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1 I
WASH-1 222 UC-80
.
AN EVALUATION
OF THE MOLTEN SALT BREEDER REACTOR
I
[----N0 1 I C E
I
This report was prepared-as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor m y of their contractors, subcontractors. or their employees, makes any warranty, express or implied, or assumes any y, legal liability or responsibility for the a c c ~ ~ a c completeness or usefulness of m y information, apparatus product or process disclosed, or represents that its US; would not infringe privatelv owned richts
$8
F Od~ e by the Superintendent of Documents, U.8 0 veroment Prlntfng Ofaee Washfngton, D.C., 20402
*;
v4
I)ISTRIBUTION OF THIS DOCUMENT IS UNLlMn
This report was prepared as input to the Office of Science and Technology's Energy Research and Development Study conducted through the Federal Council for Science and Technology.
I
The contents
represent the views of the panel members and not necessarily those of the Office of Science and Technology.
TULE OF CONTENTS
Page
I. INTRODUCTION
. 0 0 0 0 . . 0 0 . . 0 o . o 0 0 0 . . 0 0 0 0 0 0 o 0 0 0 0 0 0 0 o o o 0 o o * o o o 0 o o o o o
........ ............................................. 111. RESOURCE UTILIZATI .......................... I V o HISTORICAL DEVELOPMENT OF SALT REACTORS ................ V. SALT CONCEPT ............... VI . STATUS OF MSBR TECHNOLOW ..................................... . MSRE - The Reference Point €or Current Technolwv ........ . Continuous Fuel Processing - The t o Breeding ......... 1. pment ....................... . Fuel P u t a 1 Materials ............... . Molten S a l t R ...... 1. Fuel and Coolant S a l t s ............................. 2 . Reactor Fuel Containment Materials ................. 3. Graphite ........................................... 4 . Other S t r u c t u r a l Materials ......................... . T r i t i u m . Problem of Control ........................... II. SUMMARY
MOLTEN
MOLTEN
DESCRIPTION
BREEDER REACTOR
A
B
Key
ic
2
C
D
E
A
.
Reactor Equipment and Sktems Development
.
................
......................................... . ............................................ F . Maintenance . D i f f i c u l t Problem f o r t h e HSBR ........... f o r the .................. C . Safetv .Different H . Codes. Standard High Temperature DesiRn Methods .... 1
2
Components Svstems A
es
IAL PARTICIPATI
MSBR
MSBR PRWvRAM
0
. ..
............. ................................. ............. .................................. A ............. ..................................
1 8
10 12 14 19 19 22 22
27 30 33 34 35 38 38 41
43 45 47 48 51
REFERENCES
53
c .
.
25 26
CONCLUSIONS
APPENDIX
\
A- 1
.
LIST OF ThLES AND'FIGURES
Page Tables I
I1 111
2.
Selected Conceptual D e s i g n D a t a for a Large MSBR.
17
Important Dates and Statistics for the MSRE .
Camparison of Selected Parameters for the MSRE and 1000 W ( e ) MSBB
21
Flowsheet for Processing a Single-Fluid MSBR.
24
-. .
~
..
.. ,
, .
. . ..
.
..
. .
AN EVALUATION OF TEE MOLTER SALTBREEDER REACTOR I.
INTRODUCTION
The Division of Reactor Development and Technology, USAEC,was assigned -..~.. the responsibility of~assessing the status of the technology of the. -( Molten Salt Breeder Reactor (MSBR)& part of the Federal Council of _ . : Science and Technology Research and Development Coals Study. In. conducting this review, the attractive
features and _ areas ~. problem associated wi-th the concept have been examined; but more impor-tantly, _ the assessment has beendirected~.to -. __I-- provide a view of the Atechnologp and engineering development efforts and the associated government and i -;-
~.
industrial
commitmentswhich tiould be required to dkvelop the MSBR ._. safe, reliable and economic power source for central station
into-a
applicatiou. The MSBR 'concept, currently '.
&er
study at the Oak Ridge National
Laboratory (ORNL), is based on use of~'a.circulati&
fluid
reactor coupled with on-line continuous‘ fuel processing.
f&l As presently
envisioned, it would operate as a thermal spectrum reactor system ~utiliti& the
a thorium-u&i&
@otent&i;
for
bro&e&d
fuel cycle. u&izat$o+of.the .
Thus, the concept kuld nation's
offer
natural
~resources through operation of a b.reeder system employing &&her _ _~ _
prg~i~“‘-f;bN
0~ THIS DOCUMENT IS UIWMiTED
-2I
sustaining, competitive i n d u s t r i a l c a p a b i l i t y â‚Źor producing economical
power i n a r e l i a b l e and s a f e manner.
A b a s i c ' p a r t of achievement of
t h i s objective is t o gain public acceptance of a new form of power production.
Success i n such an endeavor is required t o permit the
,
u t i l i t i e s and others t o consider t h e concept as a v i a b l e option f o r
generating electrical power i n t h e f u t u r e and t o consider making the heavy, long-term commitments of resources i n funds, f a c i l i t i e s and personnel needed t o provide the t r a n s i t i o n from t h e e a r l y experimental f a c i l i t i e s and demonstration plants t o f u l l scale commercial r e a c t o r power plant systems.
Consistent w i t h t h e policy established for a l l power r e a c t o r development programs, the NSBR would r e q u i r e t h e successful accompl$shment of three b a s i c research and development phases:
. An i n i t i a l research and development phase in which t h e b a s i c technical aspects of the MSBR concept are confirmed, involving exploratory development, laboratory experiment, and conceptual
i n which t h e engineering and man c a p a b i l i t i e s are developed.
This includes t h e conduct of
in-depth engineering and prooftesting of first-of-a-kind components, equipment and systems.
These would then b e
incorporated i n t o experimental i n s t a l l a t i o n s and supporting
LJ
,
-3test
LJ
facilities
to assure adequate xtnderst&ding
and performance experience
associated
environmental efforts
characteristics,
As these research
the technohgical
progress,
and decision
is sufficiently
. A.third
~phase in.which
acceptable
,'
-,
Significant
When the in the
;;
-plants.
plants
manner.
make large scale commitments
by developing
construction,
in a safe,
experience
confidence.
developed and confidence
the utilities
to electric-generating
power plants
would:need
reached that would permit
necessary
demonstration
manage the design,
and development
the next stage would be the construc-
-system was attained, .tion of large
economic and
uncertainties
points
development to proceed with technology
as to gain overall
with major operational,
parameters.
to be resolved
as well
of design
reliable,
.test
the capability
and operation
economic,
of
to
these
and environmentally
f <
with
the Light
Water Reactor- (LWR), the high
Temperature Gas-cooled Reactor ~(HTGR) and the Liquid Metal-cooled Fast . Breeder Reac.tor (L?JFBR) has been gbined over the.past two~decades pertaining nuclear
to the 'efforts reactors
This experience
that
to the point
are~required of public
to-develop
and commercial
and advances ~-. acceptance.
has clearly demonstrated that the phases of develop_ m&-and d&nonstration'should be simiiar'regardless of the-energy _.. _-. ~ -~~ ~_~_ _ .~ . through each of concept being exploredi that the logicai.progression a.
-4-
t.e phases is essential; and that completing t h e work through t h e t h r e e phases is an extremely d i f f i c u l t , t i m e cunsuminp: and c o s t l y undertaking, requiring the h i g h e s t level of t e c h n i c a l management, professional competence and organizational s k i l l s .
This has again
been demonstrated by t h e recent experience i n t h e expanding LWR design, corm t r u c t i o n and l i c e n s i n g activities which emphasize clearly t h e need f o r even s t r o n g e r technology and engineering e f f o r t s than .
were i n i t i a l l y provided, although i h e s e k r e
cases for t h e * f i r s t .experiments and demonstration p l a n t s .
The LWBR
program, which is r e l a t i v e l y w e l l advanced i n its development, tracks c l o s e l y this LWR experience and has f u r t h e r reinforced t h i s need a. i t a p p l i e s t o t h e technology, development and engineerinff a p p l i c a t i o n areas
.
It should also be kept in mind t h a t t h e l a r g e backlog of commitments
and t h e shortage of q u a l i f i e d engineering and t e c h n i c a l management personnel and proof test facilities 'in t h e government , i n -2ndusttv and
i n t h e u t i l i t i e s make i t even more necessary t h a t a l l t h e r e a c t o r systems be thoroughly designed and t e s t e d b e f o r e a d d i t i o n a l s i g n i f i c a n t commitment t o , and construction o f , commercial power p l a n t s are
.
initiated. ~
With regard t o t h e XSBR, p r e l f m h a r y r e a c t o r designs w e r e WASH-1097 ("The Use of T h o r i d I n Nuclear Pawer Reactors") based upon .., ~
-5t
JJ
the information -supplied
by &IL.
Identified
requiring
Two reactor design concepts-were . :. 1..- --considered -Y a two,_fluid and fertile .=. _-: reactor In which the __fissile i. _ ~.~ _. salts were. separated r-byL .-graphite and a single fluid concept in which ~-_<.' : i. the fissile and fertile salts 'were i completely mixed. This evaluation = problem-areas
resolution .i _.
through -conduct :of an
intensive research and developm-ent program. since the publication I .I 1: -. = .__WASH-1097, all efforts related to the two fluid system have-been discontinued
of
because of mechanical
of processes which would,
if
design problems and the development * *: ., j developed into engineering systems,
permit
the,on-line reprocessing of fuel from single ,f_luid reactors. -.. .~ _ At present, the.*BR concept essentially ,; _ isI .-=-_ ._ _ -._-in_ the initial -- research 7~'~. and development phase, with emphasis ,on the development of basic MSBR Z~ technology. The technology program is centered at ORNLwherefair. 7. _.
.
essentially all research and development on molten salt reactors has --~ .z 1 .-1 ~. _ _ been performed to date. The program is currently funded at a level _', _ .; of $5 million per year. ~Expenditures to date on molten salt reactor -_ .- -_.~ ; : _ . technology both for military and civilian have -. ~, L power applications amounted to approximately $150 million of which approximately $70 million _~. -: _i _. ;.-e has been in support of central station ptier plants. These efforts date . -. -~.I. _~.: .~. ..; .-- 7.-.- .~ _ .in I ~. ~~ back to the 1940's. _ ;
_
_. -.~~
pi .;~~~_
In considering the MSBR for central station power it .- plant application, ...I~:. ~:- = ~~. pi : 7~- :~~ ,- is noted that this concept has several unique and features: â&#x20AC;&#x2DC;. desirable j --~ _* -at-the -same time, it is characterized b-y both complex technological and : _ ~I. - * ,rI .- ; .~ in-__
-6-
\J' p r a c t i c a l engineering problems which are s p e c i f i c t r e a c t o r s and for which s o l u t i o n s have not been developed.
Thus,
this concept introduced major concerns t h a t are d i f f e r e n t in kind and
magnitude from those cammanly associated with s o l i d f u e l breeder reactors.
The development of s a t i s f a c t o r y experimental units and
f u r t h e r consideration of t h i s concept f o r use as a commercial power plant w i l l require r e s o l u t i o n of these as w e l l
88
o t h e r problems which
are common t o a l l reactor concepts. c
As p a r t of t h e AEC's Systems Analysis Task Force (AEC report A
WASH-1098) and the "Cost-Benefit Analysis of t h e U.S.
Breeder
Reactor Program" (AEC r e p o r t s WASH-1126 and WASH-1184).
atudies
w e r e conducted on t h e cost and b e n e f i t of developing another breeder system, "parallel" t o t h e LMFBR.
The c o n s i s t e n t conclu-
s i o n reached i n these s t u d i e s is t h a t s u f f i c i e n t information is a v a i l a b l e t o i n d i c a t e t h a t t h e projected b e n e f i t s from the LMFBR program can support a p a r a l l e l breeder program.
Hawever, these
results are highly s e n s i t i v e t o t h e assumptions on p l a n t c a p i t a l c o s t s with t h e recognition, even among concepts i n which ample experience e x i s t s , t h a t c a p i t a l costs and e s p e c i a l l y small estimated differences i n costs are highly speculative f o r p l a n t s t o he b u i l t 15 or 20 years from now. analyses based upon such c
is ques t ionab l e ould c o n s t i t u t e a major b a s i s f o r
7
making decisions r e l a t i v e t o t h e d e s i r a b i l i t y of a p a r a l l e l breeder effort.
-7-
-*
-_ Lid
Experience
development programs In this country and abroad
in reactor
has demonstrated -that different costs of introducing
organ;itations;
in evaluating the pro.jected
a reactor development-program and 'carrying it forward
to the point of large scale commercial utilization, - different
would arrive at.
estimates of the methods,- scope of development and engineering
efforts,
and the costs and time required toâ&#x20AC;&#x2122;bring
of successful large scale application&d
&at
pro&m
koâ&#x20AC;&#x2122;a
stage
publfc"acceptance.
Based uponthe ARC's experience with other complex reactor development programs, it
is estimated that a total
about 2 billion
dollars
government 1nvestmektu~â&#x20AC;&#x2DC;to
in undiscounted direct
costs* could be required
to bring the molten salt breeder or any parallel
breeder to fruitfon
as
A magnitude of funding up to .&Is reactor. ._ it_ I level could be needed .to estab&h'the necessa&' technology and, ~._. .. . engineering bases ; obtafn the required industrial capability; and _ advance through a series of test facilities, reactor experiments, and _ : _-~, demonstratfon plants to.a commercial MSBRsafe and suitable to aeme
a viable;
commercial
power
, as a major energy option for central
~.
utility
environment.
_
.-~ ~,.~ w
_.
=
station
_
power generation in the
_I -..
.-
.
i ,-
'WASH-ilEi4- -.Updated (1970) Cost-Benefit Analysis of the U.S. Breeder Reactor Program, January 3972W
1
-8-
SUMMARY
11.
The MSBR concept is a thermal spectrum, f l u i d f u e l r e a c t o r which operates on t h e thorium-uranium
f u e l c y c l e and when coupled with
on-line f u e l processing has t h e potential f o r breeding a t a meaningful level.
The marked differences i n t h e concept as compared. I
~
t o s o l i d fueled r e a c t o r s , make t h e MSBR a d i s t i n c t i v e a l t e r n a t e . Although t h e concept has a t t r a c t i v e f e a t u r e s , t h e r e are a number of d i f f i c u l t development problems t h a t must b e resolved: many of t h e s e
are unique t o t h e MSBR w h i l e - o t h e r s are p e r t i n e n t t o any complex r e a c t o r ,spstem
The t e c h n i c a l e f f o r t accomplished s i n c e t h e p u b l i c a t i o n of WASH-1097 and WASH-1098 has i d e n t i f i e d and f u r t h e r defined t h e problem areas; however, t h i s work has not advanced t h e program beyond t h e i n i t i a l Although progress has been made
phase of research and development. in several areas -(e.g.,
reprocessing and improved g r a p h i t e ) , new
problems not addressed in WASH-1097 have a r i s e n which could a f f e c t t h e p r a c t i c a l i t y of designing and operating a MSBR.
Examples aâ&#x201A;Ź
major u n c e r t a i n t i e s relate t o materials of construction, methods f o r c o n t r o l of tritium, and t h e design of components and systems along with t h e i r spec;ial handling, inspection and maintenance equipment. Considerable research and development e f f o r t s are required i n order t o obtain t h e d a t a necessary t o resolve t h e u n c e r t a i n t i e s .
.
'
-9-
Assuming t h a t p r a c t i c a l s o l u t i
o these problems can be found, a
f u r t h e r assessment would have t o be made as t o t h e a d v i s a b i l i t y of proceeding t o t h e next s t a g e of t h e development program. 'In advancing t o t h e next phase, i t would be necessary t o develop a g r e a t l y expanded i n d u s t r i a l and u t i l i t y p a r t i c i p a t i o n and commitment along with
tl
s u b s t a n t i a l increase in'govermnent support.
Such broddened involve-
ment would require an evaluation of t h e %BR
i n terns of already
e x i s t i n g commitments t o o t h e r nuclear power and'high p r i o r i t p energv development e f f o r t s .
-10111.
RESOURCE UTILIZATION
It has long been recognized t h a t t h e importance of nuclear f u e l s f o r
power production depends i n i t i a l l y on t h e u t i l i z a t i o n of t h e n a t u r a l l y occurring fissile U-235; b u t it is t h e more qbundant f e r t i l e materials,
U-238 and Th-232, which w i l l b e t h e major source of nuclear puwer generated i n t h e future.
The b a s i c physics c h a r a c t e r i s t i c s of f i s s i l e
plutonium produced from U-238 o f f e r t h e p o t e n t i a l f o r high breeding gaiqs i n f a s t reactors, and t h e p o t e n t i a l t o expand g r e a t l y t h e u t i l i z a t i o n of uranium resources by &king f e a s i b l e t h e u t i l i z a t i o n of a d d i t i o n a l vast q u a n t i t i e s of otherwise uneconomic low grade ore.
In
a s i m i l a r manner, t h e b a s i c - p h y s i c s c h a r a c t e r i s t i c s of t h e thorium cycle w i l l permit f u l l u t i l i z a t i o n of t h e nation's thorium resources
while a t t h e same time o f f e r i n g t h e p o t e n t i a l f o r breeding i n thermal reactors
.
The estimated thorium resemes are s u f f i c i e n t t o supply t h e world's
electric energy needs f o r many hundreds of y e a r s i f the thorium is used in a high gain breeder reactor.
It ie projected t h a t if t h i s
q u a n t i t y of thorium were used in a breeder r e a c t o r , approximately
1000 Q (1 Q = 10l8 Btu) would be r e a l i z e d from t h i s f e r t i l e material. It is estimated t h a t t h e uranium reserves would a l s o supply 1000 Q*
of energy i f t h e uranium were used i n LMFBRs.
I n c o n t r a s t , only 20 Q
*Uranium recoverable a t U308 p r i c e up t o $lOO/lb.
-11would b e available i f thor
e f e r t i l e material in
an advanced converter reactor because the reactor would be dependent upon U-235 availability for f i s s
inventory make-up
(Note :
conservative estimate is that between 20 and 30 Q w i l l be
electric power generatio
n w and the year 2100.) .-
\
-12-
IV.
U
HISTORICAL DEVELOPMENT OF MOLTEN SALT REACTORS
The i n v e s t i g a t i o n of molten s a l t r e a c t o r s began in t h e late 1940's as p a r t of t h e U.S. A i r c r a f t Nuclear Propulsiod (ANP) Program,
Subsequently,
t h e A i r c r a f t Reactor Experiment (ARE) was b u i l t a t Oak Ridge and i n 1954 i t w a s operated successfully f o r nine davs a t power level@up t o
2.5 MW(th)
and f u e l o u t l e t temperatures up t o 1580.F.
mfxtute of NaF, ZrP4, and UF4.
The ARE f u e l w a s a
The moderator w a s BeO.and t h e pipinn and
vessel were constructed of Inconel.
I n 1956, ORNL began t o study molten s a l t r e a c t o r s f o r a p p l i c a t i o n as c e n t r a l s t a t i o n converters and breeders.
These s t u d i e s concluded thatl
g r a p h i t e moderated, thennal spectrum r e a c t o r s operating on a thoriumuranium cycle were most attractive f o r economic power production,
Based
on t h e technology a t t h a t time, i t was thought t h a t a two-fluid r e a c t o r in which t h e f e r t i l e and f i s s i l e salts w e r e kept s e p a r a t e was required i n order t o have a breeder system.
The s i n g l e f l u i d r e a c t o r , while not
a breeder, appeared simpler i n design and a l s o seemed t o have t h e
p o t e n t i a l f o r l o w power costs.
Over t h e next fw years, ORNL continued t o studp both t h e two f l u i d and s i n g l e f l u i d concepts, and i n 1960 the' design of t h e s i n g l e f l u i d ' 8 MW(th) Holten S a l t Reactor Experiment (MSRE) was begun.
The MSRE was
completed i n -1965 and operated s u c c e s s f u l l y during t h e period 1965 t o 1969.
The %RE
experience is t r e a t e d in more d e t a i l i n a later s e c t i o n .
Concurrent wlth.the
construction
and development on means for
of the MSRE;OR&performed research . proceseinp; molten salt fuels. In-1967
new discoveries were madewhich suggested that a single fluid
reactor.
could be combinedwith continuous on-line fuel .. processing; to becomea I. breeder sys tern. Because of the mechkcal design prob1en-mof the two fluid
concept
and the laboratory-&ale
would permit20n~lke
reprocessing,
development of
it was deierdninkd that 8 shift
emphasis to the single fluih breeder concepk e_._.; system fs being studied at the present. -2
/â&#x20AC;&#x2122;
which-
process&
should be make;
this
in
-14-
litoLTEN SALT BREEDER REACTOR CONCEPT DESCRIPTION
V.
The breeding reactions of the thorium cycle are:
+
2 3 2 ~ n-233~
22
B min.
B
233Pâ&#x201A;Ź4 27.4d e-
,233v
Because of t h e number of neutrons produced per neutron absorbed and t h e mall f a s t f i s s i o n bonus associated with U-233 and Th-232 i n the
thermal spectrum, a breeding r a t i o only s l i g h t l y g r e a t e r than unity is achievable.
I n order t o realize breeding with t h e thorium cvcle i t is
necessary t o remove the bred Pa-233 and t h e various nuclear poisons produced by t h e f i s s i o n process from t h e high f l u x region as quickly a s possible.
The Molten S a l t Breeder Reactor concept permits rapid removal
of Pa-233 and the nuclear elements).
POiSOnS
(e.g.
Xe-135 and the rare e a r t h
The r e a c t o r is a f l u i d fueled system containing uF4 and
ThF4 dissolved i n LiF
- BeF2.
The molten f u e l s a l t flows through a
graphite moderator where t h e nuclear reactions take place.
A side
stream is continuously processed t o remove t h e Pa and rare e a r t h elements, thereby permitting the achievement of a calculated breeding r a t i o of about 1.06.
The MSBR is attractive because of the following:
1.
U s e of a f l u i d f u e l and on-site processing would eliminate t h e
problems of s o l i d f u e l f a b r i c a t i o n and the handling, and
u
-15shipping and reprocessing of spent fuel elements which are
LJ
associated with all other reactor types under active consideration. 2.
*
I
MSERoperation on the thorium-uranium fuel cycle would help conserve uranium and thorium resources by utilizing reserves with high efficiency.
._
thorium; ~, ,
I j
3.
The MSBRis,, projected to have attractive The majorsuncert,ainty in the fuel
fuel cycle costs.
cycle cost is associated
with the continuous fuel processing plant whi.ch has not been I. developed. .. .4.
The safety issues ass&a&d different
.from those of &id
I with the MSBRare genetallv -fuel reactors.
Thus, there
might be safety advantages for the KSBRwhen considering major accidents.
An accurate- assessment of MSBRsafety is
not possible today because of the early state of development. 5.
Like other advanced reactor systems such as the LMFBRand - HTGR, the MSBRwould employ modem'steam technologv for power generation with high thermal efficiencies.
This would reduce
the amount of waste heat to be discharged to the environment.
\
-16-
Selected conceptual design data for a large MSBR, based primarily on design s t u d i e s performed a t ORNL, are given i n Table 1.
There are, however, problem areas associated with t h e MSBR which wst be overcome before the p o t e n t i a l of t h e concept could be a t t a i k e d * These include development of continuous f u e l processing, reactor and processing s t r u c t u r a l materials, tritium c o n t r o l methods, reactor equipment and systems, maintenance techniques, s a f e t y technolonv, and
MSBR codes and standards.
Each of these problem areas w i l l now be
evaluated i n some d e t a i l , which w a s demonstrated by t h e Molten S a l t Reactot Experiment (â&#x20AC;&#x153;2) during its design, construction and operation a t Oak Ridge and the conceptual design parameters presented I n Table I and i n Appendix A. A conceptual flawsheet for t h i s system is shown i n FiRure 1.
Lid
-17-
Table I . Selected Conceptual Design Data for a Large KSBR
Power, W(e)
1000
1 Pawer, W(th)
2240 3500 psia, 100O0F,
Steam System
*
44%net efficiencv 72% 7LIF, 16%BeF2,
~
lng and V e s s e l %dera tor
Sealed Unclad Craphlke
Breeding Ratio Specific F i s s i l e Fuel
ded Doubling T h e , Y e a - -
Core Temperatures,
OF
1050 Inlet, 1300 outlet
-18-
\
\
-19STATUS OF:MSBR .TECUNOLOGY
VI. A.
MSRE- The Reference Point
for Current,Technology .~. I~ Reactor Experiment (MSRE) was begun in .1960 at
The Zlolten Salt
ORNL as part of the Civilian of the experiment molten salt
Nuclear Power Program.
was to demonstrate
power'reactors.
were achieved-during
its
All
first
reactor
in the world
the more signifikt are givenin
In spite salt
objectives
successful
December 1969.. These included
the basic
feasibility
of
of the experiment
operation'from
the distinction
June 1965 to
of becoming the
on U-233. Someof
to operate".solely
dates and statistics
Th? purpose
pertinent
tom the MSRE
Table II.
of the MSRE, there are -many areas
of thesuccess
technology
of molten
which must be expanded and developed in order to
proceed fram this
small
non-breeding
and economic 1000 W(e)
experiment
to a safe,
reliable,
MSBR with-a
30-year life. To illustrate ,. -. this point, some of the most important differences in basic design . and performance characteristics between the MSRE and a conceptual .
lOOO.hW(e) MSBR are given inTable be accomplished size. scale-up
requirements
problems associated
plants
of increasing
provides an appreciation of the
in ,going from the MSRE to a large MSBR. Some with
progressing
high performance
i
Scale-up would logically
through development of reactor
Examination of Table III
commercial,
III.
from a smaJ1 experiment to a
power plant
are not adequately
-20-
LJ
Table 11.
Dates: i
,
. . . . . . . . . . . . . . J u l y 1960 Critical with 235U Fuel . . . . . . . . . June 1, 1965 operation a t f u l l power - 8 MtJ(th) . . May 23, 1966 Complete 6-month run . . . . . . . . . . . . March 20, 1968 End Operation with 235U f u e l . . . . . . . March 26, 1968 Critical with 233U f u e l . . . . . . . . . October 2, 1968 Operation f u l l power with 233U f u e l ... . January 28, 1969 Reactor operation terminated . . . . . . . . December 12, 1969
Design i n i t i a t e d
~
at
Statistics: Hours c r i t i c a l
. . . . . . . . . . . . . 17,655
Fuel loop t i m e c i r c u l a t i n g s a l t (Hrs).
Equiv. f u l l power hours with 235U f u e l Equiv. f u l l power hours with 233U
fuel
..
...
...
21,788 9,005 4,167
-21-
Ld
Table III. Comparison of Selected Parameter8 of the MSREand 1000 MN(e) MSBR-11 > MSREMSBR Y v ; General ThermalPcwer,MW(th) 2250 ' s Electric Pwer, MN(e) 0 1000 Plant lifetime, year8 /: .-, ~ 4- ~_ ~_ ~30 Fuel Proceasing Scheme Off-line, batch -- On-line, continuous : -7 I .: processing:; -I processing ... Breeding Ratio.- ~:_ k88 thSUl 1.0 1,06IT ~(No Th present) Reactor Fuel Salt Moderator
7 7 .-. _ =LiF-BeFZTZrF4-UF4 Lif-BeF2TThF4-UP4 Unclad, '. ~-Unclad, unsealed 'graphite- sealed graphite Reactor Vessel Mater;al Standard Hastelloy-N Modified Hastelloy-N Pwer DeAity,~RW/iiter 2.7 -- 22 - ;=c, ; Exit T&perature, l F -~--~.'- ~. 1210 1300Temperature Rise ACrO88 Core.1OF 40~.. _ ~m.250 Reactor Vessel Height, Ft. 0 -20 Reactor Vessel Diameter, Ft. 5 22 Vessel Design Pressure,-psia'~. 65 ~--' - 75.-. 'I* ,8.3-x lO14 A Neutrons/m‘+sec
Other Componentsand Systems Data m.mz,I-=_ . _; = Number of Primary Jm. ' ~' _.:.. -,: Circuits .I. - * ".y1200--. r. I L,. ~Fuel Salt PumpFl&; j$m Fuel
Lb
s&t
mp
@ad,.
ft,::;
I’ ~48;s
~. _
7LIF-BeF2 Secondary Coolant Salt Numbersof Secondary ~Circufts :-l _ ~: ,._ 1 Secondary-Salt PumpFlaw, Epm -850 -_ Secondary~Salt PumpHead, ft. Seqzmdary~Salt 78 ~0 Nuniber~of Steam Gener&ors Number~of Steam Generator Capacity, MW(th) 0
-*-_ 4 .-16,0& ~. 150. ~~~:
: -
,.
NaF-NaBF4 4. _~ - :~ 20,000 300 i6 . 121
_
A/ Based on information from "Conceptual Design Study of a Single Fluid Molten Salt Breeder Reactor," ORNL-4541,June 1971.
-22-
represented by t h e comparison presented i n t h e Table. -Therefore, it is u s e f u l t o examine a d d i t i o n a l f a c e t s of MSBR technology i n r
more d e t a i l .
B.
Continuous Fuel Processing
- The Key t o Breeding
I n order t o achieve nuclear breeding i n t h e s i n g l e f l u i d NSBR it is necessary t o have an on-line continuous f u e l processing system.
This would accomplish t h e following:
.
I s o l a t e protactinium-233 from t h e r e a c t o r environment so it can decay i n t o t h e f i s s i l e f u e l i s o t o p e uranium-233 before being transmuted i n t o o t h e r isotopes by neutron f r r a d i a t i o n .
.
Remove undesirable neutron poisons from t h e f u e l s a l t and thus improve t h e neutron economy and breeding~performance
of t h e system.
.
Control t h e f u e l chemistry and remwe excess uranium-233 which is t o b e exported from t h e breeder system.
1. Chemical Process Development The Oak Ridge National Laboratory has proposed a f u e l processing scheme t o accomplish breeding i n t h e MSRR, and t h e flowsheet processes involve:
a. Fluorination of t h e f u e l s a l t t o remove uranium as UF6.
w
-23-
b. Reductive extraction salt-with
of protactinium
a mixture of lithium
by contacting the
and bismuth.
c. Metal transfer processing to preferentially remove the . rare earth~fission product poisons which would-otherwise hinder breeding performance. The fuel proc"eseing system shown in Fig. 2 is in an earlv stage . of development at present and this tvpe of system has not been demonstrated on an. operating reactor. required only off-line, from fuel salt..
By comparison, the MSRE
batch fluorination
to recover uranium
I
At this time, the basic chemistry involved in the MSBR processing schemehas been demonstrated in laboratory scale _i experiments. Current efforts at O&k Ridge are being-directed toward development of subsystems incorporating many of the _required processing steps. Ultimately a complete breeder processing~ experiment would be requfred to demonstrate the system with all~the chemical conditions and operational requirements which would be encountered with any MSBR. Not shown on the flowsheet is a separate which would require injecting
~~ .~~
processing svstem
helium bubbles into the fuel
_ .-
c
z
cn
W
0
s
I '
I
I I
I I
I I
a I
E t 4 2
I
.
W YI YI W
0 J
0
c
U
P I I
a ul
1I
I
I
a
1
g
0
---
-ti
J
Y
-24-
a 0
W
I
bl!!
I I I
I I
v) v)
W
a a W LL
U v).
z U
a
? J U
e I W
I
W
z
e a
a 2
0 c V a x
a c
W
2
W
c
3
V
F
0 W
2
z
0 e a
P 0 3 J LL
fn
a
Li
salt,
allowing
~. they -collect
in the reactor system until
thgn to~~ciryzulate
flsslon~product
xenon, and then removing the
bubbles and xenon from the reactor system.. Xenan.ls a-highly undesirable neutron poison which will - ante by capturing neutr& ~. -fuel.
hamper breeding -perf&n-
which -would otherwise breed new
This concept for xenon stripping
was -demonstrated~ln
prlnclple.by~the~MSRE, althouih more efficient-and controllable : stripping syst+ns[ willmbe-desirable ior <the MSBR. The kenon polsonlng in the MSRE_was ~reduceh by a. factor of six by xenon stripping;
the goal for the MSBRls a factor of ten reduction. _ :
2. Fuel Proceselng~Structural Aside fromthe
chemical
Materials
processes
:
themselves,.
development requirements assoclated.wlth for the fuel.processlng presents dif flcult
containment Fterlals
systems. -19 ~artlcular~~llquid
concentrated on using moly)denum
and graphite for contalnlng,blsmuth. -molybdenumand-graphite -are;-difficult
U>fortunately,
both
to use for +uch~englneerlng
T)us, it w&Xl-be necessary to~d&@op.improyed
applications. : ~techniques-for .
fabrication-a@ joining-before ~~ ~~ .posslble ln.the'reproi~filnlF; system._ -_ -'.-
bismuth
compatlbl&lty p?oblems wlth most structural
metals., and present effortsTare
__
there are also
j
.-. .I -i.
their.use
1s.
~.
-,._
\
-26-
A second material8 problem of t h e current fuel processing system
ia the contafnment f o r the f l u o r i n a t i o n s t e p i n which uranium is v o l a t i l i z e d from t h e f u e l salt.'
The f l u o r i n e and f l u o r i d e s a l t
mixture f9 corrosive t o most s t r u c t u r a l materials, including graphite, and present ORNL flowsheets show a "frozen w a l l "
f l u o r i n a t o r which operates with a p r o t e c t i v e l a y e r of frozen f u e l s a l t covering a Aastelloy-N vessel w a l l .
This component
would r e q u i r e considerable engineering development before i t is t r u l y practical f o r use i n on-line. f u l l proceasing systems.
C.
Molten S a l t Reactor Design
- Materials Requirement8
I n concept, t h e molten s a l t reactor core is a comparatively uncomplicated type of heat source.
-
The MSRE reactor core, f o r
example, consisted of a p r h m a t i c s t r u c t u r e of unclad graphite moderator through which f u e l s a l t flowed t o be heated by the self-sustaining chain r e a c t i o n which took place as long as t h e
salt was in t h e graphite.
The e n t i r e reactor i n t e r n a l s and f u e l
s a l t were contained i n vessels and piping made bf Haetelloy-N,
R
high-strength n i c k e l base a l l o y which was developed under t h e A i r c r a f t Nuclear Propqlsion Program.
Over t h e four-year lifetime
of t h e E R E , t h e reactor ' s t r u c t u r a l b t e r i a l s performed satisf a c t o r i r p for the purposes of the experiments although operation of t h e MSRE revealed possible problems w i t h long term use of
t -27-
hi
Hastellay-N in contact with f u e l s a l t s containing f i s s i o n
'
products.
The MSBR a p p l i c a t i o n is mare demanding i n many respect8 than the MSRE, and a d d i t i o n a l d
ork would b e 'required in
several areas of mater
om b e f o r e
s u i t a b l e materials
-
could become available.
1.
The E R E f u e l s a l t was a mixture of 7LiF-BeF-ZrP4-UF4 .0-29.1-5.0-0.9
proportions of
in
,
mole X , respectively. 7
Zirconium f l u o r i d e
included ae p r o t e c t i o n against U02
p r e c i p i t a t i o n should inadvertent oxide contamination of t h e s
epatem occur.
MSRE operation i n d i c a t e d that c o n t r o l of
oxides was not a major problem and thus it is not considered necessary t o include zirconium i n f u t u r e molten s a l t r e a c t o r fuels.
It should a l s o be noted t h a t t h e MSRE f u e l contained whereas the proposed MSBR f u e l s would include f o r breedfnk.
With the
p o s s i b l e except l s f a c t o r i l y throughout the
r r e n t l y proposed by ORNL, would be a i n proportions of 71.7-16-12-0.3
-28-
Lid mole X , respectively.
930.F
This s a l t has a melting point of about
and a vapor pressure of less then 0.1 mm Hg
operating temperature of llSO'F.
a t t h e mean
It a l s o has about 3.3 t i m e s
the density and 10 tbes t h e v i s c o s i t y of water.
Its thermal
conductivity and volumetric h e a t capacity are comparable t o
water.
The high melting temperature is an obvious l i m i t a t i o n f o r a 8ystem using thfs s a l t , and t h e MSBR is limited t o high temperature operation.
fa a d d i t i o n , t h e lithium component
must be enriched in Li-7 i n order t o allow nuclear breeding, s i n c e n a t u r a l l y occurring lithium contains about 7.5% Li-6
.
L i d is undesirable i n t h e MSBR because of its tendency t o .
capture neutrons, thus penalizing breeding performance.
The chemical and phyeical c h a r a c t e r i s t i c s of t h e proposed
MSBR f u e l mixture have been and are being i n v e s t i g a t e d , and they are reasonably w e l l known for w i r r a d i a t e d salts.
The
major unknowns are associated with t h e r e a c t o r f u e l a f t e r i t
has been i r r a d i a t e d .
For example, not enough is-known about
t h e behavior of f i s s i o n products.
The a b i l 3 t y t o p r e d i c t
f i s s i o n product behavior i e important t o p l a n t s a f e t y , operation, and maintenance.
While t h e MSRE provided much.
u s e f u l iafonaation, t h e r e is s t i l l a nee
-29$artGzularly -.-
,.
with regard toi‘the- fate of the- so-called "noble
metal" fission
prqducts’such aS mol~bd&ium, niobiti
others which are cenerated-%n substantial ~_
&ose behaVioriti
and
quix%i~ies and
tkisystem is not well ‘unders toad. _I -.- _ - .--‘I..
.-- A more &%plete'underst~anding of the uhy&cal/chenical --~-characteristics
of the-irradiated
fuel salt is also needed.
As an illustrationof
this point, anomalous powef pulses were ~. observed- d&&g early‘operation oi the ?¶SRE with U-233 fuel \ : ' _*. ._ : &hich were attributed to unusual behavior of helium pas
:
..= ' bubbles as they circulated _,.-
j
through the reactor.
This
~-
-.-
behavior is bkeved
to have been due to some physical and/or _chemical characteristics of the fuel salt which were never
fully
understood.
f&ion
Out-of-reactor
Vork on molten fuel salt
product chemistry is currently _ -~.~~~
under way.
Eventually,
the behaviorSVofthe f&l salt-would need to be confirmed in an .I "~7 operating reactor.
.-The cools& salt_ -in the secondary system of the WRE was of ~. .: .. .. molar composition 66%'L$F-34XBeF2. While this coolant / performed satisfactorily (noldetectable.corrosidn or~reactfon ,-could-be observed In-the in-the secondary system), the salt-has a _
_ high melting..temperature=(8500F) melting. -temperature =(850?F) and la relatively Thtis,
it
expensive.
may not be-the appropriate choice- for power reactors
-30-
for two reasons:
(1) l a r g e r volumes of coolant s a l t w i l l be
used t o generate steam i n the MSBR, and (2) s a l t temperatures i n t h e steam generator should b e l a w enough, i f poesible, t o u t i l i z e conventional steam system technology with feedwater temperatures up to about 550'F.
The operation of MSRE was
less a f f e c t e d by the coolant s a l t melting temperature s i n c e i t dumped t h e 8 MW(th)
of h e a t via an air-cooled radiator,
The high melting temperatures of p o t e n t i a l coolant s a l t s remain a problem.
The current choice i s a e u t e c t i c mixture
of sodium f l u o r i d e and sodium fluoroborate with a molar composition of 8% NaF-92% NaBF
4; t h i s salt melts
at 725.F.
It is c m p a r a t i v e l v inexpensive and has s a t i s f a c t o r v heat
transfer properties ,
However, t h e e f f e c t s of h e a t exchanger leaka between t h e coolant and f u e l salts, and between t h e coolant s a l t and
steam systems, must b e shown t o b e t o l e r a b l e .
The
fluoroborate s a l t is c u r r e n t l y being studied with respect t o both its chemistry and compatibility with H a s t e l l o y - N .
2.
Reactor Fuel Containment Materials , A p r e r e q u i s i t e t o success f o r the MSBR would be the clbilitv t o assure reliable and s a f e containment and handling of molten
LJ j
-51-
u
fuel salts at ail: times dur'ink the life ' w&d
be necessary,; therefore;
of the reactor.
"develop suitable
to
-merit materials -for MSBRapplication
contain-
before plants could be . . .
con&ructed;- -. r
It
.--- I
-5I
A serious question concerning the constituents
compatability
.irradiated
of
of Eastelloy-N with
fuel salt was raised by the
post-
operation examlnation:of the-MSRE-in 1971. Although the MSRE materials performed satisfactorily operation,
for that sqstem during its
subsequent examintition of metal which was exposed to
MSREfuel salt revealed that the alloy had experienced intergranular attack
~of about 0.007 inch.
The attack was
not obvious. tit%1 metal specim&s were tensile
tested, at which
to’cleptha
time &a&-*opened
up as‘the metal was strained.~
examination-revealed that several fission tellurium,
Further
products, including
had-penetrated the metal to depths comparable to
‘those of the cracks.
At the present'time,
the intergrantilar-attack,was Subsequent laboratory
it is thought that
due to the.presence of tellurium.
tests have verified
produce', under~certainconditlons, .' aast&rop N* . .~ '- *, +
that tellurium
interffranular -, ._
.,
-
can
cracking in
.~
.~ Although the limited penetration of cracks presented no problems for the-MSRE, concern &
exists,with
respect to the chemical
-32compatabilitv of Hastellay-N and MSBR f u e l salts when subjected t o t h e more s t r i n g e n t MSBR requirements of higher power density and 30-year l i f e .
If the observed i n t e r m a n u l a r a t t a c k was
indeed 'due t o f i s s i o n product a t t a c k of t h e Hastelloy-N, then t h i s material may not. be s u i t a b l e f o r e i t h e r t h e piping or the v e s s e l s which would b e exposed t o much higher f i s s i o n product concentrations f o r longer periods of t i m e .
E f f o r t s are under
way to understand and explain t h e cracking problem, and t o
determine whether alternate r e a c t o r containment materials
I
should b e a c t i v e l y considered.
I n addition t o t h e i n t e r g r a n u l a r corrosion problem, t h e standard H a s t e l l o y - N used i n t h e MSRE is n o t a u i t a b l e f o r use i n t h e MSBR
because its mechanical p r o p e r t i e s d e t e r i o r a t e t o an unacceptable
level when subjected t o t h e higher neutron doses which would The problem
occur i n the higher power density, l o n g e r - l i f e MSBR.
is thought t o b e due mainly t o impurities i n t h e metal which are transmuted t o helium when exposed t o therrnal neutrons.
The helium
is believed t o cause a d e t e r i o r a t i o n of mechanical p r o p e r t i e s by its presence a t grain boundaries wlthin t h e a l l o y .
It would be
necessary t o develop a modified Hastelloy-N with improved i r r a d i a t i o n r e s i s t a n c e f o r t h e MSBR, and some progress is beinff made i n t h a t direction.
It appears a t t h i s t i m e t h a t s m a l l additions af
c e r t a i n elements, such as titanium, improve t h e i r r a d i a t i o n .
LJ
-33.-
performance
of Hastelloy-N substantially.
Development work on
modified. alloys with imprwed,irradiationresietance 1 currently under way.
is
3. &phi& Additional developmental effort -produce graphites suitable
problems'&
for VSBR&piicatian.
associated with irradiation results
ont&
required to The-first
damageto graphite structures
from fast neutrons.
1s which
Under high neutron doses, -of the
order of 1022neut.rons)~2 , most graphites tend to-become dimen&nally~unstable Based on testsof-
small
commercially available
and grossswelling graphl&sam$es graphites
of the material occurs. at ORNL, the best.
at this time-may be usable to
about 3 x i0 22 neutrons/cm2 , before the core graphite would have ~l_ . This corresponds to roughly a four-year graphite to. be replaced. -~ . lifetime-for the ORNL~reference design. Wile this might be .-i. acceptable, there are still uncertainties about-the fabrication I L and performance of large graphite pieces, and additional work -, would
be required
before
a four-year
life
could be assured
at
the higher MSBRpower densities~how;a being considered.. In anv : 0 --..-1 event, -there would be an obvious~econamfc;.+ncentive~t~~develop longer lived graphites for MSBR application s$nce:a--four-year .I _ life for graphite is- estitiated to represent a fuel &le cost ' _"'_
w
,
.-
-34-
penalty of about 0.2 mills/kw-hr r e l a t i v e t o a system w i t h
W
t h i r t y year g r a p h i t e l i f e .
The second major problem associated with Rraphites f o r MSBR a p p l i c a t i o n is t h e development of a s e a l i n g techniaue which will keep xenon, an undesirable neutron poison, from d i f f u s i n g
i n t o t h e c o r e g r a p h i t e where it can capture neutrons t o the detriment of breeding performance.
While g r a p h i t e s e a l i n g
may not b e necessary t o achieve nuclear breeding i n t h e MSBR, t h e use of sealed g r a p h i t e would certainly enhance breeding performance.
The economic Incentives o r p e n a l t i e s of g r a p h i t e
s e a l i n g cannot be assessed u n t i l a s u i t a b l e s e a l i n g process is developed
.
Sealing methods which have been investigated t o d a t e i n c l u d e p y r o l y t i c carbon coatiag and carbon impregnation.
Thus f a r ,
hawever, no sealed g r a p h i t e t h a t has been t e s t e d remained s u f f i c i e n t l y impermeable t b gas a t MSBR design i r r a d i a t i o n doses, and research and develo
4.
i n t h i s area 1s continuing.
Other S t r u c t u r a l Materials i
In- addition t o t h e
c t u r a l materials requirements f o r t h e
! L
r e a c t o r and f u e l processing systems proper, components and systems which have s p e c i a l materials reauirementa.
Such components as t h e primary h e a t exchangers and
-3L
_
:
steam generators~must function"while different
working fluids.
in contact with two
.~~~
,At the present time, Hastelloy-N is cormldered to be the most promising material
for'use in all salt containment systtie,
including the secondary piping and components. Research to date indiicates that sodium fluoroborate ~_
and
are
HaStellOy-N
colnpat-
/
ible as long as the water content of the _ : fluoroborate low; otherwise, accelerated Corrosion can occur.
is kept
Additional
testing would be needed and is underway. ..
Hastelloy-Nil'ha8 not been adequately evaluated for service under a range of steam condition8 and whether material D.
will
for use in steam generator8 is still
be a suitable not kncm.
Tritium - A Problem of Control
Because of the lithium present
in fluoride
fuel salts,
concept has the inherent problem of generating isotope
it
of hydrogen.
Due primarily
the present WBR
trithun,
Tritium is produced by the- foll&ing ti1
'Li
(nia)
(2)
'Li
(n,aa)
to 'these interactians,
of about 2400 curies/day in a 1000 We
tritlum
a radioactive
reactions:
%. 'H. would be-produced at a rate
MSBR.
This
compares
with about
- 30 40 t o 50 curiesldrtv f o r li.rht water, Raa-cooled,
and f a q t hrecrter
r e a c t o r s , i n which t r i t i u m I s produced primarilv as a l o w v i e l d fisston product.
T r i t i u m productinn in heavy w a t e r reactors of cormarable size
i s Renetallv in the r m E e 3509 t o 5800 curies/dav, due t o neutron i n t e r -
a c t i o n s with t h e deuterium nresent i n heavy w a t e r . -
=
To' f u r t h e r compound tile prnhlem tritim d i f f u s e s r e a d t l v throuzh As a tesri Hactellov-N a t elevate> tempetatuteq. , .
t o prevent tritium C r a m dfffrrstnE throtlRh t h e p i p l n p
t h e VSRQ system (such
conlprinmts
of
heat exchancars) and everrtuallv reachin? t h e
3te.m system where f t mip.ht he discharr.ed
th
t h e environwnt
ttftintd
water. -..
The problem of trCtitm c o n t r o l i n t h e T B P fs hef.n$ studied In d e t s i l a t r)!?NL.
The followins ate b e i n r considered as potentfa1 mathoas
for
trâ&#x201A;Źtium c o n t r o l :
1.
Exchangfnp t h e t r t t i m far an-? hvdronen nresent
f n the
secondat-.
coolant, therebv retaininp t h e t r t t l t i m tn t h e sacnnriarv cnolnnt.
2.
Usinp, coatinEs an metal surfaces f n order t o l n h i b f t t r i t t l t m
diffusion.
..
-37.
L
:
3.
OperatIny, the reActor
with-the
causing, the form&ion
of-tritium
removed fn the off-Eas
salt
more axid*zfn?,~
fluoride
therebv ' I'
which~.couldS>be
svstems.~ .
4;
Usfnpa
different
pracesstnfi
secondant coolant,.
this
coolant
to~remove
erP;., *odium or helium,
and
tritium;
1
..
S.-
Using--another
intermediate
~steam to "Setter"
6.-
with
Of these notentisl
fn either
the heat exchanger or steam
~-
a purge pas between the-walls:
solutions,
the use of an additional
between the secondary -and steam.systems method technically. , equipment required
and
tritium-mm
Using duplex tubing generator
loop between the fluorahorate
intermediate
IS considered
but -it Twould also be extiensive and the loss of thermal
loop
the most effective
due tti the adbitirrnal
efflcfencv.-
;T -.
From
an economfc viewpoint,-
does not. significantlv -tritium
the mostdesYrable
.coinplictite
solutlnn
is one which
-â&#x20AC;&#x2DC;the svstem, :kuch RS exchazipe of
-for.-hydrogen ~present =in--the seconda?+ coolant;-~~Thls
is being investigated
a<part
The tritium
problem may be eased bv the low nertneahilftv
.TI:.e
retention
of-the
.~ _
ORNL~efforts
techninue
I.
onâ&#x20AC;&#x2DC; trttium
I m.
chemistrv.
.I-
._ '~ . -*I
of
.
Oxide coatings which occur
011
steam generator materials i n contact with
steam, and t h i s is a l s o being i n v e s t i g a t e d a t ORNL.
E.
Reactor Equipment and Sfstems Development
While t h e XSBR would u t i l i z e some e x i s t i n g engineering technology from
o t h e r r e a c t o r types, there are s p e c i f i c components and systems f o r which a d d i t i o n a l development work is required.
Such work would have t o take
i n t o account t h e induced a c t i v i t y t h a t those components would accumulate i n t h e MSBR system, i.e.,
also need t o b e developed.
s p e c i a l handling and maintenance equipment wouThe previous discussion has already d e a l t
with a number of these, such as f u e l processing components and systems, but a d d i t i o n a l discussion is appropriate.
1. C0mPonents
As i n d i c a t e d i n the Table 111, a number of components must b e scaled up s u b s t a n t i a l l y from t h e %RE is possible.
s i z e s before a l a r g e MSBR
The development of these l a r g e r components along
with t h e i r s p e c i a l handling and maintenance equipment is probI
ably one of t h e most d i f f i c u l t and c o s t l y phases of HSBR development,
However, reliable, s a f e , and maintainable
components would need t o be developed i n o r d e r for any r e a c t o r system t o b e a success.
The MSBR pumps would l i k e l y be similar i n b a s i c design t o those f o r t h e MSRE, namely, vertical s h a f t , overhung impeller pumps.
-39-Substantial experience has been-gained over the years in the design,-'fabrication
and operation of smaller salt pumps, but
-the size Gould have~to be increased substantially application.
for
The deve.lopment and proof testingof
MSBR
such units
~along with their handling and maintainence equipment ,and test facilities
are expected to be costly and time consuming.
The intermediate heat exchangers for the MSBRmust perform with a minimumof salt inventory in order to improve the breeding performance by lowering the fuel inventory-.
Special surfaces to
enhance heat transfer would help achieve this,
and more studies
would be in order. ~Basedsn previous experience with other systems6 it
is believed-that
theseIunits
cult development and proof testing effort;
The steam generator for MSBRapplications difficult
would require a diffiI
is probably the most
large component to develop since it
represents an
item for Which there has'been almost.no experience to.'date., .is believed that a difficult
m--As discussed -previously,~the.~hij3h'melting candidate secondary koola&si
It
develo$ment~and~proof testing pro-
' -gram would'beneeded to provide reliable
:-
reactor '
and maintainable units. temperatures- of
such ;as'sodium fluoroborate,
present problems -of matching with conventional iteain ~technology~ -At this timeimcentral
station
system
so&r plants utilize j, )
-40-
feedwater temperatures only up t o about 550째F.
Therefore,
coupling a conventional feedwater s y e t e m t o a secondary coolant which freezes at 72S'F design and control.
presents obvious problems i n
It might be necessary t o provide modifi-
cations t o conventional steam system designs t o help resolve t h e problenk.
Because of these factors, a etudy r e l a t e d t o
t h e design of steam generators has been i n i t i a t e d a t Foster Wheeler Corporation.
Control rods and drlves f o r t h e MSBR would also need t o be developed.
The MSRE c o n t r o l rods w e r e air cooled-and operated
i n s i d e Hastellop-N thimbles which protruded dawn i n t o the f u e l
salt.
The MSBR would require more e f f i c i e n t cooling due t o the
higherepower densities involved.
Presumably rods and drives
would be needed which p e m i t t h e rods t o contact'and be cooled by the f u e l salt.
The s a l t valves f o r l a r g e MSBR's
r e p r e s e n t another development
problem, although the f r e e z e valve concept: which was employed successfully i n t h e MSRE could l i k e l y be scaled up i n s i z e and u t i l i z e d f o r m y MSBR applications.
Mechanical t h r o t t l i n g
vdlvek would a l s o be needed f o r t h e MSBR s a l t ~ y s t e m s , even though no t h r o t t l i n g valve w a s used with t h e MSRE.
Mechanical
shutoff valves f o r salt systems, i f required, would have t o be developed
.
.:.
Other componentswhichwould devklcjpment and testing
.gas strippers
require considerable engineering
include-the helium bubble generators and
which are- +roposed.for
i product xenon from&the ~fuel salt. this area is cukently
use in removing the fission
1Research and development in
way as part of the technology
&der
." program at ORNLm " 2.
i
.
:
.~
_-
Systems .The integration
into a complete MSDR
of all-‘required’components
central statiocpower
plant would involve a number ofmsystems for
which development:work is still--required.~. that some components; such asfiumps’and require their
-
own individualsystems
cooling -and lubric’ation.~
control~rod
drives,
would
for functions such as
1
Given the required -ctiponents~and -basic reactor $kima&
It should be noted
materials
and secondaj,
of conktruction,
the
flow systems can be designed.
<
However, the primary flow system would require supporting supporting 'systems ‘systems ‘-'--
~-dor-continuous fuel~$rocessiag,on-line fuel~processing,,"an-line --:i-mof~salt chemistry, active:gasesi
reactor
controland-.safety,
fuel-=dralning--froWevery
1 ->-componentCand -equipment, -afterheat
control during-non-nuclear
fuel~analysis- and control
possible holdup area in
control,
operations.
handling of radio-
and- temperature =~---..
-
-42-
The continuous f u e l processing systems proposed t o d a t e are q u i t e complicated and include a number of subsystems, a l l of which would have t o operate s a t i s f a c t o r i l y , within t h e c o n s t r a i n t s of economics, s a f e t y , and r e l i a b i l i t y .
The e f f e c t s of off-design
conditions on these systems would have t o b e understood so t h a t c o n t r o l would b e possible t o prevent inadvertent contamination of t h e primary system by undesirable materials.
The f u e l d r a i n system is important t o both operation and s a f e t y
s i n c e i t would b e used t o contain t h e molten f u e l whenever a need arises t o d r a i n t h e primary system o r any component or instrument for maintalnence or inspection.
systems would b e required, each with its maintaiaing and c o n t r o l l i n g temperatures.
Thus, additional
OWTI
system f o r
The f u e l s a l t d r a i n
tank would h a v e - t o b e equipped with an a u x i l i a r y cooling Bystem capable of r e j e c t i n g about 18 MJ(th) of h e a t should t h e need a r i e e t o d r a i n the s a l t immediately following nuclear operation.
The secondary coolant system would a l s o require subsystems f o r draining and c o n t r o l l i n g of s a l t chemistry and temperature. a d d i t i o n , the secondary loop might r e q u i r e systems t o c o n t r o l
t r i t i u m and t o handle the consequences of eteam generator or h e a t exchanger leaks.
In
-43The steam system for-the
MSBRmight require a departure from _
conventional designs hu& to the unique &Able& .associated.with using a coolant having a high melting temperature.
Precautions
-would have to be taken against freezing the secondary salt as it travels
through the steam generator; suitable methods for system
startup and control would need to be'lncorporated. proposed the &e of a supercritical
ORNLhas
steam system which operates
at 3500 psia and provides 700째F feedwater by mixing of supercritical steam and high pressure feedwater.
This -system would introduce
major new development requlrem&nts because it differs . ionveiational stesrn cycles:
F.
Makename
- A Difficult
from
Problem for the &BR
Unlike solid fueled reactors in which the primary system contains act&v&ion products and only those fission products which may leak from < defective fuel pins, the *BR would have the bulk of the fission products dispersed throughout the reactor system. Because of this dispersal of . . __ ., .I radioactivity, remote tech&ues would be required for many maintenance functions if the utility
the reactor were
tu
h&e an acceptable plant availab'ility
in
environment, :
'I;he MSREwas d&&&d
.
for remote maintenance of highly radio&tive '.. . &mponents; however, no ajar roainte&&e problems -(rem&ai~ or .repair of _ _large components) were encountered after nuclear opera&on was- initiated.
-44-
Thus, t h e degree t o which t h e MSRE experience on maintenance is applicable to l a r g e commercial breeder r e a c t o r s i s open t o question.
As has been evident i n p l a n t layout work on n u c l e a r ~ c i l i t i e st o d a t e , this requirement f o r remote maintenance w i l l s i g n i f i c a n t l y a f f e c t t h e u l t i m a t e design and performance of t h e p l a n t system.
The MSBR would
r e q u i r e remote techniques and tools f o r inspection, welding and c u t t i n g
of pipes, mechanical assembly and disassembly of components and systems, and removing, transporting and handling l a r g e component items a f t e r they become highly radioactive.
The removal and replacement of core
i n t e r n a l s , such as graphite, might pose d i f f i c u l t maintenance problems because of t h e high r a d i a t i o n levels involved and t h e contamination p r o t e c t i o n which would be requfred whenever t h e primary system is opened.
Another p o t e n t i a l problem is t h e a f t e r h e a t generation by f i s s i o n products
which d e p o s i t i n components such as t h e primary h e a t exchangers. Auxiliary cooling might b e required t o prevent damage when t h e f u e l s a l t is drained from t h e primary system, and a requirement f o r such cooling would z
f u r t h e r complicate inspection and malnteaance operations.
I n some cases, the Inspection and maintenance problems of t h e MSBR could be solved using present technology and p a r t i c u l a r l y experience gained from f u e l reprocessing plants.
However, a d d i t i o n a l technology development
would be required i n o t h e r areas, such
88
remote cuttinp., alignment,
;
Ld
-45-
7
cleaning and-welding of metal*members. Depending to same~-degreeon the ~particular~plant
arrangement~~~ather~special tools and-equipment would
alsohave to be-'designed 'and developed to accomplish inspection and ' -maintenance operations. , In the final-analysis,
.
-
~_
the-developinent,of adequate inspection and
maintenance techniques and proceduresland hardware for the~MSBR< hinges on the success of other facets of the program, such as materials and~coinponentdevelopment,'and eonthe~requirement that adequate care be taken during plant design to assure that all systems ahd components which would -require mainte&nce=over the life maintainable within
G.
--Safe*
the--constraints
- Different
of futility
Issues -d-P for-the
operation.
MSBR
The MS&concept-has certain characteristics advantages:relating
of the plant are indeed
to-safetym~particularly
:which might provide
with~respect
to postulated
major types of accidents currently -considered in licensing â&#x20AC;&#x2DC;aktivities. I Since the fuel would be in a molten form,:cons.ideration of the core meltdown accident~is'not of
-a fuel
spill;-
applicable to'the MSBR. Also,
secondary..criticality'is
in
the event
not-a~problem siuce-this'
is-
athermal re'actor Zystem requiring moderator for nuclear criticality. :__. -~. .~ Oihei safety features'include~ the ofact that Ukprimary operate at lti'pres&e'vith~fuel'salt
system would
that is more~than~lOOO'F~bel& -
-46-
its b o i l i n g point, t h a t f i s s i o n product iodine and strontium form
s t a b l e compounds i n t h e f l u o r i d e salts, and that t h e salts do not react rapidly with air or water.
Because of t h e continuous f u e l processinfz,
t h e need f o r excess r e a c t i v i t y would b e decreased and some of the f i s s i o n products would b e continuously removed from t h e primary system.
A prompt
negative temperature c o e f f i c i e n t of r e a c t i v i t y i s a l s o a c h a r a c t e r i s t i c
of t h e f u e l s a l t .
Safety disadvantages, on t h e o t h e r hand, include t h e very high radioa c t i v e contamination which would be present throughout the primary system, f u e l processing p l a n t , and a l l a u x i l i a r y primary systems such
as the f u e l d r a i n and off-gas systems. systems would have t o b e assured.
Thus, containment of these
A l s o , removal of decay h e a t from f u e l
s t o r a g e systems would have t o b e provided by always ready and r e l i a b l e cooling systems, p a r t i c u l a r l y f o r t h e f u e l d r a i n tank and t h e Pa-233 decay tank i n the reprocessing p l a n t where megawatt q u a n t i t i e s of decay h e a t must b e removed.
The t r i t i u m problem, already discussed, would have t o
I
I
be controlled t o assure safety.
Based on t h e present state of MSBR technology, i t i s not possible t o provide a complete assessment of MSBR safety r e l a t i v e t o o t h e r r e a c t o r s . It can b e s t a t e d , hawever, t h a t the s a f e t y i s s u e s f o r the MSBR are
generally d i f f e r e n t from those f o r s o l i d f u e l reactors, and t h a t more d e t a i l e d design work must b e done before t h e s a f e t y advantages and disadvantages of t h e MSBR could be f u l l y evaluated.
,
w
-47H.
LJ
,
_
Codes, Standards,,~and High Temperature
Codes and standards conjunction be built. currently in. nuclear
with
Design Methods
for MSBR equipment and systems must be developed in ~,
other research
In particular,
and development before _ of construction Ithe materials
large MSBR's can ~. which are
being developed and tested would have to be certified power.plant
for use
applications. . .'
The need for high temperature design-technology _-. .~ .~ as well as for other high temperature systems. under.way a program in support
is a problem for The ARC currently
of the LMFBR which is providing
the MSBR has
materials
data and structural analysis methods for design of systems employing : various temperatures up to 12OO'F. This program would ~~ steel alloys-at .. need to be broade$ed to include MSBR structural materials such as : tiastelloy-N and to include temperatures as high as 1400eR to provide ._ the _design applicable3 to high-temperature, long-term . . technology .-. : :operating conditions which would be expected for MSBR vessels, components, and core structures. es
! -48-
VII.
wxi
INDUSTRIAL PARTICXPATIOX I N THE MSBR PROGRAM
P r i v a t e l y funded conceptual design s t u d i e s and evaluations of HSBR technology were performed i n 1970 by t h e Molten S a l t Breeder Reactor Associates (HSBRA), a study group headed by t h e engineering firm of
Black & Veatch and including f i v e m i d w e s t u t i l i t i e s .
The MSBRA con-
cluded that the economic p o t e n t i a l of t h e MSBR is a t t r a c t i v e r e l a t i v e t o l i g h t water r e a c t o r s , b u t they recognized a number of problems which must b e resolved i n order t o realize t h i s p o t e n t i a l .
Since t h a t t i m e
t h e HSBRA has been r e l a t i v e l y i n a c t i v e .
A second p r i v a t e l y funded organization, t h e Molten S a l t Group, i s headed by Ebasco Services, Incorporated and includes f i v e o t h e r i n d u s t r i a l firms and f i f t e e n u t i l i t i e s .
I n 1971 t h e Group completed an evaluation of t h e
MSBR concept and technology and concluded t h a t e x i s t i n g technology is s u f f i c i e n t t o j u s t i f y construction of an MSBR demonstration p l a n t althdugh t h e performance c h a r a c t e r i s t i c s could n o t be predicted with confidence.
Additional support f o r f u r t h e r s t u d i e s has r e c e n t l y been
committed by t h e members of t h i s group.
I n a d d i t i o n t o these s t u d i e s , manufacturers of g r a p h i t e and Hastelloy-N have been cooperating vith ORNL t o develop improved materials. I
There has been l i t t l e o t h e r i n d u s t r i a l p a r t i c i p a t i o n i n t h e MSBR
Program a s i d e from ORNL subcontractors
.
At t h e p r e s e n t time, t h e r e are
-49td'
two ORNLsubcontracts
Ebasco Services,
in effect.
industrial,
firma who are: p_articip+nta
performing
a deiPfgn and evaluation
currently
performing
Inc.,
utilizing
the Molten,_ Salt_. croup is
ir
-study. 1-~
Foster Wheeler Corporation : i8
on~steam generators
design studies
the
for MSBR
application.
A number of factors indukrial
1.
which tend to.lbnit
can be identified
Vinvolvement at this
time,
namely:
The existing major industrial
and utility
LWR, HTGR, and LMFBR.
2.
The lack of incentive-for fuel
cycle
further
commitment8 to the ., -,
industrial
investment
services, , such as those required
in
supply&
for solid
fire1
rektors.
, 3.
The overwhelming manufacturing solid _
fuel
merit with
4.
reactor8 in contrast with fluid
fueled
The less advanced state deknstrated
and operating
solutious
associated with
experience
the very limited
with
hvolve-
reactors.
of MSBR technology to the major technical
the MSBR concept.
and the lack of problems ~\
-50-
It should be noted that these factors are also relevant considerations
i n establishing the level of governmental support for the MSBR program
which in turn, to some extent, affects the interest of the manufacturing and u t i l i t y industries.
-51-
VK
CONCLtiSIONS
The Molten Salt Breeder Reactor, if successfully
developed and marketed,
'&uld provide a useful supplement to the currently plutonium reactor economy. This concept offers .
Breeding in a thermal spectrum reactor;
.
Efficient
.
developing uranium-
the potential
for:
use of thorium as a fertile~material; ", i Elimination.of 'fuel fabrication and spent fuel shippink;
. High thermal efficiencies. Notwithstanding
these attractive
features,
,'
this assessment has
reconfirmed the existence of major technological and engineering problems affecting feasibility ~-'of: the'concept as a reliable and 'i -~The principal _. concerns economic breeder for the utility industry. include uncertainties
with ..~ materials, _
with methods
of
controlling
I
tritium, and with the design of components and systems along with , , their special handling, inspection and maintenance equipment. Many of thes'e problems are &mpounded ~- by the use of a fluid fission
products
and delayed neutrons are distributed
primary reactor and reprocessing The resolution
of'an IntkUive _ the major effork
throughout
the
systems.
_ of the problems of the MSBR will
the\ conduct research and development program. -kcluded among that &&la
. Proof testing.of .
fuel in which
require
have to bt? accompiishedare:
.
an integrkged reprocessing system;
Development of asuitable
; : containment material;
_. ~I *
;-
-52-
.
Development of a s a t i s f a c t o r y method f o r t h e c o n t r o l and _
I
r e t e n t i o n of t r i t i u m ;
. Attainment of a thorough understanding of
t h e behavior of f i s s i o n
.
products i n a molten s a l t system;
.
Development of long' l i f e moderator graphite, s u i t a b l e for breeder ' applica
.
-
Conceptual d e f i n i t i o n of the engineering f e a t u r e s of t h e many components and s y s tems ;
-
.
Development of adequate methods and equipment for remote inspection, handling, -and maintenance of t h e p l a n t .
The major problems associated with the MSBR are rather d i f f i c u l t In n a t u r e and many are unique t o t h i s concept.
Continuing support of the research
and development e f f o r t w i l l b e required t 6 o b t a i n s a t i s f a c t o r y s o l u t i o n s
t o the problems.
When s i ~ q ~ i f i c a n evidence t i s a v a i l a b l e t h a t demonstrates
realistic s o l u t i o n s are practical, a f u r t h e r assessment could then be made as t o t h e a d v i s a b i l i t y of crdvancini i n t o engineering phase of the development process inclu+g involvement.
t h a t of i n d u s t r i a l
Proceeding with t h i s next s t e p would -a l s o be contingent
upon obtaining a f i r m demonstration of i n t e r e s t and commitment t o t h e concept by t h e power industry and t h e u t i l i t i e s and reasonable assurances t h a t large scale government and i n d u s t r i a l resources can b e made a v a i l
on a continuing b a s i s t o t h i s program i n l i g h t of o t h e r canrmitments t o the commercial nuclear power program and higher p r i o r i t y energy
development e f f o r t s
.
L4
-53p.
.
REFERENCES
_
u
U. S. Atomic l%uergy'Conunission,"The 1967~Supplementto the 1962‘
1.
Rep&t to the Preaidkt
on Civilian
Nuclear Power" USMC Repoit,
- February 1967. U. S. Atomic Energy -&nrkk&m,
2.
%he ~&x-of Tho&un%a Nuclear
Power Reactom" USA&-Rep& WASH-1097,
1963.' -A
U. S. Atomic Ene&y Comaiissibn, "P;'otentkl
, 3.
-Patterns,"
US&i%-Rep& tiASiGlCJ98,
' ” Nuclear Power Gro&h
Decder
U. S. Atomic Energy Cormission~ "Cost-Benefit
'4.
U. S. Breeder ~R&&or%ogrk;"
1970. &naly&
of-
the
USAEC'keiort GASR-1126,“1969.
U. S. Atomic Energy Comi&Lss&n~"Updateid-(1970) Co&B&efit
5.
'
Analysis
of the U. S. Breeder Reactor Program,' USAECReport WASH-1184, . ,. .January 1972.. ~~ .Edison Electric'ktitute,
6.
"'&port bn the EEf Reactor Assessment Panel,"
EEI Publication No. 70-30, 1970. AnntkHekngs
7.
"
on &i&tor$ev~lo~m&t
&mmisS~ionh 1972~Aut&i&g Committee on Atomic
P~o&.m, U. S.'At&ic
Energy
'%&$lation;
Hear&&s before -&Joint 7~ Energy, Congress of the United-States p. 820-830,
P.
Molten salt Breeder Reactor,'
-_..
ORAL-4541, June 1971.' .
\
-54-
.
10
Roaenthal, M. I?.,
et al.; "Advances in t h e Development of Molten-Salt
Breeder Reactor8," A/CONF. 49/P-048, I n t e r n a t i o n a l Conference
011
Fourth United Nations
t h e Peaceful Uses of Atomic Energy,
Geneva, September 6-16, 1971.
11.
(ea.),
Trinko, J. R.
"Molten-Salt Reactor Technology," Technical
Report of t h e Molten-Salt Group, P a r t I, December 1971.
.
12
Trinko, J. R. (ed), "Evaluation of a 1000 We Molten-Salt Breeder Reactor," Technical Report of t h e Molten S a l t Croup, P a r t 11,
.
Navemb er 1971
13.
Ebasco Seroices Iuc., "1000 M W e Molten S a l t Breeder Reactor Conceotual
Design Study," F i n a l Report Task I, Prepared under ORNL subcontract 3560, February 1972.
14
.
"Project f o r I n v e s t i g a t i o n of Molten S a l t Breeder Reactor,''
Final
Report, Phase I Study f o r Molten S a l t Breeder Reactor Associates, Sepfember 1970. Cardwell, D. W. and Haubenreich, P. N.,
"Indexed Abstracts of
Selected References on Molten-Salt Reactor Technology," ORNL-TM-3595,
December 1971. 16
.
Kasten, P. R.,
B e t t f s , E. S. and Robertson, R. C.
, "Design
Studies
of 1000 MW(e) Molten-Salt B k e d e r Reactors ,"ORNL-3996, August 1966 17.
Molten S a l t Reactor Progr'am SPm$annual Reports beginning i n February 1962.
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W
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General Thermal capacity of reactor Gross electrical generation Net electrical output Net overall thermal efficiency Net plant heat rate
2250 MW(t) 1035 MW(e) 1000 MW(e) 44.4% 7690 Btu/kWhr
2250 MW(tj 1035 MW(e) 1000 MW(e) 44.4% 2252 J/kW-SC
Structures Reactor cell, diameter X height Confinement building, diameter X height
72 x 42 ft 134 X 189 ft
22.0 X 12.8 m 40.8 X 57.6 m
20 ft 2 in. 3 in. 75 psi 13ft ' 1412 30 in. 0.13 0.37 22.2 kW/liter 70.4 kW/liter 2.6 X 1014 neuGons cm-' sec-' 8.3 X lo'* neutrons sec-' 3.5 x 10" neutrons cm" sec-' 3.3 x 1014 neutrons cm-? set"
6.77 m 6.1 m 5.08 cm 7.62 cm 5.2 X los N/mz 3.96 m 1412 0.762 m 0.1 3 0.37 22.2 kW/liter . 70.4 kW/liter 2.6 X 1014 neutrons scc-' 8.3 x neutrons cm-l scc-l 3.5 x 1014 neutronscrnm2 scc-l 3.3 x 1014 neutroks cm-3 sw-'
1284' F
969째K
1307"F
982째K
4 years 669,000 Ib 8.5 fps 1074 ft3 1720 ft3 3316 Ib
4 years 304,OOOkg 2.6 m / x c 30.4 m3 48.7 m3
Reactor Vessel ID Vessel height at center (approx) Vessel wall thickness Vessel head thickness Vessel design pressure (abs) Core height > Number of core elements Radial thickness of reflector Volume fraction of salt in central core zone Volume fraction of salt in outer core zone Average overall core power density Peak power density in core Average thermal-neutron flux Peak thermal-neutron flux Maximum graphite damage flux (>SO keV) Damage flux at. maximum damage region (approx) Graphite temperature at max flux region Graphite temperature at maximum graphite damage region Estimated useful life of graphite Total weight of graphite in reactor Maximum flow velocity of salt in core Total fuel salt in reactor vessel Total fuel-salt volume in primary system Fissile-fuel inventory in reactor primary system and fuel processing plant Thorium inventory Breeding ratio Yield Doubling time, compounded continuously. at 80%power factor
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, 1
Primary heat exchangers (for each Thermal capacity. each Tube-side conditions (fuel salt) Tube OD Tube length (approx) Number of tubes Inlet-outlet conditions Mass flow rate Total heat transfer surface Shell-side conditions (coolant salt) Shell ID Inlet-outlet temperatures Mass flow rate Overall heat transfer coefficient (approx)
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I50.000 Ib 1.06 3.2 %/year 22 years
1.06 3.2 %/year 22 years
\
c
,-
U
17.6 X lo6 lb/hr 850 Btu lu" ft-2
(OF)-'
1.73 m 727-894" K 2218 kg/sec 4820 W m-2
(OK)-@
3
A-2
Appendix A (continued) Engineering unit? Primary pumps (for each of 4 units) Pump capacity, nominal Rated head Speed Specific speed impeller input power Design temperature Secondary pumps (for each of 4 units) Pump capacity, nominal Rated head Speed, principal Specific speed Ihpcller input power Design temperature
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Fuel-salt drain tank (1 unit) Outside diameter Overall height Storage capacity Design pressure Number of coolant U-tubes Sue of tubes, OD Number of separate coolant circuits Coolant fluid Under normal steady-state conditions: Maximum heat load Coolant circulation rate Coolant temperatures, in/out Maximum tank wall temperature Maximum transient heat load
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Fuel-salt storage tank (1 unit) Storage capacity Heat-removal capacity Coolant fluid Coolant-salt storage tanks (4units) Total volume of coolant salt in systems Storage capacity of each tank Heat-removal capacity. f m t tank in series Steam generators (for each of 16 units) Thermal capacity Tube-side conditions (steam at 3600-3800 psi) Tube OD Tube-sheet-to-tube-sheet length (approx) Number of tubes I nletbutle t temperatures Mass flow rate Total heat transfer surface Shell-side conditions (coolant salt) . Shell ID Inlet-outlet temperatures Mass flow rate Apparent overall heat transfer coeffKient range
International systcm.units*
16,000gpm 150 ft 890 rpm . 2625 r p m ( g p ~ n ) ~ ' ~ / ( f t ) ~ * ~ ~ 2350 hp 1300"F
1-01 m'/sec 45.7 m 93.2 radianshc 5.321 radian~/sec(m~~sec)~.~/(m)~~~~ 1752 kW 978°K
20,000 gpm
1.262 m3/sec 91.4 m 124.6 radianJsec 4.73 radianJsec(m' /sec)O 4 /( m)O *7s 2310 kW 978°K
300 ft 1190rpm 2330 rpm(gpm)0'5/(ft)0'7s 3100 hp 1300°F
4.27 m 6.71 m 70.8 m3 3.79 X los N/m2
14 ft 22 ft 2500 ft3 55 psi 1500
9'
1500 1.91 cm
in.
40 'LiF-BeFs
18 MW(t) 830 gpm 900-1050°F -1260°F 53 Mw(t)
18 MW(t) 0.0524 m3/sec 755-839°K -955°K 53 MW(t)
2sOOft3 -
70.8 m' 1 MW(t)
1 MWW Boiling water
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Boilingwater
8400 ft' 2100 ft3 400 Irw
237.9 m3 59.5 m' 400 kW
120.7 MU'( t)
120.7 MW(t)
y in.
1.27 cm 23.3 m 393 644-811°K 79.76 kg/sec 365m' .
76.4 ft 393 700- 1000"F 633,000 lbbr 3929 ft2 .
1.5 ft
1150-850°F 3.82 X lo6 lb/hr 490-530 Btu hr" 11-
CF)"
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0.457 m 894-727% 481.3 Wsec 2780-3005 W m-2 ("K)'t
A-3
Appendix A (cont h u e d ) .~
Engineering unit.@ Steam reheaters (for each of 8 units) Thermal capacity Tube-side conditions (steam at 5sO psi) Tube OD Tube length Number of tubes Inlet-outlet temperatures Mass flow rate Total heat transfer surface Shell-side conditions (coolant salt) Shell ID Inlet-outlet temperatures Mass flow rate Overall heat transfer coefficient Turbine-generator plant (see “General” above) Number of turbme-generator units Turbine throttle conditions Turbine throttle mass flow rate Reheat steam to IP turbine Condensing pressure (abs) Boiler feed pump work (steamturbine-driven), each of 2 units Booster feed pump work (motordriven), each of 2 units Fuehalt inventory, primary system Reactor Core zone 1 Core zone 11 Plenums, inlets, outlets 2-in. annulus Reflectors Primary heat exchangers Tubes Inlets, outlets Pump bowls Piping, including drain line
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International system unitsb
36.6 W ( t )
36.6 MW(t)
36 in. 30.3 ft 400 650- 1000”F 64 1,000 Ib/hr 2381 Et2
1.9 cm 9.24 m 400 616-811°K 80.77 kg/sec 221.2 m2
21.2 in. 1150-850°F 1.16 X IO6 Ib/hr 298 Btu hr’’ ft’)
0.54 m 894-727°K 146.2 kg/sec 1690 W m-2 (OK)-*
3500 psia. 1000°F 7.15 X lo6 lb/hr 540 psia. 1000°F 1.5 in. Hg 19.700 hp
C
1 24.1 X IO6 N/m2, 811°K 900.9 kg/sec 3.72 X lo6 N/m2. 81 1°K 5,078 N/m2 14,690 kW
- 6200hp
4620 kW
290 ft’
8.2 d 10.8 m5 6.2 m3 3.8 m’ 1.4 m3
382 ft’ 218 ft’ 135 ft’ 49 ft’ 269 ft’ 27 ft3 185 ft3 145 ft’
Off-gas bypass loop
7.6 m’ 0.8 m’
5.2 4.1 0.3 0.3
m’
m3 m3
m3 m3
Tank heels and miscellane Total enriched salt in primary system Fuet-processing system (Chemical Treatment
48.7
Cyde time for salt inventory Heat generation in sa!t to processing plant
LJ
propertie Components Compositioh . Molecular weight (approx) Melting temperature (approx) Vapor pressure at 1150°F (894.3%) Wan’) = 3.752 6.68 x 10*r P (Uft’) = 235.0 0,02317t(%“)
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At 1 3 W F (978°K) At 1175°F (908°K) At 1050°F (837K)
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204.9 Ib/ft’ 207.8 lb/ft3 210.7 lb/ft3
3283.9 Wm3
A-4
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Bd
Appendix A (continued) Engineering unitP Viscosity:d p (centipoises) = 0.109 exp [4090/T('K)];p (Ib it-' h-l) = 0.2637 exp [7362/T("R)] At 1300°F (978°K) At 1 175°F (908°K) At 1050°F (839°K) Heat capacitf (specific heat, cp) Thermal conductivity (k)f At 1300°F (978°K) At 1175°F (908°K) At 1050°F (839°K)
17.3 Ib hr-' ft-' 23.8 lb hr-' ft" 34.5 Ib hr-' ft-' 0.324 Btu Ib" PF)"
Design properties of cootant salt Component s Composition Molecular weight (approx) Melting temperature (approx) Vapor pressure:g log P (mm Hg) = 9.024 - 5920/T("K) At 850°F (727°K) At 1150°F (894°K) Density:=p (g/cm3) = 2.252 - 7.1 1 X 104t ("C); p Curt3) = 141.4 0.0247r (OF) At 1150°F (894°K) At 1000"F(81l0K) At SSO" F (727"K) Viscosity:d p (centipoises) = 0.0877 exp [2240/T (OK)];p (Ib, ft" hr-l) = 0.2121 exp [4032/T('R)) At 1150"F(894°K) At 1000°F (811°K) At 850°F (727°K) Heat capacityh (specific heat, cp) Thermal conductivity (k)' At 1150"F(894°K) At 1M)O"F(811OK) At 850°F (727°K)
* 4%
International system unit#
0.007 N sec m-2 0.010 N sec rn3 0.015 N sec m-2 1.357 J g" PIC)''
0.69 Btu hr-' Po-' ft" 0.71 Btu hr" ("F)" ft-' 0.69 Btu hr" ("F)-' ft-'
1.19 W m-' CK)-' 1.23 W m-' VK)" 1.19 W m-' PK)"
NaBF4-NaF 92-8 mole % 104 725°F
NaBF4-NaF 92-8 mole % 104 658°K
8 m m Hg 252 mm Hg
1066 N/m2 33,580 N/m2
113.0 lb/ft3 116.7 Wft3 120.4 lblft'
1811.1 kg/m3 1870.4 kg/m3 1929.7 kg/m3
2.6 Ib ft-' hr" 3.4 lb ft-' hr-' 4.6 lb ft-' hr-' 0.360 Btu lb-' (OF)"
0.001 1 N sec m-2 0.0014 N sec in-2 0.0019 N sec m-2 1507 J kg-' (OK)-'
* 4%
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Design properties of graphitei Density. at 70°F (294.YK) Bending strength Modulus of elasticity coefficient Poisson's ratio Thermal expansion coefficient Thermal conductivity at 1200°F. unirradiated (approx) Electrical resistivity Specific heat At 600°F (588.8"K) At l 2 W F (922.0"K) Helium permeability at STP with sealed surfaces
,
0.23 Btu hr" 0.23 Btu hr" 0.26 Btu hr"
'
* 2% ft-' ft" ft"
(OF)"
(OF)" (OF)"
115 lb/ft3 4000-6000 psi 1.7 X lo6 psi 0.27 2.3 X 10df'F 18 Btu hr-' (OF)" ft-' 8.9 X 104-9.9X
lo4 n-cm
0:33 Btu lb-' (OF)-' 0.42 Btu lb" (OF)-' 1 x 10" cm2/sec
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* 2%
0.398 W m-l (OK)" 0.398 W an-' ( O K ) - ' 0.450 W m-' (OK)-' 1843 kg/m3 28 X 106-41 X lo6 N/m2 11.7 X lo9 N/m2 0.27 1.3 X 10dPK 31.2 W m-I (OK)" 8.9 X 104-9.9 X 1380 J kg" (OK)" 1760 J kp-' (OK)" 1 x IO* cm2/sec
lo4 n t m
A-5
Appendix A (continued) Engineering unitP Design properties of Hastelloy Nk Density At 80°F (300°K) At 1300°F (978°K) Thermal conductivity At 80°F (300°K) At 1300°F (978°K) Specific heat At 80°F (300°K) At 1300°F (978°K) Thermal expansion At 80°F (300°K) At 1300"F(978"K) Modulus of elasticity coefficient At 80°F (300°K) At 1300°F (978°K) Tensile strength (approx) At 80°F (300°K) At 1300°F (978°K) Maximum allowable design stress C t 80"F (300°K)
At 1300°F (978°K) Melting temperature
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International system unitsb
557 lblft3 541 lb/ft'
8927 kg/m3 8671 kg/m3
6.0 Btu hr-' CF)" ft-' 12.6 Btuhr" ("F)" ft-'
10.4 W m-' (OK)-' 21.8 W m-I (OK)-'
0.098 Btu lb-' (OF)-' 0.136 Btu lb-' (OF)"
410 J kp-' ("IC)-' 569 J kg-' (OK)-'
5.7 xFP'OI 9.5 x 10"PF
3.2X 1O"PK 5.3 X 10df'K
31 X 25 X
lo6 psi lo6 psi
214 X lo9 N/m2 172 X lo9 N/m2
115,000psi 75.000 psi
793X 106N/m2 517 X lo6 N/m2
25,000 psi 3500 psi 2500"F
172 X IO6 N/m2 24 X lo6 N/m2 1644°K
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'English engineering units as used in MSR literature. bMeter-kilogram-second system. Table closely follows International System (SI). See Appendix C for conversion factors from engineering to SI units.