0 ~ ~ ~ 2 9 8 5 Reactors-Parer TD-4500 (15th ea.)
Contract No. W-7405-eng-26
TBE FEASIBILITY OF AN UNALPTEXDED IWCLEAR POWER PLANT
M. M. C. R.
W. Rosenthal M. Yarosh
F. Baes, Jr. C. Robertson
SEP 8 1960
OAK RIDGE NATIONAL Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY CWSSION
TABU OF COIWENTS
........................................................ INTRODUCTION .................................................... SOME CO-S ON THE ACHIEVEMENT OF RELIABILITY ................. SELECTION OF A REACTOR POWEB PLAl!ZP .............................. Reactor Type .................................................... Abstract
........................................... CONCEPT OF A SIMPLIFIED PRESSTJRIZED-Wp3cER REACTOR ............... Primary System .................................................. Reactor Control ............................................ Fuel Elements .............................................. Primary Coolant Pumps ...................................... Core Pressurizer ........................................... Main Heat Exchanger ........................................ Water Treatment ............................................ Secondazy System ................................................ Power Recovery System
............................. .............................. ........................... .......................... EFFECT OF ALTERED REQUIRESIENTS ON THE DESIGN CONCEPT ............ Communication with the Plant .................................... Simplified Conventional System Semiconventional Steam System Hermetically Sealed Steam System Basic Features of t h e Steam Cycle
........................................... Long-Term R e l i a b i l i t y and Expendability ......................... Variable Loads .................................................. Effects of Time Schedule ........................................ SUMMARY AX0 CONCLUSIONS ......................................... REFERENCES ....................................................... Plant Size and Weight
17 24 26
30 33 34 35 36
37 53 53
56 56 60
THE FEASIBILITY OF AN UNATTENDED NUCLEAR POWER PLANT
A study has been made of the f e a s i b i l i t y of constructing a small nuclear power plant capable of operating one year completely unattended. A system which w i l l produce 1 Mw of e l e c t r i c i t y without interruption i s required f o r f i e l d use i n about four years. It i s t h e opinion of t h e authors t h a t t h e four-year requirement g r e a t l y r e s t r i c t s t h e opportunity f o r development of new concepts and new technology and t h a t t h e objective i s most l i k e l y t o be a t t a i n e d by extreme simplification of a type of r e a c t o r with which there has been favorable experience. A pressurized-water r e a c t o r w a s selected f o r t h e application, both because of t h e extensive experience with it and because it appears r e a d i l y adaptable t o simplification. For example, t h e power plant suggested has hermetically-sealed primary and secondary systems, and a l l control systems, except t h e turbine governor, have been eliminated. The authors conclude that it i s f e a s i b l e t o develop a r e l i a b l e , simplified, pressurized-water reactor system f o r unattended service i n t h e a l l o t t e d time.
INTRODUCTION A study w a s undertaken by staff members of t h e Oak Ridge National
Laboratory t o evaulate t h e f e a s i b i l i t y of constructing a small nuclear power plant capable of operating unattended f o r one year.
of t h e study were, specifically, (1)t o assess t h e p o t e n t i a l f o r success
Y A s w i l l be evident throughout t h e report, t h e design considerations
have been r e s t r i c t e d t o meeting only t h e s t a t e d objectives of t h e study,
It w a s t h e opinion of t h e authors t h a t t h e achievement of one year of unattended operation was a problem of such d i f f i c u l t y as t o dominate a l l other design considerations.
Thus the question, w h a t makes a plant
cease t o function, was continually i n the forefront throughout t h e study. That this apparent obsession with r e l i a b i l i t y i s a r e q u i s i t e may be a t t e s t e d t o by t h e experience records of those reactors which a r e operating today. For t h e present study the following system requirements were established:
1. The plant i s t o operate unattended a t f u l l power continuously
for a minimum of one yeax. 2.
The net e l e c t r i c a l output i s t o be 1000 kw of three-phase a l t e r -
nating current a t 4160 v.
The frequency can be selected by t h e designer within a range
of 60 t o 1000 cycles, but the frequency selected must be held t o within
kl$ throughout t h e l i f e of t h e plant.
The reactor plant will be t h e only power source connected t o
i t s e l e c t r i c a l load.
A reactor power plant capabie of satisfying t h e preceding
requirements i s t o be i n operation on location within approximtely four years, 6.
Interruption of t h e generation of e l e c t r i c i t y f o r any reason
within one year of i n i t i a t i o n of operation i s considered a plant f a i l u r e .
Short-term (onemyear) r e l i a b i l i t y i s of utmost importance and
w i l l not be sacrificed t o provide long-term l i f e . 8. The system could be considered expendable, i f necessary, a t t h e end of one year of operation.
During s t a r t u p and shakedown of equipment, semiremote operation
of t h e plant will be possible, and special equipment may be provided f o r s t a r t u p requirements.
After t h e beginning of unattended operation, no comrm;mication
with t h e system will be possible other than knowledge that t h e e l e c t r i c a l
load i s being supplied. 11. Within reason, plant s i z e and weight a r e not considerations.
Within reason, cost i s not a consideration.
Plant efficiency i s not an important consideration.
14. The system w i l l operate stationary.
The plant will not operate i n a populous area.
The sink f o r heat r e j e c t i o n w i l l be determined by t h e appli-
cation, but adequate means of cooling will be available. c
The time l i m i t a t i o n s imposed on this study precluded an extensive examination of a l l systems which might conceivably s a t i s f y t h e requlree ments.
Hence, only those systems t h a t appeared t o o f f e r t h e maximum
p o t e n t i a l f o r r e l i a b i l i t y were studied.
There i s no i n t e n t t o imply
that t h e systems discussed i n this report comprise t h e only ones capable
of satisfying t h e plant requirements, but, within t h e framework of r e l i a b i l i t y considerations f o r t h i s system, they appem t o be t h e most promising. The authors a r e deeply indebted t o various members of t h e Oak Ridge National Laboratory and t h e O a k Ridge Gaseous Diffusion Plant staffs f o r numerous contributions t o t h i s study.
L. D. Schaffer helped apprise t h e
authors of t h e s t a t u s of small-reactor development.
E. R. Mann advised
secondary system and contributed information pertinent t o component selection and r e l i a b i l i t y . During the course of t h e study, discussions and correspondence were carried on with personnel of a number of organizations.
Many of t h e
opinions expressed i n this report a r e a r e s u l t of such discussions, and t h e authors g r a t e f u l l y acknowledge the following organizations f o r t h e i r valuable contributions : Aerojet General Corporation (GCRE)
Alco Products (Dunkirk F a c i l i t y ) AllisLChalmers Manufacturing Company Atomics International Combustion Engineering (Windsor F a c i l i t y and SL-1) The E l l i o t t Company (Jeannette, Pennsylvania) General E l e c t r i c Company (Erie, Pennsylvania; Fitchburg, Massachusetts; and Schenectady, New York) Gilbert Associates International Nickel Company Martin Company (Baltimore, Maryland) Personnel of SM-1 and GTTF, Fort Belvoir, Virginia Pierce Governor Company Westinghouse E l e c t r i c Corporation (Bettis Plant, E a s t Pittsburgh, and Lester, Pennsylvania) Woodward Governor Company Worthington Corporation (Harrison, New Jersey) Gilbert Associates kindly made available t o the authors a d r a f t of t h e i r forthcoming report'
on t h e r e l i a b i l i t y of reactor components.
SOME COMMElYTS ON T a ACHIEVEMENT OF RELIABILITY I n considering t h e objectives of this study, it i s meaningless t o speak of building a reactor system which w i l l operate without f a i l u r e f o r one year.
To do so would suggest we a r e requiring it t o be a cer-
t a i n t y t h a t a p a r t i c u l a r reactor power plant operate successfully f o r that period.
Actually, a more r e a l i s t i c specification might be t h a t we
expect nine out of t e n reactors t o be operating at t h e end of a year,
which i s equivalent t o requiring t h a t an individual reactor have a r e l i a b i l i t y of 0.9.
For purposes of this study, it was assumed t h a t a plant r e l i a b i l i t y of 0.9 w a s required f o r one year of unattended operatic Some understanding of what i s involved i n achieving a 0.9 r e l i a b i l i t y may be obtained by considering how t h e r e l i a b i l i t y of t h e individual components a f f e c t s t h e over-all r e l i a b i l i t y of the power plant.
If a system
consisting of a number of components i s t o have a r e l i a b i l i t y of 0.9, t h e r e l i a b i l i t y of each component i n t h e system must be b e t t e r than 0.9 and an average component r e l i a b i l i t y very much b e t t e r than 0.9 may be necessary. T h i s i s i l l u s t r a t e d by Eq. (1)i n which pi i s the probability that component "if' w i l l l a s t one year, and P i s t h e probability that a l l t h e components w i l l l a s t one year: = PlPZP3
. . pi .
For example, i f t h e system consists of 10 components, each having a
r e l i a b i l i t y of 0.99, t h e probability of t h e system l a s t i n g one year without failure of any of t h e components w i l l barely exceed 0.90.
there a r e 50 components, t h e f a i l u r e of any one of which w i l l stop the functioning of t h e system, t h e "average" r e l i a b i l i t y of each component must be 0.998.
The significance of a 0.99 requirement may be more
c l e a r l y understood i f one pictures a t e s t system i n which 100 i d e n t i c a l components a r e simultaneously set i n t o operation.
For a 0.99 r e l i a b i l i t y ,
99 of the 100 components must s t i l l be operating at the e n d of a year.
Actually, =the problem i s more complicated than j u s t assessing t h e probability t h a t a p a r t i c d m reactor can operate unattended f o r one
No nuclear power plant having t h e needed capacity has been b u i l t t o operate without regular maintenance. For reactors such as t h e 524-1 (APPR) and t h e SLel (ALPR), it i s year.
without shutdown, even with maintenance operations being performed during estion of whether an e x i s t i n g system t h a t period. Hence, it i s not can meet t h e specifications, b
ether a modification of an existing
system o r a new design of reactor can be made s u f f i c i e n t l y r e l i a b l e f o r unattended operation.
The problem i s t o estimate t h e likelihood t h a t a
r e a c t o r system having t h e necessary r e l i a b i l i t y can be developed i n about four years. I n t h i s study we have attempted t o d i s t i n g u i s h between (1)t h e probability that t h e development problems associated with unproven concepts or untested equipment can be solved i n a limited time period, and
(2) t h e probability that a p a r t i c u l a r system based on t e s t e d equipment and proven concepts can operate f o r a year unattended. For convenience, we shall designate t h e f i r s t probability as P1 and t h e second as P2. One can conceive of systems which, because of inherent features, appear t o be capable of long periods of unattended operation (have a high value of Pz) a f t e r some important development problems associated with them a r e solved.
An example might be a natural-convection liquid-
f u e l r e a c t o r that uses thermoelectric devices f o r t h e generation of electricity.
There would be no moving p a r t s t o wear and no control
system t o f a i l .
A power plant of t h i s type, i f developed successfully,
would probably operate unattended f o r long periods.
However, t h e r e
would be a,n appreciable r i s k (low P1) i n proceeding t o develop it f o r attainment of t h e objectives, since some of t h e development problems might s t i l l be unsolved a t t h e end of f o u r years. On t h e other hand, a r e a c t o r very similm t o t h e SM-1 r e a c t o r might be used.
If no major changes were made i n t h e control system,
water-treatment system, etc.,
one could be confident that a r e a c t o r of
this type could be b u i l t within the time allowed and that it would
operate a t i t s design capacity, but it would not operate very long without attention.
T h i s r e a c t o r concept might be said t o have a high P1
value but a low P2 value.
Since it is the product of P1 and P2 t h a t
gives t h e probability of successfully developing a system of t h e required r e l i a b i l i t y i n t h e time a l l o t t e d , close a t t e n t i o n must be given t o each i n selecting a program t o follow. The language of mathematical probability may help t o c l a r i f y t h e problem, but t h e d i f f i c u l t y of evaluating t h e p r o b a b i l i t i e s remains. The uncertainty i n Pl i s i l l u s t r a t e d by several r e a c t o r development programs which have not a t t a i n e d t h e i r objectives a f t e r a number of
years of e f f o r t .
While solutions may eventually be found t o t h e i r
problems, a requirement of success i n a limited time would not have been Estimation of P2 i s a l s o uncertain, since it requires knowledge
of t h e r e l i a b i l i t y of a number of components.
Although the r e l i a b i l i t y
of a component which has been extensively t e s t e d may be known with some assurance, even a small change i n design can a l t e r t h e value importantly. Thus l i t t l e confidence would be placed i n an estimate of t h e r e l i a b i l i t y of a complex piece of equipment not yet designed.
I n order t o assess t h e r e l i a b i l i t y of existing equipment, operation reports on a f e w reactors (SM-1, SL-1, PWR) were perused i n order t o determine the frequency and causes of f a i l u r e s , and operating experience
was discussed with personnel a t several i n s t a l l a t i o n s (SM-1, SL-1, GTTF).
A study w a s a l s o made of t h e operation h i s t o r y of a number of
moderate-sized steam turbines. The reactors examined were not meant t o operate unattended, and they were designed f o r regular maintenance.
Hence, it w a s expected that
these systems, having hundreds of mechanical components and large numbers of electronic devices, would be far l e s s r e l i a b l e than required f o r one year of unattended service.
More surprising and sobering w a s t h e con-
clusion t h a t f e w existing components i n normal use would operate f o r one year without maintenance.
This conclusion i s l e a s t applicable t o t h e
mechanical equipment i n t h e primary system of pressurized-water reactors, because intensive e f f o r t i n t h e N a v y program has produced r e l i a b l e components.
It applies very strongly, however, t o steam power equipment,
where spares are provided f o r many items, and appreciable on-stream maintenance i s t h e normal practice. vacuum tubes a r e a l s
Electronic control devices that use
t e unreliable f o r one year of service.
r a t i o n i s provided by t h e SM-1 control system, i n which
made during t h e first year of operation.2 were of p a r t i c u l a r i n t e r e s t because they indicate how r e l i a b i
From a study of t h e
ing h i s t o r y of a nwnber of conventional 25-Mw turbine-
ers3 concluded t h a t only f i v e u n i t s i n 100 would be
0 capable of operating continuously f o r one year.
Myers' study was
extended t o t h r e e 3000-kw turbine-generator units, and
for t h e r e l i a b i l i t y w a s again obtained.
value of 0.05
Further examination of t h e
f a i l u r e s of t h e smaller systems indicated, however, t h a t i f t h e need f o r replacing t h e generator brushes and repacking t h e admission valves could be eliminated, approximately 60 units i n 100 could operate continuously f o r one year. There a r e t h r e e ways one might proceed t o improve t h e r e l i a b i l i t y of a system which i s t o operate unattended:
components by eliminating t h e i r functions, ( 2 ) improve t h e r e l i a b i l i t y
of t h e equipment by design and development, and ( 3 ) duplicate l e s s r e l i a b l e items so t h a t if they f a i l , spares will automatically continue t o perform t h e function. Many components in r e a c t o r power plants appear t o perform functions that 'are not necessary under the ground rules of t h i s study.
i s an overload protection device, which w i l l sometimes i n t e r r u p t service unnecessarily and hence i s an undesirable i t e m w i t h regard t o achievement of uninterrupted operation.
Where continuous power production i s t h e
only c r i t e r i o n by which performance i s judged, stoppage of a pump because t h e c i r c u i t breaker t r i p s i s a plant failure equivalent t o stoppage because t h e windings burn out. The r e l i a b i l i t y of components can be improved i n several ways. Modifications, such as s d b s t i t u t i o n of b e t t e r materials, use of b e t t e r methods of fabrication, t i g h t e r q u a l i t y control and inspection, and more extensive t e s t i n g of an item before acceptance, may increase t h e probable l i f e of an e x i s t i n g device.
I n addition, t h e change from normal require-
ments r e s u l t i n g from t h e s p e c i f i c objectives of the design may permit t h e use of a d i f f e r e n t type of device t o perform t h e required function.
pointed out i n t h e Gilbert Associates study' of r e a c t o r r e l i a b i l i t y , t h e achievement of r e l i a b i l i t y i s not consistent with a minimumecost philosophy. I n considering t h e use of spares, an important point t o recognize i s that improved r e l i a b i l i t y does not necessarily accompany duplication. A n i n s u f f i c i e n t l y r e l i a b l e control device f o r switching t o a duplicate
component, for example, could reduce t h e over-all r e l i a b i l i t y of t h e system.
Duplication of a component which might develop a leak doubles
t h e chances of t h a t type of failure. One way t o improve r e l i a b i l i t y , p a r t i c u l a r l y i n control equipment,
i s t o use coincidence c i r c u i t r y which requires, say, t h a t two out of three p a r a l l e l devices operate before t h e function i s performed.
this type of c i r c u i t r y , spurious operation of one device does not cause t h e function t o be performed, and failure of one device t o operate does not prevent performance of t h e function.
It i s worth noting t h a t a system of t h i s type works best under supervision s o that a component
which has f a i l e d can be repaired t o return the system t o i t s i n i t i a l capability.
In principle, however, t h e system can be mafie as elaborate
as required t o achieve t h e needed r e l i a b i l i t y .
It i s l i k e l y that r e l i a b i l i t y requirements on individual components can be relaxed somewhat i f advantage i s taken of t h e concept of operating spares.
For example, suppose that a system has two c i r c u l a t i n g pumps,
both operating a l l t h e time, but each having a capacity such that one alone will provide s u f f i c i e n t flow f o r reactor operation.
system the r e a c t o r i s not prevented from continuing operation by the
failure of one pump.
If each of these pumps has a r e l i a b i l i t y of 0.9,
then t h e probability, as computed from t h e following equation, i s 0.99 t h a t a t least one of t h e p a i r will s t i l l be operating at t h e end of a year: pg
- (l-p)n ,
where P i s the probability of s V a l of a t l e a s t one component i n t h e g group, p i s the p r f survival of an individual component, and n i s t h e number of f t h e r e a r e t h r e e pumps, each having a 0.9 r e l i
t y , then t h e p r o b a b i l i t y i s 0.999 t h a t a t l e a s t one
will be operating
he end of a yeax. of a r e a c t o r system has a very program.
As s t a t e d i n t h e Introduction,
this study presumes there i s need for a reliable r e a c t o r t o be i n t h e
f i e l d i n about four years.
If a prototype reactor i s t o be operated
before the f i r s t f i e l d i n s t a l l a t i o n i s made, and this seems essential, four years i s a f a i r l y short time f o r t h e program.
The time required
f o r construction of a s m a ~reactor, excluding development time, i s indicated by the experience with three reactors i n the 1-Mw ( e l e c t r i c a l ) range :
1. The SMCl (APPR, pressurized-water reactor b u i l t by Alco Products at Fort Belvoir) was c r i t i c a l 29 months a f t e r a w a r d of the contract.
conceptual design, which included a f a i r l y d e t a i l e d description of t h e core, was available before t h e contract w a s awarded.
The PM-2A (pressurized-water reactor f o r polar region, also
b u i l t by Alco Products) was assembled i n t h e manufacturer's plant 14 months a f t e r award of t h e contract.
A f a i r amount of design work had
been done before t h e contract date.
This reactor i s a skid-mounted air-
transportable adaptation of SM-1.
The PM-1 (pressurized-water reactor being b u i l t by Martin Company
f o r location a t Smaance, Wyoming), as presently scheduled, w i l l require about 27 months from t h e beginning of a parametric study t o assembly of t h e reactor,
About s i x months of this was required f o r f i n a l design,
and about 15 months i s estimated f o r the period from beginning of order placement t o assembly of t h e system. SM-1,
The P M d i s an advance f r o m t h e
and it uses a new f u e l element design.
Fuel element development
and physics studies began considerably i n advance of the 27-month period
referred t o above.
I n t h e last two cases, it should be noted that t h e times given a r e f o r assembly of t h e reactor. They do not allow f o r t e s t i n g before going t o power. If it i s presumed t h a t a program i s t o be undertaken with some urgency and a f a i r l y l i b e r a l budget, based on t h e figures f o r t h e three reactors above and on discussions with reactor manufacturers, the time required f o r design and construction of a prototype reactor ( b u i l t f o r high r e l i a b i l i t y ) appears t o be about two years, once the design concept
i s established.
Allowing six months f o r f i n i s h i n g t h e conceptual design
and making arrangements f o r t h e d e t a i l design and manufacture, t h e prototype might be read.y f o r t e s t i n g i n about two and one-half years, as shown by Fig. 1.
UNCLASSIFIED ORNL-LR-DWG SOl06A
DETAIL DESIGN iSTRUCTlON OF PROTOTYPE
OPERATION OF PROTOTYPE
CONSTRUCTION OF FIELD REACTCTOR
Time Schedule f o r Achievement of F i e l d Reactor i n Four Y e a r s .
The design of t h e f i r s t r e a c t o r for f i e l d use might begin before completion of t h e prototype, and a l l orders f o r equipment f o r it placed a f t e r one-half year of t e s t operation.
Allowance of one year for con-
s t r u c t i o n of the f i e l d r e a c t o r would bring i t s completion time t o four years from t h e beginning of t h e program.
T h i s schedule has no provision f o r developmental work other than that which proceeds concurrently with t h e design and construction of
t h e prototype.
There i s no contingency allowance f o r d i f f i c u l t i e s i n
t h e development of t h e prototype, nor i s time permitted f o r t h e solution of t h e problems found during operation of t h e prototype r e a c t o r before construction of t h e second
c t o r commences.
Hence, even with very
l i t t l e development required and with good luck a t every step, about four years would be required t o g e t t h e f i r s t r e a c t o r i n the f i e l d .
If t h e r e
were delays i n administrative decisions, construction holdups, or
unforeseen development d i f f i c u l t i e s , well over f o u r years might expire before a r e l i a b l e power plant were achieved. I n view of the above, a development program based on t h e following general principles appears t o have t h e g r e a t e s t promise of success i n producing a reactor t o satisfy t h e r e l i a b i l i t y requirements i n a period of about f o u r years:
1. The main e f f o r t should be devoted t o perfection of a system based on concepts with which there has been favorable experience. T h i s r e s t r i c t i o n r e f e r s not only t o reactor and power plant types i n the broad
sense, but a l s o t o t h e manner i n which t h e systems a r e t o be operated. 2.
Approaches based on unproven concepts should be studied as
secondasy programs i f they o f f e r promise of giving a b e t t e r system than that based on proven concepts.
3. Exceptions t o points 1 and 2 could be made f o r p a r t s of t h e system i f a switch i n midprogram t o a proven concept w o u l d not delay t h e project.
The most promising approach t o the achievement of high r e l i a -
b i l i t y i s simplification.
The objectives of this study make simplifi-
cation of t h e reactor p a r t i c u l a r l y promising, since t h e system need not be repairable, have a long l i f e , follow a varying load, or'operate i n a populous area. 5.
A close interdependence of t h e system p a r t s i s permissible,
since i n any case a f a i l u r e would terminate t h e a b i l i t y of t h e system s
t o produce e l e c t r i c i t y .
Components should be overdesigned and underworked t o increase
t h e i r l i f e expectancy.
EX'ficiency should be sacrificed t o gain
Components which a r e r e l a t i v e l y unreliable can be used during
s t a r t u p i f they cannot a f f e c t t h e system once it i s in operation,
The individual components and the assembled system should be
t e s t e d thoroughly before the system i s l e f t unattended.
T h i s testing
should continue long enough t o pass t h e period where e a r l y f a i l u r e s r e s u l t i n g from manufacturing defects a r e l i k e l y t o occur.
The preceding p r i n c i p l e s can be summarized by saying that i n seeking t h e achievement of r e l i a b i l i t y i n t h e limited time available, one should attempt t o use proven concepts and present technology.
The success of a
program t o develop a r e l i a b l e reactor i n four years should not depend on
major advancements i n technology o r t h e perfection of new concepts.
more time were available, t h e r e s t r i c t i o n on new development would be relaxed, but it would be s t a t e d even more strongly f o r a shorter time allotment. The approach described i n t h i s report may not lead. t o t h e best system f o r unattended operation.
Nevertheless it appears t o be the
approach which has t h e highest likelihood of producing a r e l i a b l e r e a c t o r i n four years. SELECTION OF A REACTOR POWER PIANT Reactor Type Early i n t h i s study a comparison w a s made of a number of r e a c t o r concepts with regard t o t h e i r a p p l i c a b i l i t y t o an unattended system. These included t h e various liquid-fuel, organic-moderated,
and liquid-metal-cooled reactors, as well as t h e pressurized-water and boiling-water v a r i a t i o n s of light-water-moderated reactors.
t h e premises of t h e preceding section, t h e conclusion w a s reached that t h e extensive and generally favorable experience with water-cooled reactors recommends t h e m for t h i s application.
The p o t e n t i a l advantages
of other systems, such as high thermal efficiency, favorable f u e l o r neutron economy, low construction costs, and possibly l i g h t weight, a r e of secondary importance i n t h e framework of t h i s study. Once having selected a water-moderated r e a c t o r as t h e b a s i s f o r this study, a f u r t h e r decision between boiling-water and pressurized-water systems w a s required.
Althoughthese r e a c t o r types have many features
i n common, each has c e r t a i n advantages and disadvantages r e l a t i v e t o t h e The obvious advantage of t h e boiling-water system i s t h e elimination of several major items of equipment:. t h e primary heat exchanger,
the main coolant circulating pump, and the core pressurizer.
of the pressurized-water reactor is that there is no production of radiolytic gas. In contrast, the boiling-water reactors in operation at present produce radiolytic gas which is either vented to the atmosphere or recombined in a catalytic recombiner. Another advantage of the pressurized-water reactor is that reactor control is accomplished very simply, and there are no problems of stability. In addition, there has been much more experience with pressurized-water than with boiling-water systems. The heat exchanger on a pressurized-water reactor, although large and eqensive, can be made quite reliable if care is taken in the selection of materials and in fabrication. Pressurizers are not of themselves basically unreliable, but the pressure control systems sometimes are. It appears to be possible to design a system in which the control, as normally performed, is eliminated, thus making the pressurizer function reliable. The most questioned component is probably the primary coolant pump. Experience with canned-rotor pumps of the type needed indicates that an individual pump that is carefully tested and found to be sound will operate reliably well in excess of one year. In addition, the concept of operating spares is directly applicable to the primary coolant Pump. When there is an overpressure of hydrogen, radiolytically dissociated water in a pressurized-water system is recombined internally with no production of hydrogen and oxygen. All pressurized-water reactors are operated in this manner. Although, in principle, water decomposition in
a boiling-water reactor can be suppressed by increasing the pressure, operating at a high pH, and injecting hydrogen,' this mode of operation has not appeared attractive, and all boiling-water reactors in operation evolve radiolytic gas. Besides the problem of handling the radiolytic gas, 'the oxygen in the steam makes the materials problems more severe in the boiling-water reactor than in the secondary system of a pressurizedwater power plant. Highly enriched pressurized-water reactors can operate, and often do (the SM-1 always), without automatic reactivity regulation, except 14
f o r the inherent action of t h e negative temperature coefficient.
discussed l a t e r , it appears t o be f e a s i b l e t o operate a pressurizedwater system f o r one year with no movement of absorbing materials i n the core.
This can be done using only proven concepts.
Boiling-water r e a c t o r s
normally a r e operated with continuous r e a c t i v i t y regulation by movement Long-term operation without t h e use of an auto-
of absorbing materials.
matic control system would depend on more of an innovation than i n t h e case of a pressurized-water reactor.
(Boiling-water reactor control
might be considerably simplified, however, a t steady power. ) Because it appears possible t o make a simple pressurized-water r e a c t o r with few innovations, t h e pressurized-water concept w a s made the basis f o r this study.
T h i s selection w a s not made because t h e pressurized-
water r e a c t o r i s inherently more r e l i a b l e than t h e boiling-water reactor, but because t h e r e i s more reason f o r confidence t h a t a r e l i a b l e , unattended, pressurized-water r e a c t o r can be developed i n four years.
There has been
extensive experience with pressurized-water reactors, and t h e method of operation proposed here deviates l i t t l e from t h e manner i n which they have been operated i n the past. Power Recovery System The s e l e c t i o n of a power recovery system t o operate i n conjunction with the r e a c t o r plant w a s heavily influenced by the principles outlined
i n t h e discussion on r e l i a b i l i t y .
During t h e i n i t i a l phases of t h i s
study, consideration w a s given t o t h e p o s s i b i l i t y of converting heat
energy t o e l e c t r i c a l power through other than t h e use of a conventional turbine-generator system.
I n p a r t i c u l a r , a b r i e f study w a s made of
lusion of t h e study was that, i n
view of t h e present e a r l y st
evelopment of thermoelectric materials
and the r e l a t i v e l y r e s t r i c t e d amount of operating experience, thermo-
e l e c t r i c generation does not appear t o be promising f o r use a t t h i s time.5 The development of improved materials permitting higher e f f i c i e n c i e s at moderate teLperatures may i n t h e f u t u r e make such systems a t t r a c t i v e f o r unattended power stations.
With the selection of a pressurized-water reactor and application of the stated reliability concepts, it is clear that the most suitable power system is one employing a conventional steam turbine-generator set using water as the working fluid for the heat-power cycle. It is appreciated that many of the corrosion problems associated with water systems might be avoided with other fluids, but the technology and experience with equipment using less corrosive fluids are not adequate for
a reliable design. CONCEPT OF A SIMPLIFIED pRESSURIZED-WA!TER ,Bl3ACTOR Following the selection of the pressurized-water concept a critical exadnation of pressurized-water reactors was required to determine the modifications necessary to achieve the required reliability. The advantages of using proven concepts and present technology were balanced against the gain to be made by a change in design and the likelihood of the change being successful. Some components and procedures used on SM-1,
for example, appear to be directly applicable to an unattended reactor. In other cases, modifications are needed, and in a few instances the development of new components would be desirable. Wherever innovations are suggested in this report, there are alternate solutions which could be used should the new development not appear to be progressing rapidly enough. All the new components proposed represent combinations of items on which there has been successful experience. In the subsections which follow, the features of a pressurized-water reactor that affect its reliability are discussed. This section does not present a reactor design, but it does suggest concepts and techniques on which a design could be based. A schematic diagram of a simplified pressurized-water reactor is presented in Fig. 2 to indicate the type of system which emerges from the discussion. All the components that function during unattended plant operation are included.
UNCLASSIFIED ORNL-LR-DWG 50407A
Fig. 2. Simplified Pressurized-Water Reactor With Hermetically Sealed Primary and Secondary Systems. Primarv Svstem The basic heat generation and removal system of a pressurized-water reactor is relatively simple, but the auxiliary equipment associated with it is often complex, The major complexities are in the reactor control system, and, in some cases, the water-treatment system. It appears possible, however, to satisfy the present control requirements with a very simple system based on the inherent self-regulation characteristics of a highly-enriched pressurized-water reactor. Water treatment does not appear to be needed in a hermetically sealed primary system to be operated for one year. The use of a hermetically sealed system with inherent nuclear control reduces the p r i w j system to essentially the core, heat exchanger, coolant circulating pump, and pressurizer. Various featyres of the primary system are discussed below. Reactor Control The action which controls the fissioning rate in a nuclear reactor may be broken down into three functions. These are regulation, shim, and safety. In a pressurized-water reactor the regulation function maintains the primary water at a constant temperature despite small
changes in reactivity, such as those associated with changes in load. S u m control performs the same type of compensation for the large reactivity changes associated with fuel depletion and accumulation of fission-product poisons. The nuclear safety system stops the chain reaction in the event of malfunction of the reactor system. Interrelated with these functions are the scheduled actions which start up and shut down the reactor. The regulation function on a pressurized-water reactor is an inherent one. The negative temperature coefficient is of sufficient magnitude to control small changes in reactivity with only slight changes in the average core temperature. Long-term changes in reactivity during the life of a core may, however, be appreciable. The upper curve' of Fig. 3 illustrates the reactivity change which would occur in a fully-enriched
COMPUTE0 FOR PM-t (REF 6 1 ----COMPUTEDFOR SM-1 (REF 7 )
t.15 r *
10 15 eo CORE EURNUP (thermol Mw-yr)
Fig. 3 . Illustration of Effect of Burnable Poison on Reactivity Change of Pressurized-Water Reactor. f-
pressurized-water reactor if t h e r e were no compensation by t h e motion
of a shim rod.
The curve begins a f t e r the reactor has been heated from
room temperature t o i t s operating l e v e l and a f t e r t h e i n i t i a l r e a c t i v i t y has been reduced by t h e accumulation of xenon.
(The xenon poison
f r a c t i o n reaches about 90$ of i t s equilibrium value i n one day.) The addition of boron as a burnable poison serves t o reduce t h e excess r e a c t i v i t y during the e a r l y p a r t of t h e core l i f e , as i l l u s t r a t e d i n Fig. 3 .
The two lower s o l i d curves' of Fig. 3 a r e estimates of t h e
e f f e c t s on PM-1 of d i f f e r e n t amounts of lumped boron.
The dashed curve7
shows t h e r e a c t i v i t y change with burnup i n an SM-1-type core with uniformly d i s t r i b u t e d boron; t h e a n a l y t i c a l model used i n computing t h i s curve accurately predicted r e a c t i v i t y changes during the l i f e of Core I of SM-1. The dashed curve remains within a range of 1% Ak/k f o r a core l i f e of approximately 8 Mw-yr (thermal), and the lower s o l i d curve remains within t h a t range for an even longer period.
One year of operation of a reactor
producing 1 Mw of e l e c t r i c i t y would r e s u l t i n core burnup of 7 t o 8 Mw-yr (thermal). One d i f f i c u l t y with t h e use of a burnable poison i s t h a t it tends t o reduce t h e r e a c t i v i t y l i f e t i m e ( t h e period during which kerf exceeds
1.0) of t h e core.
If a shorter r e a c t i v i t y l i f e of t h e core can be ac-
cepted, curves even f l a t t e r than those i n Fig. 3 can be obtained.
measures, such as using more than one absorbing element, closely t a i l o r i n g t h e poison location and concentration, and perhaps even t a i l o r i n g t h e f u e 1 , l o c a t i o n and concentration, may extend t h e period i n which t h e rea c t i v i t y remains f a i r l y constant.
The uncertainties i n t h e physics
calculations, however, become g r e a t e r as t h e system becomes more complex
o r as it deviates more from t h e region of experience. e coefficient of -2 x
(Ak/k)/OF F t o compensate f o r a 1s change i n ref i c i e n t of the SM-18 i s more negative l e poison r e s t r i c t e d t h e r e a c t i v i t y t o a he water temperature t o vary over
a range of 50째F would elimi
ed for mechanical control during
T h i s technique appears t o be f e a s i b l e as a means of
obtaining high r e l i a b i l i t y f o r a period of one year. The preceding discussion has been based on t h e use of f u l l y enriched fuel.
Use of low-enrichment uranium might appear a t t r a c t i v e .
p a r t i a l l y enriched element an excursion i s c u r t a i l e d by t h e Doppler e f f e c t i n t h e fuel, whereas i n f u l l y enriched elements t h e r e i s dependence on t h e change i n moderator density.
Actually, a strong f u e l temperature
c o e f f i c i e n t i s undesirable with regard t o achievement of inherent control.
The proven low-enrichment f u e l elements f o r power r e a c t o r s con-
t a i n bulk U02, as discussed i n t h e next section.
The contact between
UO;! and t h e cladding i s such t h a t there may be an appreciable temperature drop between them, and t h e thermal conductivity of t h e gas i n t h e gap changes s i g n i f i c a n t l y upon t h e evolution of xenon and krypton (unless t h e element i s i n i t i a l l y f i l l e d with xenon).
I n addition t h e e f f e c t i v e
thermal conductivity of t h e UOz may possibly change with time because of
cracking and f u e l burnup.
The UO, temperature would increase during t h e
l i f e of t h e core from these e f f e c t s .
The uncertainty i n t h e temperature
change would make t h e increase i n resonance absorption unpredictable, and t h i s would make it more d i f f i c u l t t o l i m i t t h e r e a c t i v i t y change t o a specified range.
For this reason f u l l y enriched uranium appears t o be
t h e more a t t r a c t i v e f u e l f o r a reactor which does not have an automatic control system.
It should be noted t h a t t h e system proposed will be s t a b l e with respect t o xenon "transients."
A small reduction i n e l e c t r i c a l load
would r e s u l t i n an i n i t i a l decrease i n neutron flux, followed by a t r a n s i t i o n a l increase i n xenon poisoning.
The increased poisoning
would lower t h e reactor temperature and reduce t h e temperature driving force f o r heat t r a n s f e r t o t h e secondary system. continue i f t h e r e were no corrective action.
This sequence would
However, i n a system with
a fixed e l e c t r i c a l load, t h e thermal power would a c t u d l y increase at
t h e lower reactor temperature, since t h e lowered thermal efficiency would r e s u l t i n a g r e a t e r heat load f o r t h e same e l e c t r i c a l power.
The f l u x
would thus tend t o increase and s t a b i l i z e t h e system. -,
T h i s inherent s t a b i l i t y may not e x i s t i f , as discussed l a t e r , t h e
e l e c t r i c a l load i s automatically varied t o control the frequency.
control obtained by varying t h e e l e c t r i c a l load, a reduction i n reactor temperature would be followed by a reduction i n e l e c t r i c a l load and probably a reduction i n t h e thermal load on the reactor.
While a system
of t h i s type might be stable, this can only be determined by study of t h e p a r t i c u l a r case. There a r e t h r e e areas of concern r e l a t i v e t o a reactor with inherent control:
(1)t h e effect of core design on the temperature coefficient,
( 2 ) t h e uncertainties i n judging t h e proper f u e l and poison loading, and
( 3 ) t h e d i f f i c u l t y of accurately controlling t h e boron loading i n a f u e l element.
Changes i n size, shape, fuel-to-moderator r a t i o , r e f l e c t o r
thickness, poison concentration, etc., may change t h e temperature coe f f i c i e n t of r e a c t i v i t y .
I n t h e design of a self-regulating core,
achieving a favorable temperature coefficient i s as important as limiting r e a c t i v i t y changes with time, and it should be studied as thoroughly. The problem of estimating r e a c t i v i t y may, f o r convenience, be separated i n t o two parts, estimation of t h e i n i t i a l c r i t i c a l mass and estimation of t h e change i n r e a c t i v i t y w i t h burnup.
The f i r s t p a r t in-
volves an attempt t o make t h e r e a c t i v i t y a t t h e beginning of unattended operation equal 1.0 at t h e temperature desired; t h e second involves estimation of changes which are i n a range of about 0.01 k / k .
t i n c t i o n i s made between these two p a r t s of t h e problem because t h e cause a d effect of errors is different for each.
For example, a 3%
e r r o r i n t h e i n i t i a l r e a c t i v i t y might mean a 60째F e r r o r i n w a t e r temperature i f not corrected,
r e a s a 3$ e r r o r i n t h e change i n k / k
would have an i n s i g n i f i c a n t e f f e c t on operation of t h e reactor.
calculation of t h e f i r s t type i s determinable, however, and probably correctable before operation of t h e reactor.
An e r r o r of the second type
would not be known u n t i l a core had been operated. There i s an appreciable body of experience with cores of t h e general SM-1 type t h a t would be of value i n estimating t h e i n i t i a l r e a c t i v i t y ,
and c r i t i c a l experiments could be performed i f needed.
I n estimating
r e a c t i v i t y changes with burnup, the. calculational technique used would be normalized against any experimental d a t a t h a t a r e applicable.
has been l i t t l e experience i n this area, however, and t h e r e a r e uncert a i n t i e s associated with t h e calculations.
An extensive physics pro-
gram i s c l e a r l y called for, and it would be well i f the calculations could be checked against an experiment e a r l y i n t h e program.
l a r l y useful approach would be t o construct a core designed f o r inherent control and i n s t a l l it i n an existing reactor of t h e same general type. There w i l l be several suitable s m a l l reactors i n operation by t h e time a core could be ready f o r t e s t i n g .
I n any case, physics problems are not
l i k e l y t o delay achievement of a reactor with inherent control, since a core with c h a r a c t e r i s t i c s similar t o those of t h e dashed l i n e i n Fig. 3 could be used i f a core having b e t t e r r e a c t i v i t y c h a r a c t e r i s t i c s were not achieved. The preceding discussion of burnable poisons has been concerned with t h e calculation of t h e amount of poison required.
i s t h a t of insuring t h a t t h e desired amount of burnable poison i s in-
cluded i n the f u e l element.
T h i s i s discussed l a t e r under t h e section
on f u e l elements. Even i f burnable poisons a r e used f o r r e s t r i c t i n g long-term r e a c t i v i t y changes, some method must be provided f o r insuring that t h e core i s subc r i t i c a l before it i s brought t o t h e operating temperature.
I n addition,
compensation must be made f o r t h e r e a c t i v i t y change which r e s u l t s from t h e i n i t i a l accumulation of reactor poisons.
Thus there must be a con-
t r o l system f o r bringing the reactor t o t h e condition i n which t h e r e a c t i v i t y i s on the f l a t p a r t of t h e reactivity-burnup curve.
poison rods perform t h i s function during t h e s t a r t u p period would not reduce t h e r e l i a b i l i t y of t h e system, since t h e rods would not p a r t i c i p a t e i n t h e operation of t h e reactor when it i s unattended.
The system f o r
providing controlled movement can be quite simple, since rapid i n s e r t i o n would be under the force of gravity. There a r e absorber materials which have adequate r e l i a b i l i t y f o r use i n t h e control rods of an unattended reactor.
Unclad hafnium would
s a t i s f y t h e requirements, and it i s l i k e l y that a matrix of europium oxide i n s t a i n l e s s s t e e l would suffice. t e s t e d i n t h e SM-1 a t present.)
(Europium elements a r e being
Other materials may a l s o be satisfactory,
since, being normally withdrawn, t h e rods would undergo l i t t l e burnup. The inclusion of poison rods would be advantageous f o r another reason.
If l a t e i n t h e development program it were decided t h a t t h e in-
herent control system were not desirable, o r i f the objectives f o r t h e system changed, t h e rods would be available f o r s h i m control with no appreciable change i n reactor design.
The rods could be used t o change
r e a c t i v i t y during t h e l i f e of t h e core i f a means were provided f o r moving them out at i n t e r v a l s by s m a l l amounts.
For example, a watt-hour
meter located on the e l e c t r i c a l output of t h e turbine could actuate a simple mechanism f o r moving t h e rods.
Alternatively, t h e rod movement
could be programed i n advance f o r a mechanism operated by a clock.
type of control would appear t o be much more r e l i a b l e than one which monitors t h e core temperature and attempts t o compensate continuously
f o r changes, but, t o operate as designed, a l l t h e rods would have t o move when directed.
If communication with the reactor were possible, the out-
put of a thermocouple could t e l l an operator when t o move t h e rods. A fast-acting safety system does not appear t o be required f o r a
low-power-density pressurized-water reactor, p a r t i c u l a r l y i n a remote location.
Probably t h e most hazardous time during the l i f e of the reactor
i s t h e s t a r t u p period.
During s t a r t u p t h e system would be under t h e con-
trol of an operator, and means could be provided for dropping the poison
rods on e i t h e r an automatic o r a manual signal.
Once t h e reactor w a s i n
operation, there would be l i t t l e chance of a rapid addition of r e a c t i v i t y , Power removal would be steady,
r e would be no pump s t a r t u p s o r
condition changes which might
t h e normal source of a cold-water
minated along with t h e operator.
Operator e r r o r would
I n general, when a water-cooled
r i s operating at a high power
level, it i s d i f f i c u l t t o add r e a c t i v i t y at a r a t e which would endanger t h e i n t e g r i t y of t h e system. Fromthe preceding discussion, there appears t o be no need for a fast-acting safety system, except perhaps during startup.
would have t o be provided, however, f o r shutting down t h e r e a c t o r at a s i g n a l from outside t h e system o r a t a specified time i n t h e l i f e of t h e core.
I n addition, rod i n s e r t i o n would be prescribed upon f a i l u r e of t h e
e l e c t r i c a l output from t h e generator (since i n any case this would signal t h a t t h e reactor w a s not operating as designed) and possibly upon high temperature o r high pressure.
There i s t h e p o s s i b i l i t y of a spurious
signal causing an &necessary shutdown, but a m u l t i p l i c i t y of independent c i r c u i t s (perhaps a three-out-of-five protective system more r e l i a b l e .
system) could be used t o make t h e
Aside from t h e automatic shutdown equip-
ment, t h e system proposed would depend s o l e l y on t h e inherent properties of t h e core f o r controlling t h e reactor when it was unattended. The problem of removing fission-product decay energy has not been
considered, but t h e design should be such that t h e r e would be adequate coolant flow by natural convection f o r d t e r h e a t removal.
r e j e c t i o n of heat t o t h e surroundings would depend on the s p e c i f i c application of t h e reactor. Fuel Elements There a r e s a t i s f a c t o r y f u e l elements f o r use a t t h e power d e n s i t i e s and temperature l e v e l s envisioned f o r a r e a c t o r of t h e type proposed. Core I of t h e SM-1, f o r example, appears t o have performed adequately f o r over two and one-half years with a burnup of almost twice t h a t required f o r t h e present application.
Although SM-1 elements with f u l l
exposure have not yet been examined i n a hot c e l l , t h e i r behavior (and examination a t t h e r e a c t o r s i t e ) indicates t h a t t h e r e have been no f a i l u r e s . A number of cores of t h e same general type (i.e.,
f l a t p l a t e s of
f u l l y enriched U02 i n a s t a i n l e s s s t e e l matrix clad with s t a i n l e s s s t e e l ) have been b u i l t , and t h e r e i s an established manufacturing c a p a b i l i t y f o r this design.
While t h e SM-1 design would be q u i t e adequate f o r t h e
present application, improved elements may r e s u l t from changes i n f a b r i cation methods and from modifications, such as t h e use of spherical oxide particles. Elements with bulk U02 contained i n s t a i n l e s s - s t e e l tubes a l s o appear t o give r e l i a b l e performance, as indicated by exhaustive t e s t s
Bulk U02 elements, however, a r e normally only p a r t i a l l y enrichei with U235. A s discussed above under the heading
and by experience with the PWR.
Reactor Control, p a r t i a l enrichment i s l e s s a t t r a c t i v e than f u l l enrichment f o r a reactor with inherent control. such as p l a t e s of Zircaloy-2-clad
Other types of f u e l element,
zirconium-uranium alloys, o r t h e
tubular s t a i n l e s s steel-U02 matrix elements developed by Martin f o r t h e PM-1,
might be satisfactory.
From a r e l i a b i l i t y viewpoint, however,
there appears t o be no necessity f o r use of other than t h e proven SM-1 type of element. There has been a problem with the inclusion of uniformly d i s t r i b u t e d c
boron i n SM-1-type f u e l elements.
During fabrication of the SM-1 core,
a large f r a c t i o n of the boron was l o s t from the f u e l .
ducted during the past two years has, however, revealed the mechanisms by which boron i s l o s t and pointed t h e way t o achieving b e t t e r control over the f i n a l concentration.
The boron l o s s during s i n t e r i n g of t h e
s t a i n l e s s stee14J02 meat can now be closely regulated, but the l o s s on f a b r i c a t i o n of t h e f u e l p l a t e i s l e s s controllable o r predictable.
present, t h e f i n a l boron content of a f u e l p l a t e would probably be within
10% of t h e specified value.
Work i s continuing on t h i s problem, and
improvement i n control of the boron content of the f u e l element i s t o be expected.
(For lumped poison, quite accurate control of t h e boron con-
t e n t could be achieved by adding machined s t r i p s o r rods of boroncontaining materials. ) The d e s i r a b i l i t y of using poisons other than boron would be determined by t h e r e a c t i v i t y advantage t o be obtained as balanced against uncertainties associated with t h e i r physical inclusion i n t h e reactor.
It might appear desirable t o use other fuels, f o r example, plutonium isotopes,1째 t o assist i n t h e control of r e a c t i v i t y during t h e l i f e of t h e core.
The technology of including plutcnium i n a f u e l element has
not been developed nearly so far as t h a t of uranium, however, and it i s therefore advisable a t present t o use only uranium f u e l .
Primary Coolant Pumps A s was discussed above, one advantage of a boiling-water r e a c t o r
would be t h e elimination of t h e primary c i r c u l a t i n g pumps.
a natural-circulation pressurized-water system might be used.
it appears t h a t experience with canned-rotor pumps of t h e s i z e needed f o r a 1.0-Mw ( e l e c t r i c a l ) r e a c t o r i s s u f f i c i e n t l y favorable that not much
advantage would be gained by eliminating them, p a r t i c u l a r l y i f t h e i r elimination would mean going t o a system with which t h e r e has been no experience.
Although one may point t o many examples of failures of l a r g e
canned-rotor pumps, a d e t a i l e d examination w i l l show t h a t most of t h e f a i l u r e s e i t h e r occurred e a r l y i n t h e l i f e of t h e pump, as a result o f manufacturing defects, o r they were a consequence of malfunctioning of other p a r t s of t h e system.
Once a p a r t i c u l a r canned-rotor pump has been
checked and found t o be good, there i s a high probability of i t s l a s t i n g a year and possibly even w e l l beyond a year.
As an example, both of the
core c i r c u l a t i n g pumps of t h e 3 - 1 have operated without f a i l u r e f o r a period of two and one-half years i n which they have been subjected t o many s t a r t s and stops (on one occasion a c i r c u i t breaker tripped without discernible cause) .ll
I n order t o insure t h a t t h e r e a r e no d e f e c t s i n
design and construction of t h e p a r t i c u l a r pumps t o be used, they should be t e s t e d i n a loop and then run f o r a period during t h e checkout of t h e assembled reactor. For added r e l i a b i l i t y , it would be possible t o employ two continuously operating pumps i n p a r a l l e l , with each having s u f f i c i e n t capacity so t h a t i f one f a i l e d t h e flow would s t i l l be high enough t o continue t o cool
t h e core.
Some type of simple check valve would be used t o avoid by-
passing of t h e core by r e c i r c u l a t i o n through an inoperative leg.
t h e pressure drop would be low and a perfect s e a l would not be needed,
it would not be d i f f i c u l t t o design a valve t o give t h e needed r e l i a bility.
The procedure when switching from one pump t o another i n t h e
SM-1 i s t o start t h e second pump, so as t o have both operating, and then t o cut off t h e f i r s t pump.
There have been no d i f f i c u l t i e s associated
with t h i s method of operation. l1
one pump w i l l produce a flow of about 3900
I n t h e case of the SM-1,
gpm through t h e core, and i f both pumps a r e operating, t h e flow i s 4600 Hence, i n t h i s system, t h e flow would decrease only about 15$
A careful selection of pump c h a r a c t e r i s t i c s
upon t h e f a i l u r e of one pump.
and operating points may r e s u l t i n an even smaller flow change, although t h e above appears t o be satisfactory. Core Pressurizer The pressurizer i n a reactor primary system keeps t h e pressure high enough t o prevent boiling i n t h e core.
T h i s function should be retained
i n t h e reactor design, since there has been l i t t l e experience with e i t h e r l o c a l (subcooled) o r bulk boiling i n pressurized-water systems, and reactor control would be more d i f f i c u l t i f boiling were permitted.
addition t o preventing boiling i n the system, t h e pressurizer normally a c t s as a surge tank t o accommodate changes i n t h e volume of t h e primary coolant.
The pressurizer will have t o be designed t o handle appreciable
volume changes i n a system where water temperature v a r i a t i o n i s used f o r reactor control.
( A 50Â°F temperature change causes about a 5% change i n
l i q u i d volume. ) While there have been designs f o r self-pressurized reactors (which allow boiling near t h e core e x i t ) and f o r pressurizers using nuclear heat, t h e e l e c t r i c a l l y heated pressurizer should be retained because of i t s proven capability.
The major l i m i t a t i o n on the r e l i a b i l i t y of
pressurizers i n existing systems i s t h a t associated with the device which controls t h e pressure.
Normally an instrument senses t h e pressure
i n t h e system and uses a contr The r e l i a b i l i t y of t h
t u r n t h e heaters off and on. i z e r system would be increased i f t h e
necessity for sensing t h e
essure and f o r switching t h e heaters
on and off could be eliminated
system t h a t requires no control o r
sensing of pressure has been d the electrical
t o permit continuous operation of conditions i n the primary system,
the pressurizer heaters can be carefully sized so t h a t a t equilibrium conditions t h e temperature of t h e pressurizer i s at t h e design level.
Such a heat balance is obviously precarious in that the temperature of the pressurizer is a function both of the electrical heat input and the heat transfer coefficient. In the system described, the temperature difference between the pressurizer and ambient might be 400"F, and thus
a 10% change in heat generation or heat transfer coefficient would result in a temperature change of 40째F (equivalent to perhaps 4-00 psi). If the temperature drop from the pressurizer to the heat sink were small (say 50"F), the presumed 10s change in temperature differential would only amount to a few degrees. By insulating the pressurizer from the ambient and providing within the pressurizer a heat sink cooled by the exit water from the reactor core, the pressurizer-to-sink temperature differential and thus the absolute change in pressurizer temperature required to accommodate small changes in heat transfer conditions can be sharply reduced. Since it is the temperature difference between core and pressurizer that prevents boiling, the control of this difference will produce satisfactory operation. A pressurizer design based on the above principle of maintaining
the pressurizer temperature a specific amount above the core exit water temperature by using a fixed heat input would require little developmental
work, and the system should operate reliably over the life of the reactor. A system of the type described is illustrated in Fig. 4. The heat input to the water in the pressurizer is provided through a number of parallel electric heaters, each operating independently of the other and separately fused, if necessary. If there were 20 independent heaters, loss of the heat output of several would not jeopardize operation of the reactor.
Steam generated by the heaters would be condensed on tubes cooled with core-outlet water.
If the condenser tube wall were deliberately
made thick, the heat transfer resistance could be concentrated in the wall and would not be sensitive to changes in the heat-transfer coefficient on either side of the tube. One problem of this particular design is the interference of hydrogen (which would be present in the pressurizer) with the condensation process. If this appears to be a serious problem, it may be possible to keep the hydrogen from accumulating by connecting
ORNL-LR-DWG 504098 ,INSULATION
PRIMARY COOLING PUMP W
4. A Self-Regulating Pressurizer.
a small vent l i n e from t h e t o p of t h e pressurizer back t o some lower-
pressure point i n the primary system. There a r e variations of this concept which perhaps would work as well o r b e t t e r than the one i l l u s t r a t e d i n Fig. 4.
One p o s s i b i l i t y i s
t o provide f o r the vapor generated i n t h e pressurizer t o condense d i r e c t l y on a f r e e surface of water from t h e core.
Another i s t o l o c a t e a heat-
removal c o i l i n t h e l i q u i d volume of t h e pressurizer r a t h e r than i n t h e gas space.
Heat t r a n s f e r f r o m t h e e l e c t r i c heaters t o t h e coolant would
be by natural c i r c u l a t i o n of t h e water i n t h e pressurizer.
would not depend upon condensation and would not be affected by gas
coefficients might be sensitive
accumulation, but t h e nat t o changes i n w
the natural-convection coefficient
One advantage of having
off and thus tends t o stabalize essurizer and t h e primary system.
t tends t o increase as t h e heat educe t h e temperature difference ecreases . signs could be made t o work quite r e l i a b l y .
However, i f attempts t o avoid pressurizer control a r e
not successful, an automatic control system using duplication o r coinci-
dence c i r c u i t r y could be made t o be r e l i a b l e . Main Heat Exchanger Since t h e heat exchanger has no moving p a r t s and no control equipment, f o r this unit, r e l i a b i l i t y i s synonymous with leaktightness.
order t o assure leaktightness, not only i n i t i a l l y but f o r t h e desired l i f e , t h e following procedure would be followed:
1. A material would be selected t h a t i s e a s i l y fabricated, has good corrosion resistance, and i s not subject t o stress-corrosion cracking. Inconel appears t o be such a material.
I t s use i s discussed more
thoroughly under t h e heading Water Treatment.
2. The heat exchanger would be c a r e f u l l y designed with p a r t i c u l a r emphasis on t h e tube-to-header
Any c o n f l i c t between economy and
r e l i a b i l i t y would be resolved i n favor of r e l i a b i l i t y . 3.
There would be close inspection throughout f a b r i c a t i o n .
The completed exchanger would be thoroughly checked f o r leaks.
The heat exchanger would be i n s t a l l e d i n a loop and t e s t e d at
r e a c t o r conditions f o r an extended period. 6.
After loop operation, t h e exchanger would again be inspected
and t e s t e d f o r leaktightness. A heat exchanger constructed and t e s t e d as described should have a
very high r e l i a b i l i t y f o r many years of operation. Water Treatment The presence of excess hydrogen i n t h e primary system i s desirab1el2=l4 p r i n c i p a l l y because t h e r a d i o l y t i c oxygen content i s then held
a t very low (often undetectable) l e v e l s by t h e radiation-induced waterrecombinat ion r e a c t ion
* â€˜ I I
O w i n g t o t h e sealed condition of t h e system, this excess hydrogen cannot
be consumed by atmospheric oxygen and nitrogen (which normally would be introduced with makeup water), and it w i l l not be l o s t by leakage, except possibly by diffusion through t h e container w a l l s .
w i l l be produced by t h e over-all corrosion reactions of metals such as i r o n and chromium3 e.g.,
If it i s assumed that t h e primary system i s p r i n c i p a l l y s t a i n l e s s s t e e l and that the average corrosion r a t e i s 5 mg/dm20mo,â€™5 it i s estimated t h a t approximately 30 standard l i t e r s (1f t 3 ) of hydrogen w i l l be produced per 100 f t 2of surface i n one year of operation.
The r e s u l t i n g gaseous
hydrogen pressure i n t h e pressurizer (of t h e order of a few p s i ) would not be excessive. Other primary water treatment methods used i n conventional pressurizedwater r e a c t o r s are pH control (usually i n t h e range 9 t o 11) and sidestream water cleanup by f i l t r a t i o n and/or mixed-bed ion exchange.12-13 * l6 The need f o r these water-treatment procedures i n t h e present system w i l l be determined mainly by t h e degree of heat-transfer surface fouling by transportable corrosion-produce s o l i d s (crud).
primary systems, 12-13116e18 high pH ( 9 t o 11) has a d i s t i n c t l y b e n e f i c i a l e f f e c t i n reduc coolant.
It appears t h a t i n s t a i n l e s s - s t e e l and carbon-steel
a t e of corrosion
primary system (pH of 10 t o 10.5) i n which t h e fuel-element surface i s one-fifth t h e t o t a l area would r e s u l t i n a fouling r a t e of 10 mg/dm*-mo. I n a year of operation t h i s would give a deposit averaging 4.00017 in. i n thickness.
On t h e same basis, a t a neutral pH, a deposit of t h e order
of 0.001 i n . i n thickness (estimated from the e f f e c t of pH on t h e release r a t e from carbon s t e e l ) would r e s u l t . Thus it seems t h a t a basic pH i n t h e primary system i s desirable, even though t h e over-all fouling problem i s not so severe that it would be disabling i n a system w i t h moderate heat f l u x .
The pH of the coolant
could be established a t t h e outset by the addition of l i t h i u m hydroxide
o r ammonia.
The s t a b i l i t y of pH with time ( i f ion-exchange control i s
absent) i n such a sealed system i s d i f f i c u l t t o predict; however, since there a r e no base-consuming processes at once apparent, it i s expected t h a t the pH w i l l remain s u f f i c i e n t l y above t h e n e u t r a l point t o be dist i n c t l y beneficial from the standpoint of fouling. Cleanup.
Side-stream cleanup by mixed-bed ion-exchange r e s i n s i n
conventional pressurized-water systems has three functions:
(1)t o re-
move soluble water-borne a c t i v i t y , ( 2 ) t o f i l t e r out insoluble waterborne corrosion solids i n order t o reduce crud deposition, and ( 3 ) t o control pH by introducing t h e cation-exchange r e s i n i n the lithium,
In t h e present system, cleanup of waterborne a c t i v i t y i s not a primary consideration. Removal of water-borne ammonium, o r hydrogen form.
crud probably will not be an important consideration i n view of t h e expected low fouling r a t e s a t high pH, especially since side-stream processing may not g r e a t l y reduce fouling i n any case."
Finally, t h e added pH
control afforded by a mixed-bed ion exchanger will be needed only i f some unforeseen base-consuming process occurs i n t h e system.
Thus it seems
quite possible that a side-stream ion exchanger can be omitted. Since t h e crud-removal r a t e could be greater, an on-stream cleanup system would o f f e r promise of g r e a t e r reduction i n fouling than a sidestream cleanup system.
Such a system, however, must i n no way endanger
r e l i a b l e operation of t h e primary system.
of such cleanup could
be based on t h e f a c t t h a t the principal constituents of t h e crud will be ..-.
magnetic oxides of i r o n and ~hromium.'~ It seems possible t h a t on-stream
removal of crud, i f needed, could be accomplished by using a magnetic c o l l e c t o r i n a region of low flow r a t e , To summarize t h i s discussion of water treatment, it appears l i k e l y t h a t a hermetically sealed primary system would operate s a t i s f a c t o r i l y f o r a year without any treatment other than t h e establishment of desirable i n i t i a l conditions.
The preferable i n i t i a l condition appears t o be the
use of de-ionized water which i s adjusted t o a high pH and which contains dissolved hydrogen.
A research program involving high-pressure loops
would help determine t h e conditions which a r e favorable f o r operation without water treatment.
Operation of existing pressurized-water reactors
without continuous purification might provide valuable d a t a on crud transport and deposition.
If f u r t h e r study indicates t h a t side-stream
p u r i f i c a t i o n (or possibly on-stream cleanup by magnetic collection) i s desirable, i t s provision would have l i t t l e e f f e c t on system r e l i a b i l i t y . Secondary System The decision t o employ a conventional steam secondary system f o r power generation w a s based on t h e high degree t o which such systems have been developed.
It w a s recognized, however, that t h e conventional steam
cycle will have t o be simplified and improved i f t h e a b i l i t y t o operate unattended f o r one year i s t o be achieved. Three concepts of t h e steam system a r e considered i n t h i s study:
(1)a "simplified conventional" steam cycle from which feedwater heaters, hot-well pump, b o i l e r blowdown, and nated; ( 2 ) a "semiconventional s a moisture-recove
s h a f t seal; and (3
loss concepts w i l l be
enting of noncondensibles a r e elimiwhich has t h e preceding simplifit h a t avoids net l o s s of water from i c a l l y sealed steam system i n
he system i s positively eliminated.
Each of these
sed i n t u r n .
Simplified Conventional System Feedwater heaters a r e excluded from t h e power system, since thermal efficiency i s of secondary importance.
The hot-well pump i s eliminated
by using a multistage b o i l e r feed pump t o perform both pumping functions. By selection of a pump which i s sensitive t o suction head, t h e l e v e l i n the hot well, and, consequently, t h e l e v e l i n t h e s t e m generator, can be controlled automatically. Boiler blowdown can be made unnecessary f o r one year of operation by using a material such as Inconel f o r t h e steam generator and pretreating the water supplied t o t h e system.
With t h e elimination of blowdown,
makeup water i s needed only t o replace t h a t l o s t from t h e shaft seals. A water cleanup system does not appear t o be necessary, although i t s pro-
vision i n a side stream would not reduce t h e r e l i a b i l i t y of t h e power plant.
Ejection of noncondensible gases may not be required, since t h e
amount of gas accumulated in one year is estimated to be small enough
t h a t it could be accommodated by proper condenser design. A brushless type of generator would eliminate t h e frequent maintenance
required f o r brushes and commutators.
The turbine admission valve would
be t h e only valve required t o operate a f t e r s t a r t u p of t h e plant.
system could be of all-welded construction, with bellows-sealed valves. It would be leaktight everywhere except a t t h e shaft seal, and water
l o s s e s from t h e shaft s e a l could be accommodated by having a large water capacity i n the system o r by providing a makeup storage tank. The steam pressure would be r e l a t i v e l y low, perhaps 300 o r 4.00 psia, and no superheating would be required.
With proper turbine design the
steam path would t o l e r a t e t h e higher moisture content of the low-pressure stages.
Since efficiency i s not of major importance, a r e l a t i v e l y high
condensing pressure could be used i f made desirable by t h e nature of t h e heat sink and by moisture considerations i n t h e turbine. The equipment required f o r t h i s cycle i s a l l within the realm of present technology and can be manufactured with l i t t l e o r no development required.
The need t o replace l o s t water seriously handicaps t h i s type
of system, however, and precludes i t s recommendation f o r t h e p a r t i c u l a r application now under study.
Semiconventional Steam System
The simplified cycle described above may be further modified to eliminate the net loss of water by recovering the leakage fromthe turbine shaft seal and returning it to the system. Several schemes that were studied were considered to be workable. Perhaps the most practical of these is the use of a conventional bleed-off type of labyrinth seal that is vented to the main condenser, as illustrated in Fig. 5 . The seal would separate the steam at the tail end of the turbine from the gas in the containment vessel. The vessel could be initially filled with inert gas at a pressure greater than that in the condenser so that a net inleakage to the system would occur. Noncondensible accumulations in the condenser would be withdrawn by a steam-jet ejector, passed over an after-condenser to reduce the moisture content, and discharged back into the containment vessel. Once equilibrium conditions were established, no net water loss from the system would result unless there were surfaces below the dew-point temperature in the containment space. If condensation were difficult to avoid, a cold surface could be provided with means f o r returning the condensate to the condenser through a hydrostatic leg. UNCLASSIFIED ORNL-LR-DWG 504!0A
T h i s concept permits use of an e s s e n t i a l l y standard turbine with a brushless-generator s e t and, yet, largely eliminates t h e makeup water problem.
The need f o r a steam e j e c t o r and after-condenser,
complications t o the system which are preferably avoided; thus considerat i o n was given t o a completely sealed system. Hermetically Sealed Steam System By using a canned turbine-generator t o eliminate t h e shaft s e a l s from t h e otherwise leaktight semiconventional plant, a hermetically sealed system which eliminates t h e water leakage problem can be achieved. The r o t o r of t h e generator would operate i n a water o r water-vapor atmosphere and would be cooled by c i r c u l a t i o n of cooling water through the stator.
Generator manufacturers and manufacturers of canned-rotor pumps
s t a t e t h a t generators of this type and of the r e q u i s i t e s i z e can be fabricated using presently existing technology. Operation of t h e turbine-generator as a sealed u n i t makes it desirable t o use water-lubricated bearings and t o eliminate an oil-actuated hydraulic-governing system.
Turbines with water-lubricated bearings
have been t e s t e d by several manufacturers, and water-lubricated bearings have been quite successful i n canned-rotor pumps.
There has been l i t t l e
experience with e i t h e r water-actuated hydraulic governors o r waterlubricated mechanical governors, but t h e r e appears t o be no major d i f f i c u l t y i n developing such equipment,
Both the governor and t h e admission
valve could be located i n t h e turbine housing.
Other a l t e r n a t i v e s t o
conventional-speed governors could be based on e l e c t r i c a l sensing of t h e generated frequency. Boiler feedwater could be used f o r bearing lubrication and generator cooling, since it would be cool and a t high pressure.
A s shown i n Fig. 2,
it could a l s o be used f o r cooling t h e primary c i r c u l a t i n g pumps. While the hermetically sealed. system poses advanced design problems t h a t require more development work than would be required for t h e two concepts previously described, t h e extension of present technology i s not great.
The hermetic s e a l concept promises a system of maximum simplicity
with a minimum number of components.
A f u r t h e r consideration i s t h a t ,
should a design and development program on t h i s type of system meet unexpected d i f f i c u l t i e s , the e f f o r t could be diverted t o the controlledThe components
leakage concept with l i t t l e delay t o the over-all program.
included i n t h e hermetic s e a l concept were shown i n Fig. 2. Basic Features of t h e Steam Cycle Special problems a r e associated with t h e achievement of r e l i a b i l i t y of some major components of the proposed steam system. cussed i n t h i s section. j
These a r e d i s -
I n addition, a discussion i s included on methods
of handling control functions and of eliminating t h e water-treatment problem. I n order t o assess t h e degree of r e l i a b i l i t y t o
be expected from conventional turbine-generators under normal conditions, Myers3 undertook a study of the operating h i s t o r y of seven u n i t s of 20t o 25-Mw rated capacity.
The operating h i s t o r y f o r these units, located
a t t h e Oak Ridge Gaseous Diffusion Plant, covers approximately a 14-year period.
The u n i t s a r e conventional and a r e operated i n a manner normal
f o r power plants.
Myers concluded that t h e probability of one of t h e
u n i t s operating continuously f o r one year w a s only 0.05.
I n an extension
of t h i s study by t h e authors, the operating h i s t o r i e s of three units of 3-Mw capacity were studied, and approximately t h e same order of r e l i a 9
b i l i t y was obtained.
(It should be noted t h a t t h e u n i t s referred t o i n
t h e above studies w e r e not constructed t o other than standard specifications.
I n addition, these p a r t i c u l a r u n i t s were b u i l t during World
War I1 and consequently a r e not only of
older design but undoubtedly
possess some compromises i n materials of construction.) A study of t h e maintenance work performed on t h e 3-Mw turbine-
generator units ind the exciter constit close second i n r e p a i r valve.
ed that replacement of t h e commutator brushes on t h e single most frequent cause of outages.
equency was repacking of the turbine admission
I n order t o determine the improvement t h a t might be realized by
t h e elimination of c e r t a i n repairs, a tabulation w a s prepared i n which
it was assumed that valve-packing maintenance and exciter-brush replacement could be eliminated. The elimination of these two repair items increased the probability of one year of successful operation from 0.05 to 0.60. These data may be somewhat misleading, since minor adjustments or repairs are often made without shutdown of the unit; there are wide differences in ppinion as to the effect of repairs &e
during operation. Such uncertainties in the meaning of operating data limit the value of
statistical information on component performance. Nevertheless, an examination of the faults which caused forced outages of the units showed
that the great majority were concerned with turbine auxiliary equipment and were not outages resulting from failures of basic components in the turbine-generator. Although it is not u n c o m n for turbine-generator units to operate continuously for a year without shutdown, experience in five nuclear plants, as reported by Gilbert Associates,' indicates that 30% of all forced station outages are due to the turbine-generator unit. Of these outages, the majority resulted from the exercise of protective controls to prevent equipment damage. Many of the other shutdowns were concerned with repairs to the brush and commutator systems. It appears that with the precautions of conservative design and the elimination of the brush problem, an oil-lubricated turbine-generator unit having a high probability of successful operation for one year-can be developed. As discussed elsewhere in this report, however, a waterlubricated turbine-generator suitable for use with the hermetic seal concept is a preferred design and can be developed in the allotted time.
Although it is probable that the brush life experienced on conventional generator units can be extended by improved design, the predictable reliability of brushes is insufficient to permit their use in an unattended plant. The use of slip rings might prove to be satisfactory, but generators of the brushless design would be preferable, and there is sufficient technology and experience with brushless generators to warrant their consideration. Three types of brushless generators
induction, rotating-rectifier, and inductor - are discussed below:
1. Induction Generator.
An induction motor may be operated as an
induction generator by driving it above synchronous speed t o obtain "negative s l i p . "
The operation of such a system on a network having a
lagging power f a c t o r requires t h e use of capacitance i n p a r a l l e l with t h e l o a d t o obtain a leading power f a c t o r .
A system of this nature has
inherently poor voltage control under conditions of varying reactive
load. A change i n load o r power f a c t o r causes a change i n t h e excitation current, which, i n turn, changes t h e output voltage and frequency.
correct f o r t h i s s h i f t , capacitance may be switched i n and out of t h e system as a function of load changes. i
For systems with a r e l a t i v e l y
constant load., it i s possible t o operate with a fixed capacitance and
s t i l l maintain reasonable voltage regulation.
Thus load swings of 10%
on a system with a unity power f a c t o r have been estimated t o change t h e voltage l e s s than 2% of full-load voltage.
External excitation may be
required during staxtup of an induction generator. Induction generator u n i t s i n s i z e s smaller than 1 M w have been constructed and operated successfully, and there has been extensive experience with canned induction motors on primmy pumps.
capacitors operated i n p a r a l l e l with t h e load should be inherently longlived and may be designed t o u t i l i z e a large number of small fused u n i t s
so t h a t f a i l u r e of any individual capacitor has l i t t l e e f f e c t on t h e system capacitance.
2. Rotating-Rectifier Generator.
synchronous type i s a l s o available. located on a shaft extensio titer i s f i t t e d with hermet
The r o t a t
The e x c i t e r f o r such a machine i s
t h e generator.
The armature of t h e ex-
l y sealed r e c t i f i e r s (also r o t a t i n g with t o d-c current f o r t h e main generator
t h e s h a f t ) t o convert t h e ndings.
A brushless generator of t h e
t i f i e r thus eliminates t h e need f o r
commutator brushe ed i n this type of generator have good capable of operating a t r e l a t i v e l y high high vibration and centrifugal ratings, and have been used extensively and successfully.
They a r e subject t o radiation damage,
but it should not be d i f f i c u l t t o protect them with shielding.
e x c i t e r s with r o t a t i n g r e c t i f i e r s have been b u i l t i n s i z e s up t o approximately 200 kw, and, currently, a t l e a s t one e x c i t e r of 1300-kw capacity i s under construction.
Another brushless type of generator con-
sidered f o r t h i s application i s t h e inductor generator. obtained from t h e s t a t o r windings.
Excitation i s
A s e r i e s of staggered "teeth" on t h e
r o t o r provide the path f o r t h e main pole flux, and a l t e r n a t i n g voltage
i s induced i n t h e main s t a t o r windings by v a r i a t i o n i n the reluctance of t h e air gap. windings.
R e c t i f i e r s supply t h e d-c current t o the excitation s t a t o r
Inductor generators have been b u i l t f o r frequeccies i n t h e
range of 1000 t o 10 000 cps.
The output voltage wave from an inductor
generator w i l l be influenced by a greater percentage of the higher harmonic components than i s normally present i n a synchronous machine, and this could a f f e c t t h e operation of other equipment i n t h e system.
experience with t h i s type of machine, p a r t i c u l a r l y f o r three-phase application and f o r lower frequencies, i s somewhat limited.
There i s therefore
some hesitance i n recommending it f o r t h i s application. The operation of e i t h e r an induction generator o r a synchronous generator appears t o be e n t i r e l y f e a s i b l e f o r one year of unattended operation.
Generator efficiency f o r these designs i s estimated t o be
above SO$. An important additional consideration i n t h e discussion of turbinegenerators i s t h e method of bearing lubrication.
generator units employ o i l lubrication, and t h e semiconventional system previously discussed does not preclude the use of an o i l system.
such a system normally implies o i l pumps, o i l f i l t e r s , and an o i l cooler, t h e o i l pump could be d i r e c t l y driven off t h e turbine shaft, and s e l f cleaning oversized f i l t e r s could be i n s t a l l e d i n p a r a l l e l t o preclude flow stoppage due t o clogging.
A s pointed out by Burns and Roe,*'
of t h e complexity-of a circulating-oil system might be avoided with properly designed bearings using o i l reservoirs and multiple r i n g s t o eliminate t h e need f o r forced-oil lubrication.
Oil s l i n g e r s o r
deflecting vanes could be mounted on the shaft to prevent oil leakage, and cooling could be provided in the oil reservoir. The operation of conventional turbine-generator units has demonstrated that the interchange of oil and water in the systems is essentially negligible, particularly if oil temperatures are high enough to prevent moisture condensation. Oil wiper rings effectively reduce leakage of the oil around the shaft. Water-vapor leakage to the oil system may be prevented by.the use of centrifugal oil seals or other techniques. Thus oil-water vapor interchange will not prove a limiting factor f o r one year of operation.
An investigation was made of the possibility of using water-lubricated bearings for the turbine-generator unit. Water-lubricated bearings have been employed successfully in canned-motor pumps, and some units have had motor capacities in excess of 1000 hp. Recently, steam-turbine units with water bearings have been built and operated successfully. Units of. up to approximately 750-hp capacity are currently under test and in operation.21a22 Water bearings use hard-surface materials, such as aluminum oxide, and thus are resistant to scoring by impurities. Bearings of both hydrostatic and hydrodynamic design have been tested. Water temperatures have been in the order of 10O0F, but somewhat higher temperatures are considered to be acceptable, and widely varying water supply pressures have been employed. In order to assess the probability that water-lubricated bearings could be successfully developed and applied to turbine-generator units within the scope of the ground rules for this study, discussions were held with manufacturers experienced with the operation of steam turbines designed with water-lubricated bearings. Discussions were a l s o held with manufacturers of canned-motor pumps employing water-lubricated bearings. It was the unanimous opinion of these consulted that water bemings suitable for this application could be developed within the time required and that this development did not represent any major extrapolation of present technology.
Direct coupling of turbine t o generator i s probably preferable t o
t h e use of a geared unit, since reduction gearing would introduce an additional component that would be subject t o failure.
The d i r e c t
coupling, however, requires a compromise between optimum turbine speed and optimum generator speed.
It i s desirable t h a t t h e steam flow path i n t h e turbine be'conservat i v e i n design v e l o c i t i e s . 1
Turbine clearances may be l a r g e r than i n
current practice t o insure against rubbing failures, although a small
l o s s i n efficiency will r e s u l t .
Protective coatings can be used on t h e
turbine blades where t h e moisture content of t h e steam i s high. Water Chemistry.
One of t h e s t r i k i n g complexities of a conventional
steam secondary system i s t h e equipment required f o r maintaining adequate
water p u r i t y i n t h e operating system.
It w a s f e l t t o be of fundamental
importance t o t h e success of an unattended reactor plant that this portion of the secondary system be simplified, and therefore a study w a s made of
t h e w a t e r chemistry of t h e secondary system.
The study of this aspect
of operation of t h e steam power cycle included consideration of t h e e f f e c t s of corrosion products and t h e necessity f o r steam-generator blowdown, the magnitude of water losses and makeup requirements, radiol y t i c gas formation and venting problems, use of chemical additives, and means of cleaning up the c i r c u l a t i n g stream. 1. Water Treatment, Makeup, and Blowdown.
A conventional steam
plant incorporates e i t h e r a continuous o r periodic blowdown of t h e s o l i d s from the steam generator.
Makeup water i s supplied t o replace water l o s t
i n blowdown, from vents, and i n system leakage.
Since the unattended
reactor i n s t a l l a t i o n must be e n t i r e l y self-contained,
t h e need f o r sub-
s t a n t i a l quantities of makeup water would require e i t h e r a large storage f a c i l i t y , a water-recovery system, o r a processing system f o r providing makeup water.
The plant must be capable of one year of operation, and
therefore even nominal makeup requirements would assume large proportions. Furthermore, a system f o r supplying o r processing makeup w a t e r normally requires sensors and controls, and such components would decrease overa l l system r e l i a b i l i t y .
I n conventional steam cycles the principal corrosive
conditions a r i s e f r o m t h e presence of oxygen and carbon dioxide i n the feedwater and from t h e leakage of these atmospheric gases i n t o t h e conI
denser system.23 The usual treatments a r e deaeration of the feedwater; t h e addition of oxygen scavengers, such as hydrazine o r sodium s u l f i t e ; and t h e use of v o l a t i l e basic compounds, such as ammonia, organic amines, o r morpholine, t o react with carbon dioxide and t o increase the pH of
I n t h e present system, oxygen and carbon dioxide i n i t i a l l y
t h e condensate.
present could be removed by purging and using s u i t a b l e scavengers.
r a d i o l y t i c oxygen i s not produced i n appreciable amounts, water control i n a closed system should be considerably l e s s d i f f i c u l t than i n a conventional system. The operating r e l i a b i l i t y of t h e secondary system i s c r i t i c a l l y affected by t h e accumulation of corrosion-produce s o l i d s i n the steam generator.
I n conventional systems, corrosion-product production and
accumulations are minimized by maintaining a high pH i n t h e b o i l e r (with caustic or basic phosphate s a l t s ) and i n the condenser system (with v o l a t i l e amines) and often by the use of filming amines (10-18 carbon a l k y l amines). blowdown. I
Accumulated s o l i d s are periodically removed by b o i l e r
Since blowdown i s undesirable i n the present system because
of makeup-water requirements, consideration w a s given t o holding s o l i d s accumulation i n t h e steam generator t o within acceptable limits by (1) t h e choice of suitably corrosion-resistant materials, ( 2 ) t h e use of permanent water additives, and ( 3 ) t h e use of demineralizers o r other water cleanup methods. he production r a t e of radiotem w i l l be much lower than i n t h e primary system, but t h
ination r a t e probably w i l l
a l s o be less because these
accumulate i n t h e vapor
The oxygen produc
from t h e reactor flux w i l l i n t h e primary water.
been reported f o r t h e Idaho
steam generator well shielded o the radiation from N16 1 of 100 pc/ml (which has he gamma energy production
rate is approximately 3 x lo7 Mev/sec*ml. If this were totally absorbed by the secondary water [taking the
as 0.2],25 approximately
6 X lo1' molecules of oxygen would be produced per second per milliliter of secondary water. If the back reaction due to recombination is ignored, this could amount to an accumulation of several milliliters of oxygen per kilogram of secondary water in a year. T h i s estimate is, of course, high, since not all the gamma energy
w i l l be absorbed by the secondary water, and since the recombination rate w i l l become appreciable as the oxygen and hydrogen ejected to the steam
phase are recirculated in the sealed system.26 In particular, as excess hydrogen builds up in the secondary system from the corrosion of iron and chromium, recombination will occur more readily. Thus, with a wellshielded steam generator, no appreciable steady-state oxygen level is expected; however, it is not known what oxygen level would result if the steam generator were not shielded f r o m the reactor. Assuming an iron oxide--chromiumoxide production rate of 10 mg/dm2.mo (which is reasonable f o r a stainless steel or Inconel system that includes some carbon steel), 2 7 it is calculated that the corresponding hydrogen production rate w i l l be 60 standard liters (2 ft3) per year per 100 ft2 of hot water-steam surface area. This hydrogen w i l l collect predominantly in the condenser vapor space.
If unacceptably high hydrogen production
is anticipated, a void volume can be provided in the condenser for hydrogen gas accumulation. Alternatively, excess hydrogen might be filtered off to a separate tank through a palladium metal barrier.
4 . Corrosion Solids Accumulation in the Steam Generator. The order of magnitude of the corrosion-product accumulation in the steam generator can be estimated from the corrosion rate of 10 mg/dm2-mo assumed above. If there were 500 ft2 of hot water and steam area, approximately 750 g ("1.7 lb) of corrosion products would be produced in one year. A uniform distribution of these corrosion products in a steam generator of 500-gal capacity would result in a solids concentration of 400 ppm. This is comparable with the allowable suspended-solids concentration of 250 ppm set by the American Boiler Manufacturer's Association28 for a 300- to 450-psi m
If a l l solids were uniformly deposited on 200 f t 2 of
heat exchanger surface, t h i s would give a scale approximately 0.006 i n . Since not a l l corrosion products w i l l be transported
thick (400 m g / d m 2 ) .
t o t h e steam generator, and since they w i l l not all be deposited on heat t r a n s f e r surfaces, these numbers suggest upper limits f o r good materials of construction.
From the above rough calculation, it seems
quite possible t h a t a sealed secondary system would operate successfully As i n the
f o r one year o r longer without any water treatment a t a l l .
primary system, however, it a l s o seems l i k e l y t h a t t h e quantity of transportable corrosion products could be reduced by increasing t h e pH of t h e
secondary water with a permanent additive, and this should be seriously considered. O f t h e various pH additives available, t h e most promising from the
standpoint of long-term s t a b i l i t y a r e (1)phosphate buffers, and ( 3 ) ammonia. generator.
( 2 ) caustic,
The f i r s t two would provide a basic pH i n t h e steam
The use of caustic? however, involves some r i s k of caustic
s t r e s s corrosion as a result of concentration by boiling.
provide a basic pH both i n the steam generator and i n the condenser. use of morpholine should a l s o be considered.
Like ammonia, it would
provide a basic pH i n both t h e steam generator and t h e condenser, but
i t s long-term s t a b i l i t y would require investigation. I n recent years, the use of long-chain a l k y l amines (e.g., octadecylamine) as corrosion i n h i b i t o r s i n steam-plant condenser
systems has met with considerable success. the residence time of t h e amine
only a few hours,
and, when us
added continuously o
It has been stated that
e corrosion-inhibiting film i s conventional pxants, amines a r e t h e same time, it i s reported2â€™
t h a t no appreciable decomposit
temperature boiling t h a t t h e normal continuous addi
lming amines i s necessary f f u r t h e r investigation
n t radiation resistance and useful protective lives, they should be considered. L.
Since conventional b o i l e r blowdown i s not desirable
i n t h e present system, a l t e r n a t i v e cleanup methods should be considered. A s pointed out above, large amounts of corrosion-product
s o l i d s a r e not
anticipated, unless, of course, significant quantities of carbon s t e e l
o r other r e l a t i v e l y corrosive alloys a r e used i n t h e secondary system. A margin of safety can be provided i n the design of t h e steam generator I
by making provisions f o r s o l i d s accumulation i n low parts.
cleanup methods include side-steam f i l t r a t i o n , side-stream evaporation, and side-stream o r on-stream magnetic collection. of these methods e x i s t .
Limitations of each
The f i l t e r may be subject t o plugging (depending
on the nature of t h e solids)3 the evaporator system may require automatic control; and t h e magnetic c o l l e c t o r would be e f f e c t i v e only i n removing magnetic oxides.
The need f o r recourse t o these methods i s not a n t i c i -
pated, however, unless r e l a t i v e l y corrosive alloys a r e used i n t h e secondary system.
From the foregoing it seems c l e a r t h a t operation of the secondary system without continuous water treatment o r blowdown i s quite possible. I n addition, t h e use of permanent additives might reduce t h e quantity
of transportable corrosion products, and side-stream cleanup methods could be employed i f shown t o be necessary.
It i s evident t h a t investi-
gations of t h e problems associated with the secondary system water chemistry should be i n i t i a t e d e a r l y i n a program t o design and construct an unattended nuclear power plant. Steam Power Cycle Controls.
Instrumentation and controls f o r the
steam power cycle represent t h e major t h r e a t t o r e l i a b i l i t y , according t o t h e survey made by Gilbert Associates.'
A s s t a t e d i n t h e i r report,
"Failures i n this category [instrumentation and controls] have produced more unscheduled losses of load than t h e total. of a11 other categories.'' Clearly it i s desirable t o remove a l l unnecessary controls from the power system.
The elimination of a reactor control system i s made possible by
s h i f t i n g a portion of t h e control function from the primary system t o t h e secondary system.
Thus changes i n primary-system conditions, such
as temperature d r i f t s due t o r e a c t i v i t y changes, must be accommodated by t h e secondary system.
For a system i n which the load i s very nearly constant, the specified
frequency control t o within I$ i s a loose requirement.
turbine-generator units, where much t i g h t e r frequency control i s normally exercised, hydraulic governor u n i t s a c t through hydraulic p i l o t valves t o operate t h e turbine admission valve.
Operating experience with hydraulic
governor u n i t s i s extensive, and such u n i t s a r e considered highly r e l i a b l e . If oil-lubricated bearings a r e employed on the turbine-generator
u n i t , then a common o i l system may be used with an oil-actuated hydraulic governor.
Experience indicates t h a t o i l leakage from t h e governor system
i s negliglble, and, as w a s t h e case with oil-lubricated bearings, t h e is
system can be designed t o t o l e r a t e a small amount of leakage.
It i s
imperative t h a t hydraulic governor systems have e f f i c i e n t f i l t e r s t o remove s o l i d s from the hydraulic f l u i d . Operation of an oil-actuated governing system within a hermetically sealed steam system i s undesirable because of possible intermixing of t h e o i l and w a t e r .
Use of water as the hydraulic f l u i d with an e s s e n t i a l l y
conventional hydraulic governor i s an a t t r a c t i v e possibility,
since t h e
backlog of experience with hydraulic governors demonstrates t h e i r r e l i a bility.
Discussions were held with a manufacturer presently t e s t i n g a
water-actuated governing system on a steameturbine u n i t .
system being t e s t e d i s of a force-balance type similar t o t h a t for an oil-actuated system.
Pressure i s supplied from small pump vanes located
on t h e turbine shaft.
The degree of regulation achievable with this -actuated governor, and, although
s y s t e m i s comparab
of' t h e maaufacturer
A second ty-pe of governor capable of operating v i t h i n a hermetically
sealed turbine-generator unit i s t h e mechanical governor.
of t h e hydraulic governor over the mechanical governor f o r t h e majority of present-day steam turbine-generator applications i s based i n part on the increased precision of frequency c o n t r o l available with t h e hydraulic system. The r a t h e r loose requirement of 1$ frequency v a r i a t i o n with an e s s e n t i a l l y
constant load may permit t h e use of t h e mechanical governor f o r this appliThe essential. changes required f o r operation of a mechanical gover-
nor include s u b s t i t u t i o n of materials t o permit operation i n a vapor environment.
A d i f f i c u l t y encountered i n operation of turbine-generator controls
i s sticking of t h e turbine admission valve.
The force required f o r valve
operation i s an important consideration i n t h e use of mechanical governing systems where t h e power amplification i s somewhat limited.
problem, which results from oxide formation on valve stems, has been associated primarily with p l a n t s operating at high steam temperatures and press w e s and should n o t exist at the temperatures anticipated f o r t h i s system. Other methods f o r achieving governor control i n a hermetically sealed system could be based on t h e e l e c t r i c a l output of t h e generator.
e l e c t r i c a l governing system usually functions by sensing the frequency output of t h e generator and comparing it with a desired set-point frequency t o obtain an e r r o r signal; t h e e r r o r signal operates a p o s i t i o n c o n t r o l l e r that regulates t h e turbine admission valve.
A hydraulic o r mechanical
governor external t o t h e hermetically sealed system could operate off a synchronous motor t o regulate t h e turbine admission valve, and no penetrat i o n s of t h e sealed system would be required. An a l t e r n a t e scheme f o r eliminating t h e turbine governor and admission valve was examined.
T h i s system would vary a dummy e l e c t r i c a l load, i n
p a r a l l e l with t h e normal system load, t o control t h e turbine speed.
i s e n v i s i o n e d t h a t saturable r e a c t o r s i n s e r i e s with a r e s i s t a n c e load
would vary t h e impedance through t h e dummy c i r c u i t as a function of output frequency. circuit.
Solid-state components can be employed throughout t h e
Although this system has not been developed, t h e advantage of
completely eliminating t h e need f o r a turbine governor or turbine admission valve makes it of i n t e r e s t t o t h e present application.
The exact type of frequency-governing system t o be employed would require a more d e t a i l e d analysis than that presented here.
however, t h a t several schemes are capable of operating successfully with a hermetically sealed system.
For this application it would probably be
best t o undertake t h e development of t h e water-actuated hydraulic governor system as t h e primary e f f o r t and t h e development of other types of governor as t h e backup e f f o r t . Even i n an e s s e n t i a l l y constant-load plant, it i s necessary t o regulate feedwater flow t o t h e steam generator because of short-term perturbations and longer term drifts i n t h e system's behavior.
t i o n a l b o i l e r water-level c o n t r o l l e r s and feedwater regulators contain numerous sensors and control devices that would, i f used, jeopardize t h e operation of an unattended plant. With a confitant water inventory and a f a i r l y steady load, feedwater
regulation can be obtained without t h e usual c o n t r o l l e r s by u t i l i z i n g a b o i l e r feed pump t h a t i s quite sensitive t o suction head.
capacity would vary i n response t o the depth of water accumulated i n t h e hot well.
Performance curves f o r a submerged pump having such
c h a r a c t e r i s t i c s a r e shown in Fig. 6.
T h i s pump i s designed t o regulate
t h e depth of l i q u i d above t h e pump suction by t h e e f f e c t of suction head on t h e degree of cavitation occurring at the impeller i n l e t .
Pump capacity i s sharply reduced at low hot-well l e v e l s by p a r t i a l vaporization of t h e l i q u i d a t the impeller eye.
With high suction heads in-
suring against cavitation, t h e pump performance as t o t a l head vs capacity
i s that represented by t h e dashed l i n e i n Fig. 6.
A reduction of suction
head t o below t h e l e v e l indicated by t h e dashed l i n e w i l l i n i t i a t e cavitation.
Pump capacity will then be regulated by t h e suction head i n
accordance with the s o l i d l i n e .
Burns and Roe20 s t a t e that pumps of t h i s design are currently i n use i n t e n large-capacity u t i l i t y plants. operate with c a v i t a t i o n and a r e expec
The pumps a r e designed t o t o operate well i n excess of a
year. The use of a pump t h a t i s sensitive t o t h e hot-well water l e v e l i n a fixed-inventory system obviates t h e need f o r auxiliary valves, l e v e l
PUMP CAPACITY (gpml
Fig. 6. Characteristic Performance Curves f o r a Submerged Boiler Feedwater Pump. controls, sensors, o r actuators.
The importance of t h i s t o control
system r e l i a b i l i t y i s evident. Other methods of achieving feedwater control and l e v e l control were examined.
One scheme would employ overflow l i n e s from a l e v e l drum i n
the steam generator t o the main condenser o r t o a feedwater heater.
Under normal operation t h e water l e v e l i n the drum would remain between two overflow o u t l e t s .
Thus some water could continuously r e c i r c u l a t e t o
t h e condenser through t h e lower overflow, while steam would r e c i r c u l a t e t o t h e condenser through t h e upper overflow.
A r i s e i n drum l e v e l would
send water through the upper overflow and increase t h e recirculation, while a f a l l i n l e v e l would permit steam t o e x i t at t h e lower overflow and reduce the recirculation.
By proper sizing of t h e return lines, t h i s
system, o r variations of it, could be made t o keep t h e water l e v e l nearly constant.
The above sysCem i s i n e f f i c i e n t i n that water heated i n t h e ,
steam generator i s recirculated back t o t h e condenser without contributing t o the plant e l e c t r i c a l output, but the l o s s might not be important i n t h e present application. Methods such as t h e use of eductors might a l s o be possible f o r feedwater flow control.
The submerged pump, however, o f f e r s greater simplicity
and r e l i a b i l i t y than any other method considered. Pumps.
The conventional steam plant employs hot-well condensate
pumps, b o i l e r feed pumps, and sometimes forced-circulation pumps f o r t h e steam generator.
I n t h e plant under study, consideration w a s given t o a
system u t i l i z i n g a single multistage pump t o take care of a l l t h e feedwater pumping functions.
The u t i l i z a t i o n of a single multistage pump
capable of operating from condenser pressure t o steam generator pressure has been demonstraced i n the operation of SM-1.
The additional mcdifi-
cation desired f o r this application i s t h e development of a r e l i a b l e high-head low-flow canned-rotor pump t o perform t h e required service. The very successful operation of t h e canned-rotor primary pumps i n pressurized-water service indicates that t h e u t i l i z a t i o n of t h e cannedr o t o r pump i s completely compatible with high r e l i a b i l i t y requirements. The attachment of a canned-rotor drive t o a multistage impeller i s a combination which pump manufacturers believe presents no d i f f i c u l t y o r extrapolation of known technology.
If the required development were not
accomplished within t h e t i m e available, a noncanned pump with t h e pump leakage controlled and bled back t o the condenser could be used i n a semiconventional" t h e turbine shaft and t h u s e
uld have t o operate at
f o r t h e r e a c t a r directed a t t e n t i o n t o the use of an i n j e c t o r r a t h e r than
a motor-driven feedwater pump.
The r e l a t i v e l y low temperature of con-
densate from the condenser (possibly 70 t o 80째F) and t h e minor importance
of pumping efficiency a r e conditions well suited t o t h e use of an i n j e c t o r .
It i s p a r t i c u l a r l y advantageous t h a t i n j e c t o r s a r e of simple construction and have no moving parts.
However, i n j e c t o r s have t h e disadvantages t h a t
t h e drop i n pumping efficiency at pressures above 300 psig necessitates t h a t l a r g e quantities of steam be used f o r pumping and t h a t feedwater at temperatures above 100 t o 110째F cannot be handled by a single-stage system. Present experience with i n j e c t o r s i s such t h a t they a r e not considered t o be e n t i r e l y r e l i a b l e f o r unattended operation, although this opinion i s i n p a r t based on t h e e f f e c t of load fluctuations on i n j e c t o r operation,
Aside from questions of operability and r e l i a b i l i t y , t h e feedwater regul a t i o n obtained automatically with a cavitating pump makes t h e pump more a t t r a c t i v e than the i n j e c t o r .
The inherent simplicity of t h e i n j e c t o r
suggests, however, t h a t it be studied further. Heat Exchangers.
The basic heat exchanger requirement f o r t h e
pressurized-water secondary system includes only t h e steam generator and t h e main condenser.
The use of feedwater heaters, regenerators, reheaters,
and similar equipment, which conventionally a r e employed t o increase t h e plant thermal efficiency, would be largely contingent on t h e i r e f f e c t on reliability.
A few heat-recovery devices would not endanger system
r e l i a b i l i t y , but i n general they should be kept t o a minimum. The v a s t experience i n t h e construction of heat-exchange equipment indicates t h a t u n i t s of very high i n t e g r i t y can be fabricated and can operate successfully without f a i l u r e f o r periods of more than one year. A s discussed i n connection with t h e primary system, successful operation
depends heavily on a high l e v e l of q u a l i t y control during fabrication, but no extrapolation of current technology i s required.
The r a t h e r
moderate conditions of water temperature and pressure envisioned f o r t h e secondary system amplify the confidence t h a t leaktight units capable of one year of continuous operation can be constructed. Potential materials of construction examined f o r heat exchanger service were t h e a u s t e n i t i c s t a i n l e s s steels, Inconel, and t h e coppernickel alloys.
Austenitic s t a i n l e s s s t e e l s can be used i f one can insure
operation of a system f r e e of chlorides and/or maintain a very low l e v e l
The presence, however, of even small amounts of chlorides
and oxygen i n such systems may r e s u l t i n crevice and stress-corrosion ~ r a c k i n g . ~ ’On t h e other hand, many a u s t e n i t i c s t a i n l e s s s t e e l systems having acceptable control of water chemistry do e x i s t and have operated successfully f o r years without f a i l u r e of heat exchange equipment.
this connection it should be noted t h a t t h e SM-1 has operated since 1957
and has not encountered a single tube leak i n any heat exchanger i n the system.1’
The use of Inconel f o r steam generators i n pressurized-water
reactors has been investigated i n recent work a t
apparently o f f e r s excellent resistance t o both s t r e s s corrosion and general corrosion.
It i s t h e preponderant opinion of heat exchanger
manufacturers t h a t materials can be selected and equipment can be fabricated t h a t can meet t h e one-year operational requirement.
I n view of t h e preceding discussion, t h e proposed concept of a hermetically sealed secondary system with a canned-rotor pump and a turbine-generator as the only rotating machinery and a turbine-governorfrequency control as t h e only control system appears promising f o r t h e present application.
EFFECT OF AWERI3.D REQUEGNENTS ON THE DESIGR CONCEPT The design concepts presented i n this report a r e c l e a r l y a sensitive
function of the requirements f o r t h e nuclear power plant, of t h i s section i
c a r e f u l re-e
The purpose that would require
e i n i t i a l specifications.
s i z e and weight a r e considera-
i s important; t h e plant i s i s variable.
1 achieved, information p e r t i -
I * frequency, and perhaps flow information could be transmitted t o a receiving
center where elementary corrective actions could be i n i t i a t e d by simple signals;
Fixed-increment adjustments of control rods would permit cor-
r e c t i o n f o r long-term r e a c t i v i t y changes indicated by changes i n water temperature.
similarly, fiked-increment adjustments of a governor system
could correct f o r d r i f t of t h e s e t point.
Communication a l s o would permit
nonoperating spares t o be remotely energized o r emergency shutdown of t h e plant t o be i n i t i a t e d when received information indicated t h e need.
hazard of misoperation and f a i l u r e because of malfunctioning of the communication system would, of course, exist, but precautions, such as t h e use
of coincident signals, might be employed t o prevent spurious actions. Thus provision of a system f o r communicating with the plant might permit improvement of plant r e l i a b i l i t y . Much of t h e broad simplification proposed t o achieve r e l i a b i l i t y f o r t h e unattended plant would be useful i n designing a plant f o r operat i o n with smaU crews.
Some of the modifications proposed f o r t h e control
and water-treatment systems might greatly reduce t h e operation and mainte-
nance requirements of an attended plant. Plant Size and Weight Restriction of t h e s i z e and weight probably would a l t e r t h e design of a reactor power plant i n a manner that would adversely a f f e c t i t s reliability,
The design considerations imposed by moderate r e s t r i c t i o n s
might have l i t t l e effect, whereas overriding considerations of s i z e and
weight would l i k e l y render reactors such as t h e pressurized-water concept unsuitable.
In designing a pressurized-water reactor f o r limited s i z e
and weight, increased system pressure, temperatures, and flow velocities,
higher reactor power density, and greater turbine-generator speeds would have t o be investigated.
Although incorporation of many of these changes
would not necessarily decrease system r e l i a b i l i t y , they might remove t h e design from t h e r e a l m where experience with present pressurized-water reactors i s applicable.
Long-Term R e l i a b i l i t y and Expendability The design of an unattended power plant capable of operatin@; several years on a continuous b a s i s imposes even more severe r e s t r i c t i o n s on the P a r t i c u l a r l y sensitive t o longer term operation would be t h e water-treatment requirements. Simple systems t o permit systems and components,
introduction of chemical additives might be needed, and side-stream p u r i f i c a t i o n would probably be necessasy. The problem of building i n and controlling s u f f i c i e n t r e a c t i v i t y f o r longer term operation might necessitate t h e use of an automatic control system, although it i s possible t h a t the l i f e t i m e obtainable using burnable poisons would extend well beyond one year. If automatic control were required, it might be t h a t a programmed shim-control system operated by a clock o r a watt-hour meter (as mentioned under Reactor Control) would be more r e l i a b l e than a system which employed feedback f r o m t h e reactor,
Almost certainly, longer term r e l i a b i l i t y would be
achieved i f one proceeded t o build plants i n i t i d l y capable of one year
of operation and then modified them as operating experience dictated, Two major changes i n concept would be required i f t h e plant, o r even i t s major components, were not considered t o be expendable.
ment such as the turbine-generator and t h e pumps would require protective devices capable of shutting down the plant, md deposition of a c t i v i t y might have t o be controlled so t h a t core components would be repairable.
These changes, p a r t i c u l a r l y the use of protective devices, would decrease t h e probability of the plant running a year without shutdown. A compromise between expendability and r e p a i r a b i l i t y might be
a t t r a c t i v e f o r some applications. the plant i n packages
T h i s could involve construction of
ore package, turbine-generator package,
The cost might be much l e s s
an e n t i r e plant.
last t o operate successfully
on a variable load might s i g n i f i c a n t l y change recommended components and
The use of an induction generator, f o r Instance, with
a variable load might produce voltage variations beyond t h e desired
Similarly, c e r t a i n frequency control systems (such as t h e
mechanical governor) might be l e s s adaptable t o a variable-load plant. The e f f e c t of sudden load changes on reactor control requirements would have t o be examined. If t h e load were intermittent i n nature, t h e requirements f o r t h e
plant would be even more stringent.
Both t h e primary and the secondary
systems would have t o be capable of going t o power unattended.
of a variable dummy load t o preclude zero-power operation and t o eliminate severe load. changes, if necessary, might improve t h e r e l i a b i l i t y of a load-following system.
Small o r gradual load changes not requiring
plant shutdown could probably be accommodated by t h e design concepts i n i t i a l l y proposed. Effects of Time Schedule Certainly one of the most fundamental r e s t r i c t i o n s affecting t h e design concept i s t h e time within which t h e plant i s required.
T h i s is
emphasized by t h e suggested e f f e c t s of time changes on t h e concepts presented,
If production of an acceptable system i n three years were
required, it i s probable t h a t success of t h e program would not be based on design modifications beyond t h e semiconventional system proposed i n this report (although t h e hermetically sealed secondary system would be
attempted as a secondary e f f o r t )
If, on the other hand, four t o f i v e years were allowed f o r procurement of an acceptable plant, t h e hermetically sealed plant would c e r t a i n l y be u t i l i z e d .
For periods beyond f i v e years,
one could afford t o deviate t o s t i l l greater degrees from known technolog i e s and proven concepts.
SUMMARY Am) CONCLUSIONS Power reactors a r e normally designed f o r constant a t t e n t i o n and continuous maintenance, and there a r e no existing reactors capable of producing e l e c t r i c i t y f o r one year unattended.
Not only a r e complete
power plants incapable of m e year of operation without maintenance, but few of the individual components have the r e q u i s i t e r e l i a b i l i t y . T h i s i s p a r t i c u l a r l y t r u e of control systems and of steam power systems, f o r which regular maintenance i s accepted practice.
It i s l e s s t r u e f o r the
components i n the prima;ry system of pressurized-water reactors, because t h e Navy program has l e d t o the development of r e l i a b l e components. Since existing systems do not have the required r e l i a b i l i t y , t h e f e a s i b i l i t y of achieving an unattended reactor i n four years a c t u a l l y involves t h e question of whether a r e l i a b l e system can be developed i n t h e a l l o t t e d time. r
The four-year period i s so short t h a t there would be
l i t t l e opportunity f o r perfection of new concepts, and the authors believe a program t o achieve t h e stated objectives should, wherever possible, be based on proven concepts and proven technology.
System r e l i a b i l i t y i s
most l i k e l y t o be achieved by simplification t o t h e point t h a t operation depends on a minimum number of components.
The objectives of this study
make simplification p a r t i c u l a r l y promising, since t h e system i s not required t o be repairable, have a long l i f e , follow a varying load, o r operate i n a populous area.
Although r e l i a b i l i t y i s not consistent with
minimum cost, elimination of equipment by simplification may o f f s e t t h e increased cost of individual components. A pressurized-water reactor fueled with highly-enriched uranium was
selected f o r this application because of the successful experience with it and because it i s amenable t o extreme simplification.
h i s service.
It i s , however, t h e
n t i n a limited time. evolved from this study only mechanically coolant pump, the b o i l e r feed
t e r s t a r t u p without a nuclear control system.
Burnable poison i s used t o l i m i t g r o s s r e a c t i v i t y
* changes, and the negative temperature coefficient i s used t o compensate
f o r changes which a r e not eliminated.
The e f f e c t of allowing the core
temperature t o vary (say over a range of 50째F) i s t o shift t h e reactor control problem t o t h e turbine governor.
Since t h e governor i s already
required f o r frequency regulation, no additional components are made necessary. The design of a turbine governor and admission valve i s eased by elimination of t h e normal requirement of t i g h t frequency control over a wide range of loads.
Although several existing types of governor could
be modified f o r this system, t h e r e l i a b i l i t y of a governor i s one of t h e major system uncertainties.
(The use of a variable dummy e l e c t r i c load
t o control frequency without a t h r o t t l e valve o r a governor appears t o merit serious investigation. )
2. An automatic control system f o r t h e pressurizer i s avoided by a design w h i c h permits the e l e c t r i c heaters t o remain on continuously, 3.
Boiler feedwater controls are eliminated by use of an existing
type of cavitating pump that i s sensitive t o suction head.
The pump w i l l
maintain a constant water l e v e l i n the hot well and, consequently, i n a fixed-inventory system, w i l l maintain a constant l e v e l i n the steam generatur
4 . Both the primary and secondary systems are hermetically sealed, and there i s neither water treatment nor blowdown.
It appears f e a s i b l e
t o operate f o r a year i n this manner without encountering disabling problems from corrosion o r crud buildup. The operation of a hermetically sealed secondary system requires the use of water-lubricated turbine-generator bearings and a generator which w i l l operate i n a vapor o r water environment.
brushless generators axe capable of such operation.
While only a few
turbines i n the s i z e range needed have been b u i l t with water-lubricated bearings, canned-rotor pumps employing water bearings have been successfully operated f o r m y years.
(If development of water-lubricated bearings
does not proceed rapidly enough, an oil-lubricated turbine with a brushl e s s generator can be accommodated i n a modified design.)
2 5 I
The r e m i n i n g components of t h e power plant appear t o present no r e l i a b i l i t y problems. Fuel elements with l i f e t i m e s longer than required a r e available. reliability.
Primary c i r c u l a t i n g pumps have a demonstrated h i s t o r y of With careful design, selection of materials, fabrication,
and t e s t i n g , t h e piping, heat exchangers, and pressurizer can be made t o be r e l i a b l e .
It i s probable t h a t using t h e concepts proposed, r e a c t o r s capable of unattended operation can be available i n four years.
To achieve this
objective, a conceptual design study must progress r a p i d l y t o t h e beginning
of d e t a i l e d design, i n order t h a t t h e construction of a prototype can be accomplished e a r l y i n t h e program.
Studies of water chemistry, core
physics, and reactor control, and t h e development of a frequency-control system, turbine-generator,
and b o i l e r feed pump should begin immediately.
A well-integrated and rapidly moving program can lead t o useful power
p l a n t s i n t h e a l l o t t e d time.
Experience with t h e f i r s t generation of
r e a c t o r s and continuing development w i l l increase t h e r e l i a b i l i t y beyond
what may be achieved i n i t i a l l y .
ire. i I
n Nuclear Power Plants," 1. "Reliability Study Equipment and Components i Gilbert Associates, Inc., Report No. 1512 (to be issued).
2. "APF3-1 Operating Experience,'I Nucleonics 17, 5+51
3. M. L. Myers, "Study of Outage Experience with Selected Boiler and Turbine Generator Units," ORNL CF-60-3-56 (March 31, 1960).
4. "SL-1 Reactor Evaluation, Final Report," p. V-22, IDO-19003 (July 19, 1959) 5.
E. H, Bankcker et al., KAPb2070 (March 7, 1960). (Classified)
6. "PM-1 Nuclear Power Plant Program, Third Quarterly Progress Report, 1 September to 30 November 1959," p. 111-133, MND-M-1814. 7.
T. G. Williamson, "APPR-1 Burnout Calculations," APAE Memo 126 (April 10, 1958).
"SM-1 Operating Contract Annual Report, February 195&-February 1959," p. VII-23, APAE-47 (June 30, 1959).
9. A. Radkowsky, "Theory and Application of Burnable Poisons," 1958 Geneva Conference, Vol. 13, pp. 42645.
LO. E. E. Gross, "Reactivity-Lifetime Comparison of U235 and 10,000 MWD/Ton Pu as Fuel for the APPR, ORNL CF-59-5-106 (May 21, 1959). 11. C. R. Feavyear, SM-1, personal communication, May 5, 1960.
12. D. J. DePaul, "Corrosion and Wear Handbook," pp. 3336, TID-'7006 ( W h 1957). 13. R. Hurst and S. McLain, "Technology a d Engineering," pp. 281-ff., Progress in Nuclear Energy Series, Pergamon Press, New York, 1956.
U. "The Shippingport Pressurized Water Reactor," pp. 181-ff
Wesley Publishing Company, Reading, Massachusetts, 1958.
15. DePaul, 2.
16. DePaul, 2.G.,pp. 22+38.
17. G. P. Simon, '"Vande Graff Study. Fourth Interim Report on the Deposition of Corrosion Products Under Irradiation," pp. 13446 in WAPD-BT-16 (December 1959).
.. r1 -
18, G. P. Simon, "The Use of Neutron Activated Specimens f o r the Study (March 1960). of Corrosion Product Release Rate," WAF'D-T-1114 19. B. G. Schultz, K. H. Vogel, and G. P. Simon, B e t t i s Plant, private communication (June 1, 1960). 20.
B u r n s and Roe, Inc., provided consulting service regarding c e r t a i n aspects of t h e secondary system.
L. D. Damretowski, The E l l i o t t Company, personal communication (May 16, 1960).
J. G. W i l l i a m s , Worthington Corporation, personal communication (May 14, 1960). .
R. E. V u i a , 5. A. Paget, and R. D. Winship, "Steam Generators f o r Pressure Water Reactors," Tw-1 (PECL-832) (June 1959), Section V I I , "Bibliographies on Boiler and Feedwater Technology. "
e,, p. 286.
H u r s t and McLain,
S. Gordon and E. J, Hart, "Radiation Decomposition of Water Under S t a t i c and Boiling Conditions," 1958 Geneva Conference Proceedings, Vol. 29, pp. 13-18.
C, J. Hochanadel, "The Radiation Induced Reaction of Hydrogen and Oxygen at 25째C-2500C," 1958 Geneva Conference Proceedings, Vol. 7, pp. 521-5.
D. E. White, "Evaluation of Materials f o r Steam Generator Tubing," pp. 1 5 6 9 6 i n WAPD-l3T-16 (December 1959).
V u i a , Paget, and Winship,
Go, p. IV-4.
ORNL -298 5 Reactors -Power TID-4500 (15th ed. )
1. L. G. ALexander 2-11, C. F. Baes, Jr. l2. R, J. Beaver 13, M. Bender 14. D. S. BillFngton 15. E. P. Blizard 16. A, L, Boch 17. E. G. Bohlmann 18. C , J. Borkowski 19. G. E. Boyd 20. E. J, Breeding 21. R, B. B r i g g s 22, C. E, Center (K-25) 23. R. A. C h a r p i e 24. H. C. Claibome 25. W, G. Cobb 26. W. B. C o t t r e l l 27. J, A, Cox 28, F, L. Culler 29. H. N. Culver 30. H. E. Cunningham 31. L. B. Emlet (K-25) 32. L. G. Epel 33. E. P. Epler 34, M. H. Fontana 35. J. Foster 36, A. P. Fraas 37. J. H. Frye, Jr. 38. C . H. Gabbard 39. J. L. Gabbard, ORGDP 40. W. R. G a l l 41. B. L, Greenstreet 42. J. C. Griess 43. W. R, G r i m e s 44. E. E. Gross 45. E, Guth 46. R. J. Hefâ€™ner 47. A. Hollaender 48. C. J. Hochanadel 49. H, W. Hoffman 50. A. S, Householder 51. S, Jaye 52. W, H, Jordan
P. C. M. R. 57. P. 58. J, 59. R.
53 54. 55, 56.
60. 61. 62. 63.
M. H. R. W.
64, E. 65. H. 66. H. 67. A. 68. R, 69. K. 70. J.
n. M. 72 F . 73. M, 74. P.
75. A. 76. P, 77. C. 78. B. 79. 8H9. 9W114. 115. 116. 117. 118. 119 120. 121. 122 123. 124. 125. 126. 127. 128.
M. G. H.
A, L. E. M. S. A,
I. F, J. W.
Kasten Keim Kelley Korsmeyer Lafyatis Lane Livingston Lundin MacPherson MacPherson Manly Mann McCurdy McDuffie Miller S. Miller Z. Morgan P. Murray (Y-12) L. Myers H, Neil1 L, Nelson Patriarca M. Perry H. Pitkanen A. P r e s k i t t E. Prince E, Ramsey C. Robertson W. Rosenthal Samuels W. Savage W. Savolainen D. Schaffer D. Shipley J. Skinner H, Smiley (K-25) H. Snell Spiewak J. Stanek A. Swartout C . Thwber Tobias B. Trauger
R, P, T, B. G. A. S. I, G. E. D. R. C. F. J.
129. 130. 131. 132. 133. 134.
J. J. Tudor E. Vincens C. S. Walker A. M. Weinberg G. D. Whitman C . E. Winters
135144. M. M. Yarosh 145176. Laboratory Records 177. Laboratory Records, RC 17&179, Central Research Library 18C-182. ORNL Y-12 Technical Library Document Reference Section
183. W. B. Allred, Combustion Engineering, NRTS 184-185. G. C. Banick, General E l e c t r i c Company, 253 Main S t r e e t East, Oak Ridge, Tennessee 186. A. E. Becker, Jr., Westinghouse E l e c t r i c Corp., Lester Branch P.O. Philadelphia 13, Pennsylvania 187. L. V. Bradnick, Pierce Governor Company, Anderson, Indiana 188. Gordon Brown, Alco Products, Dunkirk, New York 189. J. A. Brown, Allis Chalmers Mfg. Co., 702 Hamilton Bank Building, Knoxville 2, Tennessee 190. R, L. Cawthon, Westinghouse E l e c t r i c Corp., 605 Burwell Building, Knoxville, Tennessee 191. M. A. Cordovi, International Nickel Company, 67 Wall Street, New York 5, New York 192. F. C r i m , G"F, F o r t Belvoir, Virginia 193. L. P. Damratowski, E l l i o t t Company, Division of Carrier Corp., Jeannette, Pennsylvania 194. R. S. Etchison, Westinghouse E l e c t r i c Corp., E a s t Pittsburgh, Pennsylvania 195. C . R. Feavyear, SM-1, Fort Belvoir, Virginia 196. L, Flowers, Atomics International, Canoga Park, C d i f o r n i a 197. C. H. Fox, Martin Company, Baltimore, Maryland 198. J. G. Gallager, Alco Products, Schenectady, New York 199. M. Golik, Westinghouse E l e c t r i c Corp., B e t t i s Plant more, Maryland
y, Fitchburg, Massachusetts er, Rome, New York
212. Captain R. I,. Morgan, Idaho Operations Office 213. S. Neal, General E l e c t r i c Company, Fitchburg, Massachusetts Col. G. B. Page, Army Reactors Branch, Division of Reactor 2l-4-221. 222. 223. 224. 225. 226. 227. 228. 229. 230. 231. 232. 233. 234. 235. 236. 237
238. 239. 240-816.
Development, AEX!, Washington M. E. Peterson, General E l e c t r i c Company, Schenectady, New York 160 W. Broadway, New York 13, R. E. Ramsey, B u r n s & Roe, Inc New York R. Rhudy, General E l e c t r i c Company, Schenectady, New York B. G. Schultz, Westinghouse E l e c t r i c Corporation, B e t t i s Plant V. D. Schwartz, Combustion Engineering, Windsor, Connecticut B. Sellers, E l l i o t t Company, Division of Carrier Corp., J e axmet t e, Pennsylvania G. P, Simon, Westinghouse E l e c t r i c Corporation, B e t t i s Plant A. D. Sutton, General E l e c t r i c Company, Fitchburg, Massachusetts A. Swenson, Woodward Governor Company, Rockford, I l l i n o i s K. H. Vogel, Westinghouse E l e c t r i c Corporation, B e t t i s Plant E. Wagman, Office of Naval Research, Washington B. H. Webb, Westinghouse E l e c t r i c Corporation, E a s t Pittsburgh, Pennsylvania I. H. Welinsky, Westinghouse E l e c t r i c Corporation, B e t t i s Plant J. R. Wetch, Atomics International, Canoga Park, California J. G. Williams, Worthington Corp., Newark, New Jersey H. H. Wilson, Gilbert Associates, P.O. Box 1498, Reading, Pennsylvania P. C . 5 l a , Combustion Engineering, Windsor, Connecticut Division of Research and Development, AMI, OR0 Given d i s t r i b u t i o n as shown i n TID-4500 (15th ea,) under Reactors -Power category (75 copies OTS)