i s s i o n bonus associated with U-233 and Th-232 i n the

thermal spectrum, a breeding r a t i o only s l i g h t l y g r e a t e r than unity is achievable.

I n order t o realize breeding with t h e thorium cvcle i t is

necessary t o remove the bred Pa-233 and t h e various nuclear poisons produced by t h e f i s s i o n process from t h e high f l u x region as quickly a s possible.

The Molten S a l t Breeder Reactor concept permits rapid removal

of Pa-233 and the nuclear elements).



Xe-135 and the rare e a r t h

The r e a c t o r is a f l u i d fueled system containing uF4 and

ThF4 dissolved i n LiF

- BeF2.

The molten f u e l s a l t flows through a

graphite moderator where t h e nuclear reactions take place.

A side

stream is continuously processed t o remove t h e Pa and rare e a r t h elements, thereby permitting the achievement of a calculated breeding r a t i o of about 1.06.

The MSBR is attractive because of the following:


U s e of a f l u i d f u e l and on-site processing would eliminate t h e

problems of s o l i d f u e l f a b r i c a t i o n and the handling, and


-15shipping and reprocessing of spent fuel elements which are


associated with all other reactor types under active consideration. 2.



MSERoperation on the thorium-uranium fuel cycle would help conserve uranium and thorium resources by utilizing reserves with high efficiency.


thorium; ~, ,

I j


The MSBRis,, projected to have attractive The majorsuncert,ainty in the fuel

fuel cycle costs.

cycle cost is associated

with the continuous fuel processing plant has not been I. developed. .. .4.

The safety issues ass&a&d different

.from those of &id

I with the MSBRare genetallv -fuel reactors.

Thus, there

might be safety advantages for the KSBRwhen considering major accidents.

An accurate- assessment of MSBRsafety is

not possible today because of the early state of development. 5.

Like other advanced reactor systems such as the LMFBRand - HTGR, the MSBRwould employ modem'steam technologv for power generation with high thermal efficiencies.

This would reduce

the amount of waste heat to be discharged to the environment.



Selected conceptual design data for a large MSBR, based primarily on design s t u d i e s performed a t ORNL, are given i n Table 1.

There are, however, problem areas associated with t h e MSBR which wst be overcome before the p o t e n t i a l of t h e concept could be a t t a i k e d * These include development of continuous f u e l processing, reactor and processing s t r u c t u r a l materials, tritium c o n t r o l methods, reactor equipment and systems, maintenance techniques, s a f e t y technolonv, and

MSBR codes and standards.

Each of these problem areas w i l l now be

evaluated i n some d e t a i l , which w a s demonstrated by t h e Molten S a l t Reactot Experiment (â&#x20AC;&#x153;2) during its design, construction and operation a t Oak Ridge and the conceptual design parameters presented I n Table I and i n Appendix A. A conceptual flawsheet for t h i s system is shown i n FiRure 1.



Table I . Selected Conceptual Design Data for a Large KSBR

Power, W(e)


1 Pawer, W(th)

2240 3500 psia, 100O0F,

Steam System


44%net efficiencv 72% 7LIF, 16%BeF2,


lng and V e s s e l %dera tor

Sealed Unclad Craphlke

Breeding Ratio Specific F i s s i l e Fuel

ded Doubling T h e , Y e a - -

Core Temperatures,


1050 Inlet, 1300 outlet





VI. A.

MSRE- The Reference Point

for Current,Technology .~. I~ Reactor Experiment (MSRE) was begun in .1960 at

The Zlolten Salt

ORNL as part of the Civilian of the experiment molten salt

Nuclear Power Program.

was to demonstrate


were achieved-during





in the world

the more signifikt are givenin

In spite salt



December 1969.. These included

the basic



of the experiment


the distinction

June 1965 to

of becoming the

on U-233. Someof

to operate".solely

dates and statistics

Th? purpose


tom the MSRE

Table II.

of the MSRE, there are -many areas

of thesuccess


of molten

which must be expanded and developed in order to

proceed fram this



and economic 1000 W(e)


to a safe,


MSBR with-a

30-year life. To illustrate ,. -. this point, some of the most important differences in basic design . and performance characteristics between the MSRE and a conceptual .

lOOO.hW(e) MSBR are given inTable be accomplished size. scale-up


problems associated


of increasing

provides an appreciation of the

in ,going from the MSRE to a large MSBR. Some with


high performance


Scale-up would logically

through development of reactor

Examination of Table III



from a smaJ1 experiment to a

power plant

are not adequately



Table 11.

Dates: i


. . . . . . . . . . . . . . J u l y 1960 Critical with 235U Fuel . . . . . . . . . June 1, 1965 operation a t f u l l power - 8 MtJ(th) . . May 23, 1966 Complete 6-month run . . . . . . . . . . . . March 20, 1968 End Operation with 235U f u e l . . . . . . . March 26, 1968 Critical with 233U f u e l . . . . . . . . . October 2, 1968 Operation f u l l power with 233U f u e l ... . January 28, 1969 Reactor operation terminated . . . . . . . . December 12, 1969

Design i n i t i a t e d



Statistics: Hours c r i t i c a l

. . . . . . . . . . . . . 17,655

Fuel loop t i m e c i r c u l a t i n g s a l t (Hrs).

Equiv. f u l l power hours with 235U f u e l Equiv. f u l l power hours with 233U





21,788 9,005 4,167



Table III. Comparison of Selected Parameter8 of the MSREand 1000 MN(e) MSBR-11 > MSREMSBR Y v ; General ThermalPcwer,MW(th) 2250 ' s Electric Pwer, MN(e) 0 1000 Plant lifetime, year8 /: .-, ~ 4- ~_ ~_ ~30 Fuel Proceasing Scheme Off-line, batch -- On-line, continuous : -7 I .: processing:; -I processing ... Breeding Ratio.- ~:_ k88 thSUl 1.0 1,06IT ~(No Th present) Reactor Fuel Salt Moderator

7 7 .-. _ =LiF-BeFZTZrF4-UF4 Lif-BeF2TThF4-UP4 Unclad, '. ~-Unclad, unsealed 'graphite- sealed graphite Reactor Vessel Mater;al Standard Hastelloy-N Modified Hastelloy-N Pwer DeAity,~RW/iiter 2.7 -- 22 - ;=c, ; Exit T&perature, l F -~--~.'- ~. 1210 1300Temperature Rise ACrO88 Core.1OF 40~.. _ ~m.250 Reactor Vessel Height, Ft. 0 -20 Reactor Vessel Diameter, Ft. 5 22 Vessel Design Pressure,-psia'~. 65 ~--' - 75.-. 'I* ,8.3-x lO14 A Neutrons/m‘+sec

Other Componentsand Systems Data,I-=_ . _; = Number of Primary Jm. ' ~' _.:.. -,: Circuits .I. - * ".y1200--. r. I L,. ~Fuel Salt PumpFl&; j$m Fuel






I’ ~48;s

~. _

7LIF-BeF2 Secondary Coolant Salt Numbersof Secondary ~Circufts :-l _ ~: ,._ 1 Secondary-Salt PumpFlaw, Epm -850 -_ Secondary~Salt PumpHead, ft. Seqzmdary~Salt 78 ~0 Nuniber~of Steam Gener&ors Number~of Steam Generator Capacity, MW(th) 0

-*-_ 4 .-16,0& ~. 150. ~~~:

: -


NaF-NaBF4 4. _~ - :~ 20,000 300 i6 . 121


A/ Based on information from "Conceptual Design Study of a Single Fluid Molten Salt Breeder Reactor," ORNL-4541,June 1971.


represented by t h e comparison presented i n t h e Table. -Therefore, it is u s e f u l t o examine a d d i t i o n a l f a c e t s of MSBR technology i n r

more d e t a i l .


Continuous Fuel Processing

- The Key t o Breeding

I n order t o achieve nuclear breeding i n t h e s i n g l e f l u i d NSBR it is necessary t o have an on-line continuous f u e l processing system.

This would accomplish t h e following:


I s o l a t e protactinium-233 from t h e r e a c t o r environment so it can decay i n t o t h e f i s s i l e f u e l i s o t o p e uranium-233 before being transmuted i n t o o t h e r isotopes by neutron f r r a d i a t i o n .


Remove undesirable neutron poisons from t h e f u e l s a l t and thus improve t h e neutron economy and breeding~performance

of t h e system.


Control t h e f u e l chemistry and remwe excess uranium-233 which is t o b e exported from t h e breeder system.

1. Chemical Process Development The Oak Ridge National Laboratory has proposed a f u e l processing scheme t o accomplish breeding i n t h e MSRR, and t h e flowsheet processes involve:

a. Fluorination of t h e f u e l s a l t t o remove uranium as UF6.



b. Reductive extraction salt-with

of protactinium

a mixture of lithium

by contacting the

and bismuth.

c. Metal transfer processing to preferentially remove the . rare earth~fission product poisons which would-otherwise hinder breeding performance. The fuel proc"eseing system shown in Fig. 2 is in an earlv stage . of development at present and this tvpe of system has not been demonstrated on an. operating reactor. required only off-line, from fuel salt..

By comparison, the MSRE

batch fluorination

to recover uranium


At this time, the basic chemistry involved in the MSBR processing schemehas been demonstrated in laboratory scale _i experiments. Current efforts at O&k Ridge are being-directed toward development of subsystems incorporating many of the _required processing steps. Ultimately a complete breeder processing~ experiment would be requfred to demonstrate the system with all~the chemical conditions and operational requirements which would be encountered with any MSBR. Not shown on the flowsheet is a separate which would require injecting

~~ .~~

processing svstem

helium bubbles into the fuel

_ .-







I '





a I

E t 4 2




0 J





a ul













a 0






v) v)


a a W LL

U v).

z U


? J U

e I W




e a

a 2

0 c V a x

a c








0 W



0 e a

P 0 3 J LL






~. they -collect

in the reactor system until

thgn to~~ciryzulate


xenon, and then removing the

bubbles and xenon from the reactor system.. a-highly undesirable neutron poison which will - ante by capturing neutr& ~. -fuel.

hamper breeding -perf&n-

which -would otherwise breed new

This concept for xenon stripping

was -demonstrated~ln, althouih more efficient-and controllable : stripping syst+ns[ willmbe-desirable ior <the MSBR. The kenon polsonlng in the MSRE_was ~reduceh by a. factor of six by xenon stripping;

the goal for the MSBRls a factor of ten reduction. _ :

2. Fuel Proceselng~Structural Aside fromthe






development requirements assoclated.wlth for the fuel.processlng presents dif flcult

containment Fterlals

systems. -19 ~artlcular~~llquid

concentrated on using moly)denum

and graphite for contalnlng,blsmuth. -molybdenumand-graphite -are;-difficult



to use for +uch~englneerlng

T)us, it w&Xl-be necessary to~d&@op.improyed

applications. : ~techniques-for .

fabrication-a@ joining-before ~~ ~~ .posslble ln.the'reproi~filnlF; system._ -_ -'.-


compatlbl&lty p?oblems wlth most structural

metals., and present effortsTare


there are also


.-. .I -i.







A second material8 problem of t h e current fuel processing system

ia the contafnment f o r the f l u o r i n a t i o n s t e p i n which uranium is v o l a t i l i z e d from t h e f u e l salt.'

The f l u o r i n e and f l u o r i d e s a l t

mixture f9 corrosive t o most s t r u c t u r a l materials, including graphite, and present ORNL flowsheets show a "frozen w a l l "

f l u o r i n a t o r which operates with a p r o t e c t i v e l a y e r of frozen f u e l s a l t covering a Aastelloy-N vessel w a l l .

This component

would r e q u i r e considerable engineering development before i t is t r u l y practical f o r use i n on-line. f u l l proceasing systems.


Molten S a l t Reactor Design

- Materials Requirement8

I n concept, t h e molten s a l t reactor core is a comparatively uncomplicated type of heat source.


The MSRE reactor core, f o r

example, consisted of a p r h m a t i c s t r u c t u r e of unclad graphite moderator through which f u e l s a l t flowed t o be heated by the self-sustaining chain r e a c t i o n which took place as long as t h e

salt was in t h e graphite.

The e n t i r e reactor i n t e r n a l s and f u e l

s a l t were contained i n vessels and piping made bf Haetelloy-N,


high-strength n i c k e l base a l l o y which was developed under t h e A i r c r a f t Nuclear Propqlsion Program.

Over t h e four-year lifetime

of t h e E R E , t h e reactor ' s t r u c t u r a l b t e r i a l s performed satisf a c t o r i r p for the purposes of the experiments although operation of t h e MSRE revealed possible problems w i t h long term use of

t -27-


Hastellay-N in contact with f u e l s a l t s containing f i s s i o n



The MSBR a p p l i c a t i o n is mare demanding i n many respect8 than the MSRE, and a d d i t i o n a l d

ork would b e 'required in

several areas of mater

om b e f o r e

s u i t a b l e materials


could become available.


The E R E f u e l s a l t was a mixture of 7LiF-BeF-ZrP4-UF4 .0-29.1-5.0-0.9

proportions of



mole X , respectively. 7

Zirconium f l u o r i d e

included ae p r o t e c t i o n against U02

p r e c i p i t a t i o n should inadvertent oxide contamination of t h e s

epatem occur.

MSRE operation i n d i c a t e d that c o n t r o l of

oxides was not a major problem and thus it is not considered necessary t o include zirconium i n f u t u r e molten s a l t r e a c t o r fuels.

It should a l s o be noted t h a t t h e MSRE f u e l contained whereas the proposed MSBR f u e l s would include f o r breedfnk.

With the

p o s s i b l e except l s f a c t o r i l y throughout the

r r e n t l y proposed by ORNL, would be a i n proportions of 71.7-16-12-0.3


Lid mole X , respectively.


This s a l t has a melting point of about

and a vapor pressure of less then 0.1 mm Hg

operating temperature of llSO'F.

a t t h e mean

It a l s o has about 3.3 t i m e s

the density and 10 tbes t h e v i s c o s i t y of water.

Its thermal

conductivity and volumetric h e a t capacity are comparable t o


The high melting temperature is an obvious l i m i t a t i o n f o r a 8ystem using thfs s a l t , and t h e MSBR is limited t o high temperature operation.

fa a d d i t i o n , t h e lithium component

must be enriched in Li-7 i n order t o allow nuclear breeding, s i n c e n a t u r a l l y occurring lithium contains about 7.5% Li-6


L i d is undesirable i n t h e MSBR because of its tendency t o .

capture neutrons, thus penalizing breeding performance.

The chemical and phyeical c h a r a c t e r i s t i c s of t h e proposed

MSBR f u e l mixture have been and are being i n v e s t i g a t e d , and they are reasonably w e l l known for w i r r a d i a t e d salts.


major unknowns are associated with t h e r e a c t o r f u e l a f t e r i t

has been i r r a d i a t e d .

For example, not enough is-known about

t h e behavior of f i s s i o n products.

The a b i l 3 t y t o p r e d i c t

f i s s i o n product behavior i e important t o p l a n t s a f e t y , operation, and maintenance.

While t h e MSRE provided much.

u s e f u l iafonaation, t h e r e is s t i l l a nee

-29$artGzularly -.-


with regard toi‘the- fate of the- so-called "noble

metal" fission

prqducts’such aS mol~bd&ium, niobiti

others which are cenerated-%n substantial ~_

&ose behaVioriti


quix%i~ies and

tkisystem is not well ‘unders toad. _I -.- _ - .--‘I..

.-- A more &%plete'underst~anding of the uhy&cal/chenical --~-characteristics

of the-irradiated

fuel salt is also needed.

As an illustrationof

this point, anomalous powef pulses were ~. observed- d&&g early‘operation oi the ?¶SRE with U-233 fuel \ : ' _*. ._ : &hich were attributed to unusual behavior of helium pas


..= ' bubbles as they circulated _,.-


through the reactor.




behavior is bkeved

to have been due to some physical and/or _chemical characteristics of the fuel salt which were never





Vork on molten fuel salt

product chemistry is currently _ -~.~~~

under way.


the behaviorSVofthe f&l salt-would need to be confirmed in an .I "~7 operating reactor.

.-The cools& salt_ -in the secondary system of the WRE was of ~. .: .. .. molar composition 66%'L$F-34XBeF2. While this coolant / performed satisfactorily (noldetectable.corrosidn or~reactfon ,-could-be observed In-the in-the secondary system), the salt-has a _

_ high melting..temperature=(8500F) melting. -temperature =(850?F) and la relatively Thtis,



may not be-the appropriate choice- for power reactors


for two reasons:

(1) l a r g e r volumes of coolant s a l t w i l l be

used t o generate steam i n the MSBR, and (2) s a l t temperatures i n t h e steam generator should b e l a w enough, i f poesible, t o u t i l i z e conventional steam system technology with feedwater temperatures up to about 550'F.

The operation of MSRE was

less a f f e c t e d by the coolant s a l t melting temperature s i n c e i t dumped t h e 8 MW(th)

of h e a t via an air-cooled radiator,

The high melting temperatures of p o t e n t i a l coolant s a l t s remain a problem.

The current choice i s a e u t e c t i c mixture

of sodium f l u o r i d e and sodium fluoroborate with a molar composition of 8% NaF-92% NaBF

4; t h i s salt melts

at 725.F.

It is c m p a r a t i v e l v inexpensive and has s a t i s f a c t o r v heat

transfer properties ,

However, t h e e f f e c t s of h e a t exchanger leaka between t h e coolant and f u e l salts, and between t h e coolant s a l t and

steam systems, must b e shown t o b e t o l e r a b l e .


fluoroborate s a l t is c u r r e n t l y being studied with respect t o both its chemistry and compatibility with H a s t e l l o y - N .


Reactor Fuel Containment Materials , A p r e r e q u i s i t e t o success f o r the MSBR would be the clbilitv t o assure reliable and s a f e containment and handling of molten

LJ j



fuel salts at ail: times dur'ink the life ' w&d

be necessary,; therefore;

of the reactor.

"develop suitable


-merit materials -for MSBRapplication


before plants could be . . .

con&ructed;- -. r


.--- I


A serious question concerning the constituents




of Eastelloy-N with

fuel salt was raised by the


operation examlnation:of the-MSRE-in 1971. Although the MSRE materials performed satisfactorily operation,

for that sqstem during its

subsequent examintition of metal which was exposed to

MSREfuel salt revealed that the alloy had experienced intergranular attack

~of about 0.007 inch.

The attack was

not obvious. tit%1 metal specim&s were tensile

tested, at which


time &a&-*opened

up as‘the metal was strained.~

examination-revealed that several fission tellurium,


products, including

had-penetrated the metal to depths comparable to

‘those of the cracks.

At the present'time,

the intergrantilar-attack,was Subsequent laboratory

it is thought that

due to the.presence of tellurium.

tests have verified

produce', under~certainconditlons, .' aast&rop N* . .~ '- *, +

that tellurium

interffranular -, ._




cracking in


.~ Although the limited penetration of cracks presented no problems for the-MSRE, concern &


respect to the chemical

-32compatabilitv of Hastellay-N and MSBR f u e l salts when subjected t o t h e more s t r i n g e n t MSBR requirements of higher power density and 30-year l i f e .

If the observed i n t e r m a n u l a r a t t a c k was

indeed 'due t o f i s s i o n product a t t a c k of t h e Hastelloy-N, then t h i s material may not. be s u i t a b l e f o r e i t h e r t h e piping or the v e s s e l s which would b e exposed t o much higher f i s s i o n product concentrations f o r longer periods of t i m e .

E f f o r t s are under

way to understand and explain t h e cracking problem, and t o

determine whether alternate r e a c t o r containment materials


should b e a c t i v e l y considered.

I n addition t o t h e i n t e r g r a n u l a r corrosion problem, t h e standard H a s t e l l o y - N used i n t h e MSRE is n o t a u i t a b l e f o r use i n t h e MSBR

because its mechanical p r o p e r t i e s d e t e r i o r a t e t o an unacceptable

level when subjected t o t h e higher neutron doses which would The problem

occur i n the higher power density, l o n g e r - l i f e MSBR.

is thought t o b e due mainly t o impurities i n t h e metal which are transmuted t o helium when exposed t o therrnal neutrons.

The helium

is believed t o cause a d e t e r i o r a t i o n of mechanical p r o p e r t i e s by its presence a t grain boundaries wlthin t h e a l l o y .

It would be

necessary t o develop a modified Hastelloy-N with improved i r r a d i a t i o n r e s i s t a n c e f o r t h e MSBR, and some progress is beinff made i n t h a t direction.

It appears a t t h i s t i m e t h a t s m a l l additions af

c e r t a i n elements, such as titanium, improve t h e i r r a d i a t i o n .




of Hastelloy-N substantially.

Development work on

modified. alloys with imprwed,irradiationresietance 1 currently under way.


3. &phi& Additional developmental effort -produce graphites suitable


for VSBR&piicatian.

associated with irradiation results


required to The-first

damageto graphite structures

from fast neutrons.

1s which

Under high neutron doses, -of the

order of 1022neut.rons)~2 , most graphites tend to-become dimen&nally~unstable Based on testsof-


commercially available

and grossswelling graphl&sam$es graphites

of the material occurs. at ORNL, the best.

at this time-may be usable to

about 3 x i0 22 neutrons/cm2 , before the core graphite would have ~l_ . This corresponds to roughly a four-year graphite to. be replaced. -~ . lifetime-for the ORNL~reference design. Wile this might be .-i. acceptable, there are still uncertainties about-the fabrication I L and performance of large graphite pieces, and additional work -, would

be required


a four-year


could be assured


the higher MSBRpower densities~how;a being considered.. In anv : 0 --..-1 event, -there would be an obvious~econamfc;.+ncentive~t~~develop longer lived graphites for MSBR application s$nce:a--four-year .I _ life for graphite is- estitiated to represent a fuel &le cost ' _"'_





penalty of about 0.2 mills/kw-hr r e l a t i v e t o a system w i t h


t h i r t y year g r a p h i t e l i f e .

The second major problem associated with Rraphites f o r MSBR a p p l i c a t i o n is t h e development of a s e a l i n g techniaue which will keep xenon, an undesirable neutron poison, from d i f f u s i n g

i n t o t h e c o r e g r a p h i t e where it can capture neutrons t o the detriment of breeding performance.

While g r a p h i t e s e a l i n g

may not b e necessary t o achieve nuclear breeding i n t h e MSBR, t h e use of sealed g r a p h i t e would certainly enhance breeding performance.

The economic Incentives o r p e n a l t i e s of g r a p h i t e

s e a l i n g cannot be assessed u n t i l a s u i t a b l e s e a l i n g process is developed


Sealing methods which have been investigated t o d a t e i n c l u d e p y r o l y t i c carbon coatiag and carbon impregnation.

Thus f a r ,

hawever, no sealed g r a p h i t e t h a t has been t e s t e d remained s u f f i c i e n t l y impermeable t b gas a t MSBR design i r r a d i a t i o n doses, and research and develo


i n t h i s area 1s continuing.

Other S t r u c t u r a l Materials i

In- addition t o t h e

c t u r a l materials requirements f o r t h e

! L

r e a c t o r and f u e l processing systems proper, components and systems which have s p e c i a l materials reauirementa.

Such components as t h e primary h e a t exchangers and




steam generators~must function"while different

working fluids.

in contact with two


,At the present time, Hastelloy-N is cormldered to be the most promising material

for'use in all salt containment systtie,

including the secondary piping and components. Research to date indiicates that sodium fluoroborate ~_






ible as long as the water content of the _ : fluoroborate low; otherwise, accelerated Corrosion can occur.

is kept


testing would be needed and is underway. ..

Hastelloy-Nil'ha8 not been adequately evaluated for service under a range of steam condition8 and whether material D.


for use in steam generator8 is still

be a suitable not kncm.

Tritium - A Problem of Control

Because of the lithium present

in fluoride

fuel salts,

concept has the inherent problem of generating isotope


of hydrogen.

Due primarily

the present WBR


Tritium is produced by the- foll&ing ti1






to 'these interactians,

of about 2400 curies/day in a 1000 We


a radioactive


%. 'H. would be-produced at a rate




with about

- 30 40 t o 50 curiesldrtv f o r li.rht water, Raa-cooled,

and f a q t hrecrter

r e a c t o r s , i n which t r i t i u m I s produced primarilv as a l o w v i e l d fisston product.

T r i t i u m productinn in heavy w a t e r reactors of cormarable size

i s Renetallv in the r m E e 3509 t o 5800 curies/dav, due t o neutron i n t e r -

a c t i o n s with t h e deuterium nresent i n heavy w a t e r . -


To' f u r t h e r compound tile prnhlem tritim d i f f u s e s r e a d t l v throuzh As a tesri Hactellov-N a t elevate> tempetatuteq. , .

t o prevent tritium C r a m dfffrrstnE throtlRh t h e p i p l n p

t h e VSRQ system (such



heat exchancars) and everrtuallv reachin? t h e

3te.m system where f t he discharr.ed


t h e environwnt


water. -..

The problem of trCtitm c o n t r o l i n t h e T B P fs hef.n$ studied In d e t s i l a t r)!?NL.

The followins ate b e i n r considered as potentfa1 mathoas


trâ&#x201A;Źtium c o n t r o l :


Exchangfnp t h e t r t t i m far an-? hvdronen nresent

f n the


coolant, therebv retaininp t h e t r t t l t i m tn t h e sacnnriarv cnolnnt.


Usinp, coatinEs an metal surfaces f n order t o l n h i b f t t r i t t l t m







OperatIny, the reActor


causing, the form&ion


removed fn the off-Eas


more axid*zfn?,~


therebv ' I'


svstems.~ .





secondant coolant,.




erP;., *odium or helium,








~steam to "Setter"



Of these notentisl

fn either

the heat exchanger or steam


a purge pas between the-walls:


the use of an additional

between the secondary -and method technically. , equipment required



Using duplex tubing generator

loop between the fluorahorate


IS considered

but -it Twould also be extiensive and the loss of thermal


the most effective

due tti the adbitirrnal


;T -.


an economfc viewpoint,-

does not. significantlv -tritium

the mostdesYrable



is one which

-â&#x20AC;&#x2DC;the svstem, :kuch RS exchazipe of

-for.-hydrogen ~present =in--the seconda?+ coolant;-~~Thls

is being investigated


The tritium

problem may be eased bv the low nertneahilftv




.~ _




onâ&#x20AC;&#x2DC; trttium

I m.



._ '~ . -*I



Oxide coatings which occur


steam generator materials i n contact with

steam, and t h i s is a l s o being i n v e s t i g a t e d a t ORNL.


Reactor Equipment and Sfstems Development

While t h e XSBR would u t i l i z e some e x i s t i n g engineering technology from

o t h e r r e a c t o r types, there are s p e c i f i c components and systems f o r which a d d i t i o n a l development work is required.

Such work would have t o take

i n t o account t h e induced a c t i v i t y t h a t those components would accumulate i n t h e MSBR system, i.e.,

also need t o b e developed.

s p e c i a l handling and maintenance equipment wouThe previous discussion has already d e a l t

with a number of these, such as f u e l processing components and systems, but a d d i t i o n a l discussion is appropriate.

1. C0mPonents

As i n d i c a t e d i n the Table 111, a number of components must b e scaled up s u b s t a n t i a l l y from t h e %RE is possible.

s i z e s before a l a r g e MSBR

The development of these l a r g e r components along

with t h e i r s p e c i a l handling and maintenance equipment is probI

ably one of t h e most d i f f i c u l t and c o s t l y phases of HSBR development,

However, reliable, s a f e , and maintainable

components would need t o be developed i n o r d e r for any r e a c t o r system t o b e a success.

The MSBR pumps would l i k e l y be similar i n b a s i c design t o those f o r t h e MSRE, namely, vertical s h a f t , overhung impeller pumps.

-39-Substantial experience has been-gained over the years in the design,-'fabrication

and operation of smaller salt pumps, but

-the size Gould have~to be increased substantially application.


The deve.lopment and proof testingof


such units

~along with their handling and maintainence equipment ,and test facilities

are expected to be costly and time consuming.

The intermediate heat exchangers for the MSBRmust perform with a minimumof salt inventory in order to improve the breeding performance by lowering the fuel inventory-.

Special surfaces to

enhance heat transfer would help achieve this,

and more studies

would be in order. ~Basedsn previous experience with other systems6 it

is believed-that


cult development and proof testing effort;

The steam generator for MSBRapplications difficult

would require a diffiI

is probably the most

large component to develop since it

represents an

item for Which there has'been experience to.'date., .is believed that a difficult

m--As discussed -previously,~the.~hij3h'melting candidate secondary koola&si


develo$ment~and~proof testing pro-

' -gram would'beneeded to provide reliable


reactor '

and maintainable units. temperatures- of

such ;as'sodium fluoroborate,

present problems -of matching with conventional iteain ~technology~ -At this timeimcentral



so&r plants utilize j, )


feedwater temperatures only up t o about 550째F.


coupling a conventional feedwater s y e t e m t o a secondary coolant which freezes at 72S'F design and control.

presents obvious problems i n

It might be necessary t o provide modifi-

cations t o conventional steam system designs t o help resolve t h e problenk.

Because of these factors, a etudy r e l a t e d t o

t h e design of steam generators has been i n i t i a t e d a t Foster Wheeler Corporation.

Control rods and drlves f o r t h e MSBR would also need t o be developed.

The MSRE c o n t r o l rods w e r e air cooled-and operated

i n s i d e Hastellop-N thimbles which protruded dawn i n t o the f u e l


The MSBR would require more e f f i c i e n t cooling due t o the

higherepower densities involved.

Presumably rods and drives

would be needed which p e m i t t h e rods t o contact'and be cooled by the f u e l salt.

The s a l t valves f o r l a r g e MSBR's

r e p r e s e n t another development

problem, although the f r e e z e valve concept: which was employed successfully i n t h e MSRE could l i k e l y be scaled up i n s i z e and u t i l i z e d f o r m y MSBR applications.

Mechanical t h r o t t l i n g

vdlvek would a l s o be needed f o r t h e MSBR s a l t ~ y s t e m s , even though no t h r o t t l i n g valve w a s used with t h e MSRE.


shutoff valves f o r salt systems, i f required, would have t o be developed



Other componentswhichwould devklcjpment and testing

.gas strippers

require considerable engineering

include-the helium bubble generators and

which are- +roposed.for

i product xenon from&the ~fuel salt. this area is cukently

use in removing the fission

1Research and development in

way as part of the technology


." program at ORNLm " 2.






Systems .The integration

into a complete MSDR

of all-‘required’components

central statiocpower

plant would involve a number ofmsystems for

which development:work is still--required.~. that some components; such asfiumps’and require their


own individualsystems

cooling -and lubric’ation.~




for functions such as


Given the required -ctiponents~and -basic reactor $kima&

It should be noted


and secondaj,

of conktruction,


flow systems can be designed.


However, the primary flow system would require supporting supporting 'systems ‘systems ‘-'--

~-dor-continuous fuel~$rocessiag,on-line fuel~processing,,"an-line --:i-mof~salt chemistry, active:gasesi



1 ->-componentCand -equipment, -afterheat

control during-non-nuclear

fuel~analysis- and control

possible holdup area in



handling of radio-

and- temperature =~---..



The continuous f u e l processing systems proposed t o d a t e are q u i t e complicated and include a number of subsystems, a l l of which would have t o operate s a t i s f a c t o r i l y , within t h e c o n s t r a i n t s of economics, s a f e t y , and r e l i a b i l i t y .

The e f f e c t s of off-design

conditions on these systems would have t o b e understood so t h a t c o n t r o l would b e possible t o prevent inadvertent contamination of t h e primary system by undesirable materials.

The f u e l d r a i n system is important t o both operation and s a f e t y

s i n c e i t would b e used t o contain t h e molten f u e l whenever a need arises t o d r a i n t h e primary system o r any component or instrument for maintalnence or inspection.

systems would b e required, each with its maintaiaing and c o n t r o l l i n g temperatures.

Thus, additional


system f o r

The f u e l s a l t d r a i n

tank would h a v e - t o b e equipped with an a u x i l i a r y cooling Bystem capable of r e j e c t i n g about 18 MJ(th) of h e a t should t h e need a r i e e t o d r a i n the s a l t immediately following nuclear operation.

The secondary coolant system would a l s o require subsystems f o r draining and c o n t r o l l i n g of s a l t chemistry and temperature. a d d i t i o n , the secondary loop might r e q u i r e systems t o c o n t r o l

t r i t i u m and t o handle the consequences of eteam generator or h e a t exchanger leaks.


-43The steam system for-the

MSBRmight require a departure from _

conventional designs hu& to the unique &Able& .associated.with using a coolant having a high melting temperature.


-would have to be taken against freezing the secondary salt as it travels

through the steam generator; suitable methods for system

startup and control would need to be'lncorporated. proposed the &e of a supercritical


steam system which operates

at 3500 psia and provides 700째F feedwater by mixing of supercritical steam and high pressure feedwater.

This -system would introduce

major new development requlrem&nts because it differs . ionveiational stesrn cycles:



- A Difficult


Problem for the &BR

Unlike solid fueled reactors in which the primary system contains act&v&ion products and only those fission products which may leak from < defective fuel pins, the *BR would have the bulk of the fission products dispersed throughout the reactor system. Because of this dispersal of . . __ ., .I radioactivity, remote tech&ues would be required for many maintenance functions if the utility

the reactor were


h&e an acceptable plant availab'ility


environment, :

'I;he MSREwas d&&&d


for remote maintenance of highly radio&tive '.. . &mponents; however, no ajar roainte&&e problems -(rem&ai~ or .repair of _ _large components) were encountered after nuclear opera&on was- initiated.


Thus, t h e degree t o which t h e MSRE experience on maintenance is applicable to l a r g e commercial breeder r e a c t o r s i s open t o question.

As has been evident i n p l a n t layout work on n u c l e a r ~ c i l i t i e st o d a t e , this requirement f o r remote maintenance w i l l s i g n i f i c a n t l y a f f e c t t h e u l t i m a t e design and performance of t h e p l a n t system.

The MSBR would

r e q u i r e remote techniques and tools f o r inspection, welding and c u t t i n g

of pipes, mechanical assembly and disassembly of components and systems, and removing, transporting and handling l a r g e component items a f t e r they become highly radioactive.

The removal and replacement of core

i n t e r n a l s , such as graphite, might pose d i f f i c u l t maintenance problems because of t h e high r a d i a t i o n levels involved and t h e contamination p r o t e c t i o n which would be requfred whenever t h e primary system is opened.

Another p o t e n t i a l problem is t h e a f t e r h e a t generation by f i s s i o n products

which d e p o s i t i n components such as t h e primary h e a t exchangers. Auxiliary cooling might b e required t o prevent damage when t h e f u e l s a l t is drained from t h e primary system, and a requirement f o r such cooling would z

f u r t h e r complicate inspection and malnteaance operations.

I n some cases, the Inspection and maintenance problems of t h e MSBR could be solved using present technology and p a r t i c u l a r l y experience gained from f u e l reprocessing plants.

However, a d d i t i o n a l technology development

would be required i n o t h e r areas, such


remote cuttinp., alignment,





cleaning and-welding of metal*members. Depending to same~-degreeon the ~particular~plant

arrangement~~~ather~special tools and-equipment would

alsohave to be-'designed 'and developed to accomplish inspection and ' -maintenance operations. , In the final-analysis,




the-developinent,of adequate inspection and

maintenance techniques and proceduresland hardware for the~MSBR< hinges on the success of other facets of the program, such as materials and~coinponentdevelopment,'and eonthe~requirement that adequate care be taken during plant design to assure that all systems ahd components which would -require mainte&nce=over the life maintainable within




- Different

of futility

Issues -d-P for-the



The MS&concept-has certain characteristics advantages:relating

of the plant are indeed


:which might provide


to postulated

major types of accidents currently -considered in licensing â&#x20AC;&#x2DC;aktivities. I Since the fuel would be in a molten form,:cons.ideration of the core meltdown accident~is'not of

-a fuel


applicable to'the MSBR. Also,



the event

not-a~problem siuce-this'


athermal re'actor Zystem requiring moderator for nuclear criticality. :__. -~. .~ Oihei safety features'include~ the ofact that Ukprimary operate at lti'pres&e'vith~fuel'salt

system would

that is more~than~lOOO'F~bel& -


its b o i l i n g point, t h a t f i s s i o n product iodine and strontium form

s t a b l e compounds i n t h e f l u o r i d e salts, and that t h e salts do not react rapidly with air or water.

Because of t h e continuous f u e l processinfz,

t h e need f o r excess r e a c t i v i t y would b e decreased and some of the f i s s i o n products would b e continuously removed from t h e primary system.

A prompt

negative temperature c o e f f i c i e n t of r e a c t i v i t y i s a l s o a c h a r a c t e r i s t i c

of t h e f u e l s a l t .

Safety disadvantages, on t h e o t h e r hand, include t h e very high radioa c t i v e contamination which would be present throughout the primary system, f u e l processing p l a n t , and a l l a u x i l i a r y primary systems such

as the f u e l d r a i n and off-gas systems. systems would have t o b e assured.

Thus, containment of these

A l s o , removal of decay h e a t from f u e l

s t o r a g e systems would have t o b e provided by always ready and r e l i a b l e cooling systems, p a r t i c u l a r l y f o r t h e f u e l d r a i n tank and t h e Pa-233 decay tank i n the reprocessing p l a n t where megawatt q u a n t i t i e s of decay h e a t must b e removed.

The t r i t i u m problem, already discussed, would have t o



be controlled t o assure safety.

Based on t h e present state of MSBR technology, i t i s not possible t o provide a complete assessment of MSBR safety r e l a t i v e t o o t h e r r e a c t o r s . It can b e s t a t e d , hawever, t h a t the s a f e t y i s s u e s f o r the MSBR are

generally d i f f e r e n t from those f o r s o l i d f u e l reactors, and t h a t more d e t a i l e d design work must b e done before t h e s a f e t y advantages and disadvantages of t h e MSBR could be f u l l y evaluated.







Codes, Standards,,~and High Temperature

Codes and standards conjunction be built. currently in. nuclear


Design Methods

for MSBR equipment and systems must be developed in ~,

other research

In particular,

and development before _ of construction Ithe materials

large MSBR's can ~. which are

being developed and tested would have to be certified power.plant

for use

applications. . .'

The need for high temperature design-technology _-. .~ .~ as well as for other high temperature systems. under.way a program in support

is a problem for The ARC currently

of the LMFBR which is providing

the MSBR has


data and structural analysis methods for design of systems employing : various temperatures up to 12OO'F. This program would ~~ steel alloys-at .. need to be broade$ed to include MSBR structural materials such as : tiastelloy-N and to include temperatures as high as 1400eR to provide ._ the _design applicable3 to high-temperature, long-term . . technology .-. : :operating conditions which would be expected for MSBR vessels, components, and core structures. es

! -48-




P r i v a t e l y funded conceptual design s t u d i e s and evaluations of HSBR technology were performed i n 1970 by t h e Molten S a l t Breeder Reactor Associates (HSBRA), a study group headed by t h e engineering firm of

Black & Veatch and including f i v e m i d w e s t u t i l i t i e s .

The MSBRA con-

cluded that the economic p o t e n t i a l of t h e MSBR is a t t r a c t i v e r e l a t i v e t o l i g h t water r e a c t o r s , b u t they recognized a number of problems which must b e resolved i n order t o realize t h i s p o t e n t i a l .

Since t h a t t i m e

t h e HSBRA has been r e l a t i v e l y i n a c t i v e .

A second p r i v a t e l y funded organization, t h e Molten S a l t Group, i s headed by Ebasco Services, Incorporated and includes f i v e o t h e r i n d u s t r i a l firms and f i f t e e n u t i l i t i e s .

I n 1971 t h e Group completed an evaluation of t h e

MSBR concept and technology and concluded t h a t e x i s t i n g technology is s u f f i c i e n t t o j u s t i f y construction of an MSBR demonstration p l a n t althdugh t h e performance c h a r a c t e r i s t i c s could n o t be predicted with confidence.

Additional support f o r f u r t h e r s t u d i e s has r e c e n t l y been

committed by t h e members of t h i s group.

I n a d d i t i o n t o these s t u d i e s , manufacturers of g r a p h i t e and Hastelloy-N have been cooperating vith ORNL t o develop improved materials. I

There has been l i t t l e o t h e r i n d u s t r i a l p a r t i c i p a t i o n i n t h e MSBR

Program a s i d e from ORNL subcontractors


At t h e p r e s e n t time, t h e r e are


two ORNLsubcontracts

Ebasco Services,

in effect.


firma who are: p_articip+nta


a deiPfgn and evaluation





the Molten,_ Salt_. croup is


-study. 1-~

Foster Wheeler Corporation : i8

on~steam generators

design studies


for MSBR


A number of factors indukrial


which tend to.lbnit

can be identified

Vinvolvement at this



The existing major industrial

and utility



The lack of incentive-for fuel



commitment8 to the ., -,



services, , such as those required



for solid



, 3.

The overwhelming manufacturing solid _


merit with


reactor8 in contrast with fluid


The less advanced state deknstrated

and operating


associated with


the very limited




of MSBR technology to the major technical

the MSBR concept.

and the lack of problems ~\


It should be noted that these factors are also relevant considerations

i n establishing the level of governmental support for the MSBR program

which in turn, to some extent, affects the interest of the manufacturing and u t i l i t y industries.




The Molten Salt Breeder Reactor, if successfully

developed and marketed,

'&uld provide a useful supplement to the currently plutonium reactor economy. This concept offers .

Breeding in a thermal spectrum reactor;




developing uranium-

the potential


use of thorium as a fertile~material; ", i Elimination.of 'fuel fabrication and spent fuel shippink;

. High thermal efficiencies. Notwithstanding

these attractive



this assessment has

reconfirmed the existence of major technological and engineering problems affecting feasibility ~-'of: the'concept as a reliable and 'i -~The principal _. concerns economic breeder for the utility industry. include uncertainties

with ..~ materials, _

with methods




tritium, and with the design of components and systems along with , , their special handling, inspection and maintenance equipment. Many of thes'e problems are &mpounded ~- by the use of a fluid fission


and delayed neutrons are distributed

primary reactor and reprocessing The resolution

of'an IntkUive _ the major effork




_ of the problems of the MSBR will

the\ conduct research and development program. -kcluded among that &&la

. Proof testing.of .

fuel in which


have to bt? accompiishedare:


an integrkged reprocessing system;

Development of asuitable

; : containment material;

_. ~I *




Development of a s a t i s f a c t o r y method f o r t h e c o n t r o l and _


r e t e n t i o n of t r i t i u m ;

. Attainment of a thorough understanding of

t h e behavior of f i s s i o n


products i n a molten s a l t system;


Development of long' l i f e moderator graphite, s u i t a b l e for breeder ' applica



Conceptual d e f i n i t i o n of the engineering f e a t u r e s of t h e many components and s y s tems ;



Development of adequate methods and equipment for remote inspection, handling, -and maintenance of t h e p l a n t .

The major problems associated with the MSBR are rather d i f f i c u l t In n a t u r e and many are unique t o t h i s concept.

Continuing support of the research

and development e f f o r t w i l l b e required t 6 o b t a i n s a t i s f a c t o r y s o l u t i o n s

t o the problems.

When s i ~ q ~ i f i c a n evidence t i s a v a i l a b l e t h a t demonstrates

realistic s o l u t i o n s are practical, a f u r t h e r assessment could then be made as t o t h e a d v i s a b i l i t y of crdvancini i n t o engineering phase of the development process inclu+g involvement.

t h a t of i n d u s t r i a l

Proceeding with t h i s next s t e p would -a l s o be contingent

upon obtaining a f i r m demonstration of i n t e r e s t and commitment t o t h e concept by t h e power industry and t h e u t i l i t i e s and reasonable assurances t h a t large scale government and i n d u s t r i a l resources can b e made a v a i l

on a continuing b a s i s t o t h i s program i n l i g h t of o t h e r canrmitments t o the commercial nuclear power program and higher p r i o r i t y energy

development e f f o r t s








U. S. Atomic l%uergy'Conunission,"The 1967~Supplementto the 1962‘


Rep&t to the Preaidkt

on Civilian

Nuclear Power" USMC Repoit,

- February 1967. U. S. Atomic Energy -&nrkk&m,


%he ~&x-of Tho&un%a Nuclear

Power Reactom" USA&-Rep& WASH-1097,

1963.' -A

U. S. Atomic Ene&y Comaiissibn, "P;'otentkl

, 3.


US&i%-Rep& tiASiGlCJ98,

' ” Nuclear Power Gro&h


U. S. Atomic Energy Cormission~ "Cost-Benefit


U. S. Breeder ~R&&or%ogrk;"

1970. &naly&



USAEC'keiort GASR-1126,“1969.

U. S. Atomic Energy Comi&Lss&n~"Updateid-(1970) Co&B&efit




of the U. S. Breeder Reactor Program,' USAECReport WASH-1184, . ,. .January 1972.. ~~ .Edison Electric'ktitute,


"'&port bn the EEf Reactor Assessment Panel,"

EEI Publication No. 70-30, 1970. AnntkHekngs



on &i&tor$ev~lo~m&t

&mmisS~ionh 1972~Aut&i&g Committee on Atomic

P~o&.m, U. S.'At&ic



Hear&&s before -&Joint 7~ Energy, Congress of the United-States p. 820-830,


Molten salt Breeder Reactor,'


ORAL-4541, June 1971.' .





Roaenthal, M. I?.,

et al.; "Advances in t h e Development of Molten-Salt

Breeder Reactor8," A/CONF. 49/P-048, I n t e r n a t i o n a l Conference


Fourth United Nations

t h e Peaceful Uses of Atomic Energy,

Geneva, September 6-16, 1971.



Trinko, J. R.

"Molten-Salt Reactor Technology," Technical

Report of t h e Molten-Salt Group, P a r t I, December 1971.



Trinko, J. R. (ed), "Evaluation of a 1000 We Molten-Salt Breeder Reactor," Technical Report of t h e Molten S a l t Croup, P a r t 11,


Navemb er 1971


Ebasco Seroices Iuc., "1000 M W e Molten S a l t Breeder Reactor Conceotual

Design Study," F i n a l Report Task I, Prepared under ORNL subcontract 3560, February 1972.



"Project f o r I n v e s t i g a t i o n of Molten S a l t Breeder Reactor,''


Report, Phase I Study f o r Molten S a l t Breeder Reactor Associates, Sepfember 1970. Cardwell, D. W. and Haubenreich, P. N.,

"Indexed Abstracts of

Selected References on Molten-Salt Reactor Technology," ORNL-TM-3595,

December 1971. 16


Kasten, P. R.,

B e t t f s , E. S. and Robertson, R. C.

, "Design


of 1000 MW(e) Molten-Salt B k e d e r Reactors ,"ORNL-3996, August 1966 17.

Molten S a l t Reactor Progr'am SPm$annual Reports beginning i n February 1962.




A- 1

i f


General Thermal capacity of reactor Gross electrical generation Net electrical output Net overall thermal efficiency Net plant heat rate

2250 MW(t) 1035 MW(e) 1000 MW(e) 44.4% 7690 Btu/kWhr

2250 MW(tj 1035 MW(e) 1000 MW(e) 44.4% 2252 J/kW-SC

Structures Reactor cell, diameter X height Confinement building, diameter X height

72 x 42 ft 134 X 189 ft

22.0 X 12.8 m 40.8 X 57.6 m

20 ft 2 in. 3 in. 75 psi 13ft ' 1412 30 in. 0.13 0.37 22.2 kW/liter 70.4 kW/liter 2.6 X 1014 neuGons cm-' sec-' 8.3 X lo'* neutrons sec-' 3.5 x 10" neutrons cm" sec-' 3.3 x 1014 neutrons cm-? set"

6.77 m 6.1 m 5.08 cm 7.62 cm 5.2 X los N/mz 3.96 m 1412 0.762 m 0.1 3 0.37 22.2 kW/liter . 70.4 kW/liter 2.6 X 1014 neutrons scc-' 8.3 x neutrons cm-l scc-l 3.5 x 1014 neutronscrnm2 scc-l 3.3 x 1014 neutroks cm-3 sw-'

1284' F




4 years 669,000 Ib 8.5 fps 1074 ft3 1720 ft3 3316 Ib

4 years 304,OOOkg 2.6 m / x c 30.4 m3 48.7 m3

Reactor Vessel ID Vessel height at center (approx) Vessel wall thickness Vessel head thickness Vessel design pressure (abs) Core height > Number of core elements Radial thickness of reflector Volume fraction of salt in central core zone Volume fraction of salt in outer core zone Average overall core power density Peak power density in core Average thermal-neutron flux Peak thermal-neutron flux Maximum graphite damage flux (>SO keV) Damage flux at. maximum damage region (approx) Graphite temperature at max flux region Graphite temperature at maximum graphite damage region Estimated useful life of graphite Total weight of graphite in reactor Maximum flow velocity of salt in core Total fuel salt in reactor vessel Total fuel-salt volume in primary system Fissile-fuel inventory in reactor primary system and fuel processing plant Thorium inventory Breeding ratio Yield Doubling time, compounded continuously. at 80%power factor


, 1

Primary heat exchangers (for each Thermal capacity. each Tube-side conditions (fuel salt) Tube OD Tube length (approx) Number of tubes Inlet-outlet conditions Mass flow rate Total heat transfer surface Shell-side conditions (coolant salt) Shell ID Inlet-outlet temperatures Mass flow rate Overall heat transfer coefficient (approx)








, .e


ii i


I50.000 Ib 1.06 3.2 %/year 22 years

1.06 3.2 %/year 22 years





17.6 X lo6 lb/hr 850 Btu lu" ft-2


1.73 m 727-894" K 2218 kg/sec 4820 W m-2




Appendix A (continued) Engineering unit? Primary pumps (for each of 4 units) Pump capacity, nominal Rated head Speed Specific speed impeller input power Design temperature Secondary pumps (for each of 4 units) Pump capacity, nominal Rated head Speed, principal Specific speed Ihpcller input power Design temperature


Fuel-salt drain tank (1 unit) Outside diameter Overall height Storage capacity Design pressure Number of coolant U-tubes Sue of tubes, OD Number of separate coolant circuits Coolant fluid Under normal steady-state conditions: Maximum heat load Coolant circulation rate Coolant temperatures, in/out Maximum tank wall temperature Maximum transient heat load


Fuel-salt storage tank (1 unit) Storage capacity Heat-removal capacity Coolant fluid Coolant-salt storage tanks (4units) Total volume of coolant salt in systems Storage capacity of each tank Heat-removal capacity. f m t tank in series Steam generators (for each of 16 units) Thermal capacity Tube-side conditions (steam at 3600-3800 psi) Tube OD Tube-sheet-to-tube-sheet length (approx) Number of tubes I nletbutle t temperatures Mass flow rate Total heat transfer surface Shell-side conditions (coolant salt) . Shell ID Inlet-outlet temperatures Mass flow rate Apparent overall heat transfer coeffKient range

International systcm.units*

16,000gpm 150 ft 890 rpm . 2625 r p m ( g p ~ n ) ~ ' ~ / ( f t ) ~ * ~ ~ 2350 hp 1300"F

1-01 m'/sec 45.7 m 93.2 radianshc 5.321 radian~/sec(m~~sec)~.~/(m)~~~~ 1752 kW 978°K

20,000 gpm

1.262 m3/sec 91.4 m 124.6 radianJsec 4.73 radianJsec(m' /sec)O 4 /( m)O *7s 2310 kW 978°K

300 ft 1190rpm 2330 rpm(gpm)0'5/(ft)0'7s 3100 hp 1300°F

4.27 m 6.71 m 70.8 m3 3.79 X los N/m2

14 ft 22 ft 2500 ft3 55 psi 1500


1500 1.91 cm


40 'LiF-BeFs

18 MW(t) 830 gpm 900-1050°F -1260°F 53 Mw(t)

18 MW(t) 0.0524 m3/sec 755-839°K -955°K 53 MW(t)

2sOOft3 -

70.8 m' 1 MW(t)

1 MWW Boiling water



8400 ft' 2100 ft3 400 Irw

237.9 m3 59.5 m' 400 kW

120.7 MU'( t)

120.7 MW(t)

y in.

1.27 cm 23.3 m 393 644-811°K 79.76 kg/sec 365m' .

76.4 ft 393 700- 1000"F 633,000 lbbr 3929 ft2 .

1.5 ft

1150-850°F 3.82 X lo6 lb/hr 490-530 Btu hr" 11-



0.457 m 894-727% 481.3 Wsec 2780-3005 W m-2 ("K)'t


Appendix A (cont h u e d ) .~

Engineering unit.@ Steam reheaters (for each of 8 units) Thermal capacity Tube-side conditions (steam at 5sO psi) Tube OD Tube length Number of tubes Inlet-outlet temperatures Mass flow rate Total heat transfer surface Shell-side conditions (coolant salt) Shell ID Inlet-outlet temperatures Mass flow rate Overall heat transfer coefficient Turbine-generator plant (see “General” above) Number of turbme-generator units Turbine throttle conditions Turbine throttle mass flow rate Reheat steam to IP turbine Condensing pressure (abs) Boiler feed pump work (steamturbine-driven), each of 2 units Booster feed pump work (motordriven), each of 2 units Fuehalt inventory, primary system Reactor Core zone 1 Core zone 11 Plenums, inlets, outlets 2-in. annulus Reflectors Primary heat exchangers Tubes Inlets, outlets Pump bowls Piping, including drain line


International system unitsb

36.6 W ( t )

36.6 MW(t)

36 in. 30.3 ft 400 650- 1000”F 64 1,000 Ib/hr 2381 Et2

1.9 cm 9.24 m 400 616-811°K 80.77 kg/sec 221.2 m2

21.2 in. 1150-850°F 1.16 X IO6 Ib/hr 298 Btu hr’’ ft’)

0.54 m 894-727°K 146.2 kg/sec 1690 W m-2 (OK)-*

3500 psia. 1000°F 7.15 X lo6 lb/hr 540 psia. 1000°F 1.5 in. Hg 19.700 hp


1 24.1 X IO6 N/m2, 811°K 900.9 kg/sec 3.72 X lo6 N/m2. 81 1°K 5,078 N/m2 14,690 kW

- 6200hp

4620 kW

290 ft’

8.2 d 10.8 m5 6.2 m3 3.8 m’ 1.4 m3

382 ft’ 218 ft’ 135 ft’ 49 ft’ 269 ft’ 27 ft3 185 ft3 145 ft’

Off-gas bypass loop

7.6 m’ 0.8 m’

5.2 4.1 0.3 0.3


m3 m3

m3 m3

Tank heels and miscellane Total enriched salt in primary system Fuet-processing system (Chemical Treatment


Cyde time for salt inventory Heat generation in sa!t to processing plant


propertie Components Compositioh . Molecular weight (approx) Melting temperature (approx) Vapor pressure at 1150°F (894.3%) Wan’) = 3.752 6.68 x 10*r P (Uft’) = 235.0 0,02317t(%“)


At 1 3 W F (978°K) At 1175°F (908°K) At 1050°F (837K)


204.9 Ib/ft’ 207.8 lb/ft3 210.7 lb/ft3

3283.9 Wm3




Appendix A (continued) Engineering unitP Viscosity:d p (centipoises) = 0.109 exp [4090/T('K)];p (Ib it-' h-l) = 0.2637 exp [7362/T("R)] At 1300°F (978°K) At 1 175°F (908°K) At 1050°F (839°K) Heat capacitf (specific heat, cp) Thermal conductivity (k)f At 1300°F (978°K) At 1175°F (908°K) At 1050°F (839°K)

17.3 Ib hr-' ft-' 23.8 lb hr-' ft" 34.5 Ib hr-' ft-' 0.324 Btu Ib" PF)"

Design properties of cootant salt Component s Composition Molecular weight (approx) Melting temperature (approx) Vapor pressure:g log P (mm Hg) = 9.024 - 5920/T("K) At 850°F (727°K) At 1150°F (894°K) Density:=p (g/cm3) = 2.252 - 7.1 1 X 104t ("C); p Curt3) = 141.4 0.0247r (OF) At 1150°F (894°K) At 1000"F(81l0K) At SSO" F (727"K) Viscosity:d p (centipoises) = 0.0877 exp [2240/T (OK)];p (Ib, ft" hr-l) = 0.2121 exp [4032/T('R)) At 1150"F(894°K) At 1000°F (811°K) At 850°F (727°K) Heat capacityh (specific heat, cp) Thermal conductivity (k)' At 1150"F(894°K) At 1M)O"F(811OK) At 850°F (727°K)

* 4%

International system unit#

0.007 N sec m-2 0.010 N sec rn3 0.015 N sec m-2 1.357 J g" PIC)''

0.69 Btu hr-' Po-' ft" 0.71 Btu hr" ("F)" ft-' 0.69 Btu hr" ("F)-' ft-'

1.19 W m-' CK)-' 1.23 W m-' VK)" 1.19 W m-' PK)"

NaBF4-NaF 92-8 mole % 104 725°F

NaBF4-NaF 92-8 mole % 104 658°K

8 m m Hg 252 mm Hg

1066 N/m2 33,580 N/m2

113.0 lb/ft3 116.7 Wft3 120.4 lblft'

1811.1 kg/m3 1870.4 kg/m3 1929.7 kg/m3

2.6 Ib ft-' hr" 3.4 lb ft-' hr-' 4.6 lb ft-' hr-' 0.360 Btu lb-' (OF)"

0.001 1 N sec m-2 0.0014 N sec in-2 0.0019 N sec m-2 1507 J kg-' (OK)-'

* 4%


Design properties of graphitei Density. at 70°F (294.YK) Bending strength Modulus of elasticity coefficient Poisson's ratio Thermal expansion coefficient Thermal conductivity at 1200°F. unirradiated (approx) Electrical resistivity Specific heat At 600°F (588.8"K) At l 2 W F (922.0"K) Helium permeability at STP with sealed surfaces


0.23 Btu hr" 0.23 Btu hr" 0.26 Btu hr"


* 2% ft-' ft" ft"


(OF)" (OF)"

115 lb/ft3 4000-6000 psi 1.7 X lo6 psi 0.27 2.3 X 10df'F 18 Btu hr-' (OF)" ft-' 8.9 X 104-9.9X

lo4 n-cm

0:33 Btu lb-' (OF)-' 0.42 Btu lb" (OF)-' 1 x 10" cm2/sec


* 2%

0.398 W m-l (OK)" 0.398 W an-' ( O K ) - ' 0.450 W m-' (OK)-' 1843 kg/m3 28 X 106-41 X lo6 N/m2 11.7 X lo9 N/m2 0.27 1.3 X 10dPK 31.2 W m-I (OK)" 8.9 X 104-9.9 X 1380 J kg" (OK)" 1760 J kp-' (OK)" 1 x IO* cm2/sec

lo4 n t m


Appendix A (continued) Engineering unitP Design properties of Hastelloy Nk Density At 80°F (300°K) At 1300°F (978°K) Thermal conductivity At 80°F (300°K) At 1300°F (978°K) Specific heat At 80°F (300°K) At 1300°F (978°K) Thermal expansion At 80°F (300°K) At 1300"F(978"K) Modulus of elasticity coefficient At 80°F (300°K) At 1300°F (978°K) Tensile strength (approx) At 80°F (300°K) At 1300°F (978°K) Maximum allowable design stress C t 80"F (300°K)

At 1300°F (978°K) Melting temperature




International system unitsb

557 lblft3 541 lb/ft'

8927 kg/m3 8671 kg/m3

6.0 Btu hr-' CF)" ft-' 12.6 Btuhr" ("F)" ft-'

10.4 W m-' (OK)-' 21.8 W m-I (OK)-'

0.098 Btu lb-' (OF)-' 0.136 Btu lb-' (OF)"

410 J kp-' ("IC)-' 569 J kg-' (OK)-'

5.7 xFP'OI 9.5 x 10"PF

3.2X 1O"PK 5.3 X 10df'K

31 X 25 X

lo6 psi lo6 psi

214 X lo9 N/m2 172 X lo9 N/m2

115,000psi 75.000 psi

793X 106N/m2 517 X lo6 N/m2

25,000 psi 3500 psi 2500"F

172 X IO6 N/m2 24 X lo6 N/m2 1644°K


'English engineering units as used in MSR literature. bMeter-kilogram-second system. Table closely follows International System (SI). See Appendix C for conversion factors from engineering to SI units.