3 q45b 03bL387 3 T
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NUCLEAR CHARACTERISTICS OF SPHERIC HOMOGENEOUS, TWO-REGIONl MOLTEN-FLUORIDE-SALT REACTORS
L. G. Alexander D. A. Carrison
H. G. MacPherson J. T. Roberts
'
i
.
ORAT U N I O N CARBIDE COR for the
U.S. A T O M I C E N E R G Y
'
TlON , I SSlO
P r i n t e d i n USA.
Price
75
A v a i l a b l e from t h e
O f f i c e o f T e c h n i c a l Services Department of Commerce Washington 25, D.C.
LEGAL NOTICE T h i s report w a s prepared as on account of Government sponsored work.
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nor t h e Ccmmission, nor any person a c t i n g on b e h a l f of t h e Commission:
A.
Makes a n y warranty or representation, expressed or implied, w i t h respect t o t h e accurocy, completeness, ony
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ORNL-2751 Reactors-Power TID-4500 (14th ed.)
Contract No. W-7405-eng-26
R E A C T O R P R O J E C T S DIVISION
NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS, TWO-REGION, MOLTEN-FLUORIDE-SALT REACTORS
L. G. D. A.
Alexander Carrison
H. G.
MacPherson
J. T. Roberts
D A T E ISSUED
OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated b y UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION
3 445b 0363387 3
.
.
CONT ENT S
......................................................................................................... PRIOR WORK ....................................................................................................... METHOD OF CALCULATION .................................................................................. HOMOGENEOUS REACTORS FUELED WITH U 2 3 5 ...................................................... Initial States ...................................................................................................... Neutron Balances and Other Reactor Variables .......................................................... HOMOGENEOUS REACTORS FUELED WITH U 2 3 3...................................................... NUCLEAR PERFORMANCE OF A REFERENCE DESIGN REACTOR .............................. COMPARATIVE PERFORMANCE ............................................................................. ACKNOWLEDGMENTS ........................................................................................... ABSTRACT
1 2
4 4 4
7
9 14
18
18
...
Ill
NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS, TWO- R EGION, MOLT EN- FLUOR1DE-SALT REACTORS
L. G.
Alexander
D. A. Carrison
H. G.
MacPherson
J. T. Roberts
ABSTRACT T h e use
of
o m o l t e n - s a l t f u e l makes p o s s i b l e the production
steom w i t h o nuclear reactor operating o t l o w pressure.
of high-pressure, superheated INOR-8
T h e corrosion r e s i s t a n c e of the
series o f nickel-molybdenum a l l o y s appeors t o be s u f f i c i e n t t o guarantee reactor component l i f e t i m e s of
10 t o 20 years.
Proposed continuous fuel-processing methods show promise of r e d u c i n g
f u e l - p r o c e s s i n g c o s t s t o n e g l i g i b l e levels.
With
U233 o s
in b o t h core and blonket, i n i t i a l regeneration r a t i o s up t o
the f u e l and T h 2 3 2 0 s the f e r t i l e material
1.08 c o n be obtained a t c r i t i c a l mosses
600 kg. T h e corresponding inventory for o 600-Mw(th) central s t o t i o n power reoctor i s 1300 kg. With U235 a s the fuel, u233 i s produced, ond i n i t i a l regenerotion r a t i o s i n excess of 0.6 con be obtained w i t h c r i t i c a l mosses o f l e s s t h a n 300 kg. The corresponding c r i t i c a l i n v e n t o r i e s for 600-Mw(th) central s t o t i o n power reactors ore 600 k g or less, depending on the thorium looding. It i s concluded t h a t homogeneous, molten-salt-fueled reactors ore competiless than
i n i t i o l l y obout
t i v e i n regard t o n u c l e a r performance w i t h present solid-fuel reactors, and they may be economic o l l y superior because o f l o w e r f u e l and fuel-processing costs.
Molten f l u o r i d e s a l t s p r o v i d e t h e b a s i s of a new f a m i l y o f l i q u i d - f u e l e d power reactors. The range o f s o l u b i l i t y o f uranium and thorium compounds in the s a l t s makes t h e system f l e x i b l e and a l l o w s the consideration o f a v a r i e t y o f reactors. Suitable s a l t mixtures have m e l t i n g p o i n t s in t h e range 850 t o 950째F and are s u f f i c i e n t l y compatible with known a l l o y s t o assure l o n g - l i v e d components, i f t h e temperature i s k e p t below 1300'F. A s may be seen, molten-salt-fueled reactor systems tend to operate n a t u r a l l y in a temperature region s u i t a b l e for modern steam plants; they have the further advantage that they a c h i e v e these temperatures without pressurization o f t h e molten salt. T h e molten-salt system, for purposes other than e l e c t r i c power generation, i s n o t new. I n t e n s i v e research and development over the past n i n e years
under ORNL sponsorship has provided reasonable answers t o a majority o f t h e obvious d i f f i c u l t i e s . One o f t h e most important advances has been the development o f methods for h a n d l i n g s a l t m e l t s a t h i g h temperatures and m a i n t a i n i n g them a t temperatures above t h e l i q u i d u s temperatures. Information on the chemical and p h y s i c a l properties o f a wide v a r i e t y o f molten s a l t s has been obtained, and methods were developed for t h e i r
manufacture, purification, and handling t h a t are i n u s e on a production scale. It has been found that the simple i o n i c s a l t s are stable under r a d i a t i o n and that they suffer no deterioration other than the b u i l d u p o f f i s s i o n products. T h e molten-salt system has the usual b e n e f i t s a t t r i b u t e d t o f l u i d - f u e l e d systems. T h e p r i n c i p a l advantages claimed over solid-fuel systems are:
(1)
t h e lack o f r a d i a t i o n damage that can l i m i t fuel burnup; (2) t h e avoidance of the expense of fabricating new fuel elements; (3) the continuous removal of f i s s i o n products; (4) a h i g h negative temperature c o e f f i c i e n t o f r e a c t i v i t y ; and (5) the avoidance o f t h e need for excess r e a c t i v i t y , since makeup fuel can be added as required. The l a s t two factors make p o s s i b l e a reactor without control rods t h a t automatically a d j u s t s i t s power i n reThe sponse t o changes of the e l e c t r i c a l load. l a c k o f excess r e a c t i v i t y can l e a d t o a reactor t h a t i s safe from nuclear power excursions. In comparison w i t h aqueous systems, the moltens a l t system has three outstanding advantages: it a l l o w s high-temperature operation a t a low pressure; e x p l o s i v e r a d i o l y t i c gases are not formed; and thorium compounds are s o l u b l e in the salts. Compensating disadvantages are the high m e l t i n g
p o i n t s o f t h e s a l t s and poorer neutron economy; t h e importance o f these disadvantages cannot b e assessed properly w i t h o u t further experience. Probably t h e most outstanding c h a r a c t e r i s t i c o f the molten-salt systems i s t h e i r chemical f l e x i b i l i t y , that is, t h e w i d e v a r i e t y of m o l t e n - s a l t s o l u t i o n s which are a v a i l a b l e for reactor use. In t h i s respect, t h e m o l t e n - s a l t systems are p r a c t i c a l l y unique, and t h i s i s t h e e s s e n t i a l advantage w h i c h they enjoy over t h e uranium-bismuth systems. Thus t h e m o l t e n - s a l t systems are not to be thought o f i n terms o f a s i n g l e reactor - rather, they are t h e b a s i s for a new c l a s s o f reactors. Included i n t h i s c l a s s are a l l the embodiments w h i c h comprise t h e w h o l e o f solid-fuel-element technology: U235burners, thorium-uranium thermal converters or breeders, and thorium-urani um f a s t converters or breeders. Of p o s s i b l e short-term i n t e r e s t i s t h e U 2 3 5 burner. Because o f the inherently h i g h temperatures and because there are
o f l o w neutron absorption, h i g h s o l u b i l i t y o f uranium and thorium, and chemical inertness. In general, t h e c h l o r i d e s h a v e lower m e l t i n g points, b u t they appear to b e l e s s s t a b l e and more corr o s i v e than t h e fluorides. T h e f l u o r i d e systems appear to b e preferable for use i n thermal and epithermal reactors. Many mixtures have been investigated, m a i n l y a t ORNL and a t Mound Laboratory. T h e p h y s i c a l properties o f these mixtures, i n so far a s t h e y are known, have been tabulated, and t h e r e s u l t s o f e x t e n s i v e phase 3 s t u d i e s h a v e been reported. L i t h i u m - 7 h a s an a t t r a c t i v e l y l o w capture cross section, 0.0189 barn a t 0.0759 ev, but Li6, which comprises 7.5% o f t h e natural mixture, has a capt u r e cross section o f 542 barns a t t h e same energy. T h e cross sections at 0.0759 e v and 1 1 5 O O F for several l i t h i u m compositions are compared below w i t h t h e cross sections o f sodium, potassium, rubidium, and cesium.
no fuel elements, t h e fuel c o s t i n the s a l t system i s o f t h e order o f 1 m i l l / k w h r i n graphite-moderated, molten-salt-fueled reactors. T h e present technology suggests that homogeneous converters u s i n g a base s a l t composed o f
Cross Section Element Lithium
0.1% L i 6 0.01% L i 6 0.001% L i 6 O.OOOI% L i 6
BeF, and e i t h e r L i 7 F or N a F and u s i n g UF, for fuel and ThF, for a f e r t i l e m a t e r i a l are more suit-
a b l e for e a r l y reactors than are graphite-moderated or plutonium-fueled systems. The chief virtues
of t h e homogeneous converter reactor are that i t i s based on w e l l - e x p l o r e d p r i n c i p l e s and that the use o f a simple fuel c y c l e should lead to l o w fuel c y c l e costs. With further development, t h e same base s a l t (using Li7F) can be combined w i t h a graphite moderator i n a heterogeneous arrangement to prov i d e a s e l f - s u s t a i n i n g Th-U233 system w i t h a breeding ratio o f about 1. T h e c h i e f advantage of t h e molten-salt system over other l i q u i d systems i n pursuing t h i s o b j e c t i v e is, a s h a s been mentioned, that i t i s t h e o n l y system i n w h i c h a solub l e thorium compound can be used, and thus the problem o f slurry h a n d l i n g i s avoided. P R I O R WORK
T h e a p p l i c a b i l i t y of molten s a l t s t o nuclear reactors has been a b l y d i s c u s s e d by Grimes and others.’12 T h e most promising systems are those comprising t h e f l u o r i d e s and c h l o r i d e s o f t h e a l k a l i metals, zirconium, and beryllium. These appear t o possess t h e most d e s i r a b l e combination
2
(born s)
Sodium Potassium Rubidium Cesium
0.561 0.073 1 0.0243 0.0194 0.290 1.130 0.401 29
T h e capture c r o s s s e c t i o n s o f t h e l i g h t e r elements at higher energies presumably stand i n approximately t h e same r e l a t i o n a s a t thermal. It may be seen that p u r i f i e d Li7 h a s an a t t r a c t i v e l y l o w c r o s s section i n comparison w i t h the c r o s s s e c t i o n s o f other a l k a l i m e t a l s and that sodium i s t h e n e x t best a l k a l i metal. T h e sodium-zirconium f l u o r i d e system has been e x t e n s i v e l y studied at ORNL.3 A e u t e c t i c cont a i n i n g about
42 mole % ZrF,
melts at
910OF.
‘ W . R . Grimes, ORNL CF-52-4-197, p 320 f f ( A p r i l
1952) ( c l a s s i f i e d ) .
2W. R. Grimes, D. R. Cuneo, and F. F. Blankenship, i n R e a c t o r Handbook, ed. b y J. F. Hogertan and R . Grass, v o l 2, sec 6, p 799, AECD-3646 (May 1955).
c.
3J. A. L a n e , H. G. MacPherson, and F. Maslan (eds.), Fluid Fuel Reactors, p 569, Addison-Wesley, Reading,
Mass.,
1958.
Small additions o f UF, lower the melting p o i n t appreciably. A fuel o f t h i s t y p e was s u c c e s s f u l l y used i n the A i r c r a f t Reactor Experiment (ARE)., lnconel, a n i c k e l - r i c h alloy, i s reasonably res i s t a n t t o corrosion by t h i s f u e l system at 15OOOF. Although long-term data are lacking, there i s reason t o expect the corrosion r a t e a t 120OOF t o be s u f f i c i e n t l y l o w t h a t lnconel equipment would l a s t several years. However, w i t h regard t o i t s u s e i n a centralstation power reactor, the sodium-zirconium fluor i d e system has several serious disadvantages. T h e sodium capture c r o s s section i s l e s s favorable than t h e L i 7 cross section. I n addition, there i s t h e so-called “snow” problem; that is, ZrF, tends t o evaporate from the fuel and c r y s t a l l i z e on surIn comparison w i t h faces exposed t o t h e vapor. t h e l i t h i u m - b e r y l l i u m system d i s c u s s e d below, the sodium-zirconium system has i n f e r i o r heat transfer F i n a l l y , t h e expectaand c o o l i n g effectiveness. t i o n a t Oak Ridge i s that t h e INOR-8 a l l o y s w i l l prove t o be as r e s i s t a n t t o the b e r y l l i u m s a l t s as t o t h e zirconium s a l t s and t h a t there i s therefore n o compelling reason for s e l e c t i n g the sodiumz ircon ium system. T h e capture cross section o f b e r y l l i u m appears t o be s a t i s f a c t o r i l y low at a l l energies. A phase diagram for t h e system L i F - B e F , has recently been published.’ A mixture c o n t a i n i n g 31 mole % BeF, reportedly l i q u e f i e s a t approximately 98OoF. Substantial concentrations o f ThF, i n the core f l u i d may be obtained by b l e n d i n g t h i s mixture with A temperature diagram t h e compound 3Li F.ThF,. for the ternary system has been published.6 T h e
l i q u i d u s temperature along the j o i n appears t o l i e below 930°F for mixtures c o n t a i n i n g up t o 10 Small a d d i t i o n s o f UF, to any o f mole % ThF,. these mixtures should lower t h e l i q u i d u s temperature somewhat. T h e ARE was operated w i t h a molten-fluorides a l t fuel i n November 1954. T h e reactor had a moderator c o n s i s t i n g o f b e r y l l i u m o x i d e blocks. T h e fuel, which was a mixture o f sodium fluoride, zirconium fluoride, and uranium fluoride, flowed through the moderator i n lnconel tubes and was pumped through an external heat exchanger by
,1bid,
673.
’!bid., p 573. 6 1 b i d , p 579.
means o f a centrifugal pump. T h e reactor opIt was d i s erated a t a peak power o f 2.5 Mw. mantled after carrying out a scheduled experimental program. In 1953 a group o f ORSORT students, under the leadership o f J a r ~ i s , ~ i n v e s t i g a t e d the applicab i l i t y o f molten s a l t s t o package reactors. More recently, another ORSORT group l e d by D a v i e s prepared a v a l u a b l e study o f the f e a s i b i l i t y o f
molten-salt U235burners for central-station power production.8 F a s t reactors based on t h e U 2 3 8 - P u
c y c l e were studied by Addoms et al. of MIT and by an ORSORT group l e d by B ~ l m e r . ~Both groups concluded that it would be preferable t o u s e molten chlorides, r a t h e r than the fluorides, because o f the r e l a t i v e l y h i g h moderating power o f t h e f l u o r i n e nucleus, although it was recognized t h a t the chlor i d e s are probably inferior w i t h respect t o corrosion and radiation s t a b i l i t y . Bulmer et al. a l s o pointed out that it would be necessary t o u s e purif i e d C137 on account of t h e ( n , p ) reaction exh i b i t e d by C135. Because o f t h e disadvantages o f
t h e c h l o r i d e systems and, further, because the technology of handling and u t i l i z i n g neptuniumand plutonium-bearing s a l t s i s largely unknown, i t was decided t o postpone consideration o f c h l o r i d e s a l t reactors. I n 1953 an ORSORT group l e d by Wehmeyer” analyzed many o f the problems presently under study. The proposals set forth i n t h a t report have influenced t h e present program. A study by Davidson and Robb o f K A P L ” has a l s o been helpful. B o t h studies considered the p o s s i b i l i t y o f u s i n g thorium i n a U 2 3 3 conversion-breeding c y c l e a t thermal or near thermal energies. A recent conceptual design study12 o f a 240-Mw ( e l e c t r i c a l ) centra I - s t a t i o n mol ten-sal t-fuel ed reactor was used as a b a s i s for examining the economics and f e a s i b i l i t y of a reactor using moltens a l t fuel. A n attempt was made t o keep the 7T. Jarvis et al., ORNL CF-53-10-26 (August 1953) (classified). 8R. W. Davies et al., ORNL CF-56-8-208 (August 1956) (classified).
9J. Bulmer et a[., Fused Salt Fast Breeder, ORNL CF-56-8-204 (August 1956). ’OD. B. Wehrneyer et al.. Study of a F u s e d Salt Breeder Reactor for Power Production, ORNL CF-5310-25 (September 1953). ”J. K. Davidson and W. L. Robb, A Molten-Salt Thorium Converter f o r Power Production, KAPL-M-JKD10 (Oct. 15, 1956). 12Lane, MacPherson, and Maslan, op. cii.,
p
681.
3
technology and t h e processing scheme as simple as possible. METHOD OF CALCULATION Reactor c a l c u l a t i o n s were performed by means o f
the UNlVAC program O C U S O ~a , modification ~~ of the Eyewash pr0gra1n.l~ Ocusol i s a 31-group, mu It i region, spherical Iy symmetric, age-di ffusion code. T h e group-averaged cross sections for the various elements o f i n t e r e s t that were used were based on the l a t e s t a v a i l a b l e data.” Where data were lacking, reasonable interpol at ions bo sed on resonance theory were used. T h e estimated cross sections were made t o agree w i t h measured resonance integrals where available. Saturations and Doppler broadening o f the resonances i n thorium as a function o f concentration were estimated. T h e molten s a l t s may be used as homogeneous moderators or simply as fuel carriers i n heterogeneous reactors. Although graphite-moderated heterogeneous reactors h a v e certain potential advantages, their technical f e a s i b i I i t y depends upon the c o m p a t i b i l i t y o f fuel, graphite, and metal, which has not as y e t been established. For t h i s reason, the homogeneous reactors, although inferior i n nuclear performance, have been given prior attention. A preliminary study i n d i c a t e d that, i f the int e g r i t y o f t h e core vessel c o u l d be guaranteed, the nuclear economy o f two-region reactors would probably be superior t o t h a t o f bare and reflected one-region reactors. T h e two-region reactors were, accordingly, studied i n d e t a i l . Although entrance and e x i t c o n d i t i o n s d i c t a t e other than a spherical shape, i t was necessary, for t h e calculations, t o u s e a model comprising the f o l l o w i n g concentric, spherical regions: (1) the core; (2) an INOR-8 reactor vessel, in. thick; (3) a blanket, approximately 2 ft thick; and (4) an INOR-8 reactor vessel,
v3
’/3
in. thick. The diameter of t h e core and the concentration o f thorium i n the core were selected as independent variables. T h e primary dependent v a r i a b l e s were the c r i t i c a l concentration of the
13L. G. Alexander et al., Operatzng Instructzons for t h e Liniuac Program Ocusol-A, a Modification o / t h e E y e w a s h Program, ORNL CF-57-6-4 (June 5, 1957).
14J. H. Alexander and N. D. Given, A Machine Multigroup Calculation. T h e E y e w a s h Program for Univac,
ORNL-1925 (Sept. 15, 1955). L. G. Alexander, Cross Sections 15J. T. Roberts / o r O c u s o l - A frogram, ORNL CF-57-6-5 (June 11, 1957).
4
f u e l (U235, U233, or Pu239) and the d i s t r i b u t i o n o f the neutron absorptions among the various atomic species i n the reactor. From these, the c r i t i c a l mass, c r i t i c a l inventory, regeneration ratio, burnup rate, etc., could be r e a d i l y calculated.
HOMOGENEOUS R E A C T O R S F U E L E D WITH U 2 3 5 While t h e isotope UZJ3would b e a superior fuel i n molten-fluoride-salt reactors, it i s unfortunately not a v a i l a b l e i n quantity. Any r e a l i s t i c appraisal o f t h e immediate c a p a b i l i t i e s o f these reactors must be based on the use of U 2 3 5 . T h e study o f homogeneous reactors was d i v i d e d
i n t o two phases: (1) the mapping o f the nuclear c h a r a c t e r i s t i c s o f the i n i t i a l (i.e., “clean”) states
a s a function o f core diameter and thorium concentration and (2) the a n a l y s i s o f the subsequent performance o f selected i n i t i a l states w i t h various processing schemes and rates. T h e d e t a i l e d res u l t s o f t h e f i r s t phase are given here. B r i e f l y , i t was found that regeneration r a t i o s o f up t o 0.65 c o u l d be obtained w i t h moderate investment i n U235( l e s s than 100 kg). I n i t i a l States
A complete parametric study was made o f moltenf l u o r i d e - s a l t reactors h a v i n g diameters i n the range
o f 4 t o 10 ft and thorium concentrations i n the fuel I n these reranging from 0 t o 1 mole % ThF,.
a c t o r s t h e b a s i c fuel s a l t (fuel s a l t No. 1) was a mixture o f 31 mole % B e F 2 and 69 mole % LiF, which has a density o f about 2.0 g/cm3 at 115OOF. T h e core vessel was composed o f INOR-8. The blanket f l u i d (blanket s a l t No. 1) was a mixture o f 25 mole % ThF, and 75 mole % LiF, which has a density o f about 4.3 g/cm3 at 1150OF. In order t o shorten the c a l c u l a t i o n s i n t h i s series, the reactor v e s s e l was neglected, since the resultant error would be small. These reactors contained no f i s s i o n products or n o n f i s s i o n a b l e isotopes o f uranium other than U 2 3 8 . A summary o f t h e r e s u l t s i s presented i n T a b l e 1, i n which the neutron balance i s presented i n terms o f neutrons absorbed i n a given element per neutron absorbed i n U235[both by f i s s i o n and the(n,y) reaction]. The sum o f the absorptions i s therefore equal t o 7, that is, the number of neutrons produced by f i s s i o n per neutron absorbed i n fuel. Further, the sum o f the absorptions in and thorium i n the fuel and i n thorium i n the blanket
u238
Table
1. I n i t i a l - S t a t e N u c l e a r C h a r a c t e r i s t i c s o f ?wo-Region, Homogeneous, Malten-Fluoride-Salt R e a c t o r s F u e l e d w i t h U235
+
No. 1: 31 mole % B e F 2 69 mole % L i F + U F 4 + ThF4 No. 1: 25 mole % ThF4 + 75 mole % L i F T o t a l power: 600 Mw (heat) External f u e l volume: 339 ft3
Fuel salt
Blanket s a l t
1
C a s e number
2
3
4
5
7
6
10
9
8
11
Core diameter, ft
4
5
5
5
5
5
6
6
6
6
6
ThF4 in f u e l salt, m o l e %
0
0
0.25
0.5
0.75
1
0
0.25
0.5
0.75
1
u~~~in f u e l
0.952
0.318
0.561
0.721
0.845
0.938
0.107
0.229
0.408
0.552
0.662
33.8
11.3
20.1
25.6
30.0
33.3
3.80
8.13
14.5
19.6
23.5
124
81.0
144
183
215
239
47.0
101
179
243
29 1
~~~5 ( f i s s i o n s )
0.7023
0.7185
0.7004
0.6996
0.7015
0.7041
0.7771
0.7343
0.7082
0.7000
0.7004
u235
0.2977
0.2815
0.2996
0.3004
0.2985
0.2959
0.2229
0.2657
0.2918
0.3000
0.2996
0.055 1
0.0871
0.0657
0.0604
0.0581
0.0568
0.1981
0.1082
0.0770
0.0669
0.0631
Core v e s s e l
0.0560
0.0848
0.0577
0.0485
0.0436
0.0402
0.1353
0.0795
0.0542
0.0435
0.0388
L i and F i n b l a n k e t s a l t
0.0128
0.0138
0.0108
0.0098
0.0093
0.0090
0.0164
0.01 16
0.0091
0.0081
0.0074
Leakage
0.0229
0.0 156
0.0147
0.0143
0.0141
0.0140
0.0137
0.0129
0.0122
0.0119
0.01 16
0.0430
0.0426
0.0463
0.0451
0.0431
0.0412
0.0245
0.0375
0.0477
0.0467
0.0452
0.0832
0.1289
0.1614
0.1873
0.1321
0.1841
0.2142
0.2438
U235 atom
salt, m o l e %
density*
C r i t i c a l mass, k g of
U 235
Neutron obsorption r a t i o s * *
hY)
Be, Li, and
u~~~in f u e l
F
in f u e l s a l t
salt
Th i n f u e l s a l t
0.5448
0.5309
0.4516
0.421 1
0.4031
0.3905
0.5312
0.4318
0.3683
0.3378
0.3202
1.73
1.77
1.73
1.73
1.73
1.74
1.92
1.82
1.75
1.73
1.73
M e d i a n f i s s i o n energy, e v
270
15.7
105
158
270
425
0.18
5.6
38
100
120
Thermal f i s s i o n s , %
0.052
6.2
0.87
0.22
0.87
0.040
35
13
3
0.56
0.48
Regeneration r a t i o
0.59
0.57
0.58
0.60
0.61
0.62
0.56
0.61
0.60
0.60
0.61
Th i n b l a n k e t s a l t Neutron y i e l d ,
7
atoms/cm
3
.
**Neutrons absorbed per neutron absorbed in
U 235
.
Table
C r i t i c a l mass, k g of
U
19
20
21
22
10
10
10
10
10
0
0.25
0.5
0.75
1
0
0.25
0.5
0.75
1
0.114
0.047
0.078
0.132
0.226
0.349
0.033
0.052
0.081
0.127
0.205
4.05
1.66
2.77
4.67
8.03
12.4
1.175
1.86
2.88
4.50
7.28
79.6
48.7
81.3
137
236
364
67.3
107
165
258
417
0.25
235
18
8
ThF4 i n fuel salt, mole %
density*
17
8
8
U235 atom
16 8
7
in fuel salt, mole %
15 8
Core diameter, f t
u~~~
14
13
12
Case number
1 (continued)
Neutron a b s o r p t i o n r a t i o s * *
~~~5
(fission)
0.7748
0.8007
0.7930
0.767 1
0.7362
0.7146
0.8229
0.7428
0.7902
0.7693
0.7428
u235
h<Y)
0.2252
0.1993
0.2070
0.2329
0.2638
0.2854
0.1771
0.2572
0.2098
0.2307
0.2572
Be, Li, and F i n f u e l s a l t
0.1880
0.4130
0.2616
0.1682
0.1107
0.0846
0.5713
0.3726
0.2486
0.1735
0.1206
Core vessel
0.0951
0.1491
0.1032
0.0722
0.0500
0.0373
0.1291
0.0915
0.0669
0.0497
0.0363
L i and F
0.0123
0.0143
0.01 12
0.0089
0.0071
0.0057
0.01 14
0.0089
0.0073
0.0060
0.0049
Leakage
0.0068
0.0084
0.0082
0.0080
0.0077
0.0074
0.0061
0.0060
0.0059
0.0057
0.0055
u Z 3 8 in fuel salt
0.0254
0.0143
0.0196
0.0272
0.0368
0.0428
0.0120
0.0153
0.0209
0.0266
0.0343
Th i n fuel s a l t
0.1761
0.2045
0.3048
0.3397
0.3515
0.2409
0.3691
0.4324
0.4506
Th i n blanket s o l t
0.4098
0.4073
0.3503
0.3056
0.2664
0.2356
0.3031
0.2617
0.2332
0.2063
0.1825
191
2.00
1.96
1.89
1.82
1.76
2.03
2.00
1.95
1.90
1.83
Median f i s s i o n energy, ev
0.16
Thermal
0.10
0.17
5.3
27
Thermal
Thermal
0.100
0.156
1.36
Thermal f i s s i o n s , %
33
59
45
29
13
5
66
56
43
30
16
Regeneration r a t i o
0.6 1
0.42
0.57
0.64
0.64
0.63
0.32
0.52
0.62
0.67
in b l a n k e t s a l t
Neutron y i e l d ,
..
q
atoms/cm
3
.
**Neutrons absorbed per neutron absorbed in
U 235
0.67 -
.
,
,
s a l t g i v e d i r e c t l y t h e regeneration ratio. The l o s s e s t o other elements are p e n a l t i e s imposed on t h e regeneration r a t i o by these poisons. A graph o f c r i t i c a l mass p l o t t e d as a function o f core diameter, w i t h thorium concentration as a parameter, i s presented in Fig. 1. The masses in a 7 - f t - d i a core vary from about 40 k g of lJ235 h a v i n g no thorium in the f u e l t o about 450 k g i n the 10-ft-dia core having 1 mole % ThF, i n the UNCLASSI FlED ORNL-LR-DWG 39521
500
CORE A N D B L A N K E T S A L T S NO. 1
I
0.8
I
I
6
7
mole% ThF4 IN F U E L S A L T
0 c
a
(L
z 0.6
2 c a (L
6 0.4 W
W
n
CORE A N D B L A N K E T I N F U E L SALT
400
-
5
4
,--
Ln N 10
8
10
9
CORE D I A M E T E R ( f t )
3
0 ~
1
Fig.
300
m a
> a
_J
2.
Initial
Homogeneous,
I
with 200
[L
0
100
0
4
5
6
7
8
9
10
CORE D I A M E T E R ( f t )
Fig.
Regeneration
i n Two-Region, Reactors
Fueled
concentration i n the fuel salt, both the c r i t i c a l concentration and the regeneration r a t i o were somewhat lower for t h e No. 2 salts.
0 L
0
U235.
Fuel
Mol ten- Fluoride-Sal t
1.
I n i t i a l C r i t i c a l Masses
of U235 in Two-
Region, Homogeneous, Molten-Fluoride-Salt Reactors.
fuel. T h e corresponding regeneration ratios, p l o t t e d i n Fig. 2, range from 0.5 for the minimum mass reactor t o 0.63 for the largest mass reactor. It does not seem l i k e l y t h a t further increases i n diameter or thorium concentration would increase t h e regeneration above 0.65. T h e e f f e c t s o f changes i n the compositions o f t h e fuel and blanket s a l t s were studied i n a series o f c a l c u l a t i o n s for s a l t s h a v i n g more favorable m e l t i n g p o i n t s and v i s c o s i t i e s . The BeF, content was r a i s e d t o 37 mole % i n the f u e l s a l t (fuel s a l t No. 2), and the blanket composition (blanket s a l t No. 2) was f i x e d a t 13 mole % ThF,, 16 mole 76
BeF,, and 71 mole % LiF. B l a n k e t s a l t No. 2 i s a somewhat better r e f l e c t o r than No. 1, and fuel s a l t No. 2 i s a somewhat better moderator than No. 1. A s a result, ot a given core diameter and thorium
Reservations concerning the f e a s i b i l i t y s t r u c t i n g and guaranteeing the i n t e g r i t y v e s s e l s in large s i z e s (10 f t and over), w i t h preliminary consideration o f inventory
o f cono f core together charges
for large systems, l e d t o the conclusion that a f e a s i b l e reactor would probably have a core diameter l y i n g i n t h e range between 6 and 8 ft. A c cordingly, a parametric study o f the No. 2 fuel and blanket s a l t s i n reactors w i t h core diameters i n
t h e 6- to 8-ft range was made. I n t h i s study the presence o f an outer reactor vessel c o n s i s t i n g of in. o f INOR-8 was taken i n t o account. The r e s u l t s are presented i n T a b l e 2. I n general, the nuclear performance i s somewhat better w i t h the No. 2 s a l t s than with the No. 1 salts.
5
Neutron Balances and Qther Reactor Variables The
d i s t r i b u t i o n s of the neutron captures are
g i v e n in T a b l e s
1
and
2,
where the r e l a t i v e hard-
ness o f t h e neutron spectrum i s i n d i c a t e d by the median f i s s i o n energies and the percentages o f thermal fissions. It may be seen that l o s s e s t o lithium, beryllium, and fluorine i n the fuel s a l t and t o the core vessel are substantial, e s p e c i a l l y
i n t h e more thermal reactors (e.g., case No. 18). However, i n t h e thermal reactors, l o s s e s by radiat i v e capture i n ~~~5 are r e l a t i v e l y low. Increasing
7
Table
2. I n i t i a l - S t a t e Nuclear C h a r a c t e r i s t i c s o f Two-Region, Homogeneous, M o l t e n - F l u a r i d e - S a l t Reactors F u e l e d w i t h U 2 3 5 2: 37 mole % B e F 2 + 63 mole % L i F + UF4 + ThF4 2: 13 mole % ThF, + 16 mole % BeF, + 71 mole % L i F T o t a l power: 600 Mw (heat) External fuel volume: 339 ft3 F u e l s a l t No.
B l a n k e t s a l t No.
23
Case number
24
25
26
28
27
29
30
31
32
33
34
Core diameter, f t
6
6
6
6
7
7
7
7
8
8
8
8
ThF4
0.25
0.5
0.75
1
0.25
0.5
0.75
1
0.25
0.5
0.75
1
~~~5 i n fuel salt, mole %
0.169
0.310
0.423
0.580
0.084
0.155
0.254
0.366
0.064
0.099
0.163
0.254
u235atom d e n s i t y *
5.87
10.91
15.95
20.49
3.13
5.38
8.70
13.79
2.24
3.51
5.62
9.09
72.7
135
198
254
61.5
106
171
27 1
65.7
103
165
267
0.7516
0.7174
0.7044
0.6958
0.7888
0.7572
0.7282
0.7094
0.8014
0.7814
0.7536
0.7288
0.2484
0.2826
0.2956
0.3042
0.21 12
0.2428
0.2718
0.2906
0.1986
0.2186
0.2464
0.2712
0.1307
0.0900
0.0763
0.0692
0.2147
0.1397
0.1010
0.0824
0.2769
0.1945
0.1354
0.1016
0.0726
0.0575
0.0473
0.1328
0.0905
0.0644
0.0497
0.1308
0.0967
0.0696
0.0518
0.0215
0.0167
0.0131
0.0108
0.0198
0.0162
0.0130
0.0105
in fuel salt, mole %
C r i t i c a l muss, k g o f U
235
Neutron absorption r a t i o s * *
u~~~ ( f i s s i o n s ) u235 Be,
(n,y)
Li, and F in f u e l s a l t
Core v e s s e l
0.1098
L i and F i n b l a n k e t s a l t
0.0214
0.0159
0.0132
0.01 17
Outer v e s s e l
0.0024
0.0021
0.0021
0.0019
0.0019
0.0018
0.0016
0.0015
0.0017
0.0016
0.0014
0.0013
Leakage
0.0070
0.0065
0.0064
0.0061
0.0052
0.0050
0.0048
0.0045
0.0045
0.0043
0.0042
0.0040
0.0325
0.0426
0.0452
0.0477
0.0214
0.0307
0.0392
0.0447
0.0177
0.0233
0.0315
0.0392
Th i n fuel s a l t
0.1360
0.1902
0.2212
0.2387
0.1739
0.2565
0.2880
0.3022
0.1978
0.3043
0.3501
0.3637
Th i n b l a n k e t s a l t
0.4165
0.3521
0.3178
0.2962
0.3770
0.2566
0.3240
0.2892
0.2561
0.2280
1.86
1.77
1.74
1.72
1.95
0.3294 0.2866 - _ _ _ 1.87 1.80
1.75
1.97
1.93
1.86
1.80
Median f i s s i o n energy, ev
0.480
10.47
58.10
76.1
0.1223
0.415
7.61
25.65
51% th
0.136
0.518
7.75
Thermal f i s s i o n s , %
21
7
2.8
0.84
43
24
11
4.3
51
38
23
11
Regeneration r a t i o
0.59
0.58
0.58
0.58
0.57
0.62
0.61
0.60
0.54
0.62
0.64
0.63
u~~~i n
fuel s o i t
Neutron yield,
q
atoms/cm
3
.
**Neutrons absorbed per neutron absorbed i n U
235.
.
t h e hardness decreases l o s s e s t o s a l t and core vessel sharply (case No. 5) b u t increases the l o s s T h e numbers given for t o the (n,y) reaction. capture i n t h e l i t h i u m and f l u o r i n e in the blanket show that these elements are w e l l shielded by the
20 k g in a 5 - f t - d i a core, w i t h no thorium present, t o 130 k g i n a 10-ft-dia core having 1 mole % thorium in t h e fuel. T h e corresponding regenerat i o n r a t i o s are 0.60 and 0.90. For a given thorium
thorium i n t h e blanket, and the leakage values show that leakage from the reactor i s less than 0.01 neutron per neutron absorbed i n U235 in reT h e blanket conactors over 6 ft i n diameter, t r i b u t e s s u b s t a n t i a l l y t o the regeneration o f fuel, accounting for not l e s s than one-third o f the total, even i n the 10-ft-dia core c o n t a i n i n g 1 mole %
ThF,.
140
I
UNCLASSIFIED ORNL-LR-DWG 39523
! CORE AND B L A N K E T SALTS NO 4
120
I
1
0
2
mole%ThF4
IN FUEL SALT,
+
I
400
C-I
N CI
I)
0 LL
80
cn I
H O M O G E N E O U S R E A C T O R S F U E L E D WITH U 2 3 3
In
Uranium-233 i s a superior fuel for u s e in moltenf l u o r i d e - s a l t reactors i n almost every respect. T h e
a
f i s s i o n cross section in the intermediate range of neutron energies i s greater than the f i s s i o n cross section o f U235. Thus i n i t i a l c r i t i c a l inventories
a
z
60
J
u k n 0
40
are less, and l e s s a d d i t i o n a l fuel i s required t o override poisons. Also, the p a r a s i t i c cross sec-
20
t i o n i s s u b s t a n t i a l l y less, and fewer neutrons are l o s t t o r a d i a t i v e capture. Further, the r a d i a t i v e captures r e s u l t in the immediate formation o f a f e r t i l e isotope, U234. T h e r a t e o f accumulation
U236 i s orders U235 as a fuel,
o f magnitude smaller than w i t h and t h e b u i l d u p o f N p 2 3 7 and Pu239 i s negligible. The mean neutron energy i s somewhat nearer thermal i n such reactors than i t i s i n the corresponding U235cases. Consequently, l o s s e s t o
3 are for reactors using fuel and blanket s a l t s No. 1 w i t h ThF, concentrations ranging up t o 1 mole %. T h e c r i t i c a l masses are graphed in Fig. 3 and the regeneration r a t i o s i n F i g . 4. T h e masses range from a minimum o f about
8
6
12
10
CORE DIAMETER ( f t )
of
core v e s s e l and t o core s a l t tend t o be higher. B o t h l o s s e s are reduced s u b s t a n t i a l l y a t higher thorium concentrations because o f the hardening o f t h e neutron spectrum. R e s u l t s from a parametric study o f the nuclear c h a r a c t e r i s t i c s o f two-region, homogeneous, moltenf l u o r i d e - s a l t reactors fueled w i t h U 2 3 3 are given i n T a b l e s 3 and 4. The core diameters considered range from 3 t o 12 ft, and the thorium concentrat i o n s range from 0.25 t o 7 mole %. T h e regenerat i o n r a t i o s are very good compared w i t h those obt a i n e d w i t h U235. With 7 m o l e % ThF, i n an 8 - f t - d i a core, t h e U 2 3 3 c r i t i c a l mass was 1500 kg, and t h e regeneration r a t i o was 1.09.
4
Fig.
3.
I
I
I
I
I
I
CORE AND BLANKET SALTS NO. 1 1.0
Q
2 n
.
C r i t i c a l Masses of Two-Region, Homogeneous,
Molten-Fluoride-Salt Reactors Fueled with
U233
mole 9eThFq
IN FUEL SALT
0.8
2
0 W
06
z
W
W 0
04
1
0.2
0
3
5
4
T h e data i n T a b l e
Fig.
4.
Initial
Homogeneous, with
6
1
I
7
8
9
10
CORE DIAMETER ( f t )
F u e l Regeneration in Two-Region,
Molten-Fluoride=Solt
Reactors
Fueled
U233.
9
Table
3. I n i t i a l - S t a t e Nuclear C h a r a c t e r i s t i c s o f Two-Region, Homogeneous, M o l t e n - F l u o r i d e - S a l t Reactors F u e l e d w i t h U233 1: 31 mole % 8 e F 2 + 69 mole % L i F 25 mole % ThF4 75 m o l e % L i F T o t a l power: 600 Mw (heat) E x t e r n a l fuel volume: 339 f t 3 F u e l s a l t No.
+
B l a n k e t salt:
36
35
Case number
37
38
39
40
+ UF4 + ThF4
41
42
43
44
45
46
Care diameter, f t
3
4
4
5
5
5
5
5
6
6
6
6
ThF4 in f u e l salt, mole %
0
0
0.25
0
0.25
0.5
0.75
1
0
0.25
0.5
0.75
u~~~i n fuel
0.592
0.158
0.233
0.076
0.106
0.141
0.179
0.214
0.048
0.066
0.087
0.113
21.1
5.6
8.26
2.7
3.73
5.0
6.35
7.605
1.7
2.3
3.1
4.0
64.9
22.3
30.3
19.3
26.9
35.8
45.5
54.5
20.4
29.2
38.4
49.5
u~~~( f i s s i o n s )
0.8754
0.8706
0.8665
0.8767
0.8725
0.8684
0.8674
0.8672
0.8814
0.8779
0.8744
0.8665
u233
0.1246
0.1294
0.1335
0.1233
0.1275
0.1316
0.1326
0.1328
0.1186
0.1221
0.1256
0.1335
Be, Li, and F in f u e l s a l t
0.0639
0.1051
0.0860
0.1994
0.1472
0.1174
0.1010
0.0905
0.3180
0.2297
0.1774
0.1412
Core v e s s e l
0.0902
0.1401
0.1093
0.1808
0.1380
0.1112
0.0944
0.0821
0.1983
0.1508
0.1209
0.0989
L i and F in b l a n k e t s a l t
0.0233
0.0234
0.0203
0.0232
0.0 196
0.0 172
0.0157
0.0 146
0.0215
0.0179
0.0157
0.0139
Leakage
0.0477
0.0310
0.0306
0.0197
0.0193
0.0190
0.0189
0.0188
0.0160
0.0157
0.0157
0.0154
Th in fuel s a l t
0.0000
0.0000
0.1095
0.0000
0.1593
0.256 1
0.3219
0.3702
0.0000
0.1973
0.31 11
0.3989
Th in blanket s a l t
0.9722
0.8857
0.8193
0.7777
0.7066
0.5487
0.6255
0.6004
0.6586
0.5922
0.5539
0.5169
2.1973
2.1853
2.1750
2.2007
2.1900
2.1797
2.1773
2.1766
2.2124
2.2035
2.1948
2.185
Median f i s s i o n energy, e v
174
14.2
19.1
1.752
2.87
9.625
16.5
29.35
0.326
1.18
2.175
10.16
Thermal f i s s i o n s , %
0.0527
7.952
2.970
24.80
16.499
10.09
5.99
3.192
37.832
29.37
27.12
14.87
Regeneration r a t i o
0.9722
0.8856
0.9288
0.7777
0.8659
0.9148
0.9474
0.9706
0.5486
0.7895
0.8651
0.9158
salt, mole %
U233 atom d e n s i t y * U
C r i t i c a l mass, k g of
233
Neutron a b s o r p t i o n r a t i o s * *
(n,y)
Neutron yield,
q
atoms/cm
3
.
**Neutrons absorbed per neutron absorbed in
U233
.
I
b
Table
47
Case number
48
49
,
*
I
3 (continued)
50
51
52
53
54
55
56
57
Core diameter, f t
6
8
8
8
8
8
10
10
10
10
10
ThF4 in f u e l solt, mole %
1
0
0.25
0.5
0.75
1
0
0.25
0.5
0.75
1
u~~~i n
0.133
0.028
0.039
0.052
0.066
0.078
0.022
0.031
0.041
0.051
0.063
U233 otom d e n s i t y *
4.72
1.01
1.41
1.85
2.33
2.72
0.780
1.09
1.45
1.8
2.25
C r i t i c a l moss, k g o f U 2 3 3
58.4
29.6
41.1
54.3
68.4
86.6
44.7
63.0
83.1
103.2
131.3
0.8693
0.8876
0.8850
0.8808
0.8779
0.8755
0.8921
0.8881
0.8842
0.8814
0.8781
0.1307
0.1124
0.1150
0.1192
0.1221
0.1245
0.1079
0.1119
0.1158
0.1186
0.1219
0.1216
0.5433
0.3847
0.2896
0.2285
0.1829
0.7166
0.5037
0.3758
0.2952
0.2360
Core v e s s e l
0.0855
0.1866
0.1406
0.1112
0.0915
0.0778
0.1560
0.1168
0.0919
0.0754
0.0629
L i and F i n b l o n k e t s o l t
0.0127
0.0176
0.0141
0.0120
0.0106
0.0095
0.0133
0.0108
0.009 1
0.0080
0.007 1
Leakage
0.0152
0.0095
0.0095
0.0093
0.0091
0.0090
0.0068
0.0068
0.0066
0.0065
0.0065
Th
i n fuel s a l t
0.4580
0.0000
0.2513
0.4044
0.5055
0.5768
0.0000
0.2852
0.4585
0.5708
0.6507
Th in b l o n k e t s a l t
0.4889
0.4707
0.421 1
0.3842
0.3582
0.3344
0.3466
0.3058
0.2774
0.2564
0.2408
Neutron yield, T]
2.1820
2.2277
2.2212
2.2108
2.2035
2.1975
2.2392
2.2290
2.2194
2.2133
2.2040
Median f i s s i o n energy, e v
8.51
52% t h
0.197
0.4915
1.185
1.12
58% th
50% th
0.1735
0.455
3.25
Thermal f i s s i o n s , %
12.42
51.93
43.398
35.79
29.078
24.36
58.34
50.39
42.8
36.45
29.96
Regenerotion r a t i o
0.9470
0.4707
0.6725
0.7886
0.8638
0.9112
0.3467
0.5910
0.7359
0.8271
0.8915
f u e l salt, mole %
Neutron obsorption r a t i o s * *
u~~~ ( f i s s i o n s ) u233 Be,
b,Y)
Li, and F
in f u e l s o l t
otoms/cm
3
.
**Neutrons obsorbed per neutron obsorbed i n U
233
.
Table
4.
I n i t i a l - S t a t e N u c l e a r C h a r a c t e r i s t i c s of Two-Region, Homogeneous, M o l t e n - F l u o r i d e - S a l t R e a c t o r s F u e l e d w i t h
+
2: 37 male % B e F 2 + 63 mole % L i F UF4 + ThF4 2: 13 mole % ThF4 16 mole % B e F 2 + 71 male % L i F T o t a l power: 600 Mw (heat) External f u e l volume: 339 f t 3 F u e l s a l t No.
+
Blanket s a l t No.
58
Case number
59
61
60
63
62
64
65
66
67
U233
68
69
70
Core diameter, ft
6
6
6
6
7
7
7
7
8
8
8
8
4
Th4 in f u e l salt, mole %
0.25
0.5
0.75
1
0.25
0.5
0.75
1
0.25
0.5
0.75
1
2
u~~~in
0.062
0.081
0.10
0.121
0.047
0.059
0.074
0.091
0.039
0.049
0.062
0.075
0.619
U233atom d e n s i t y *
1.98
2.6
3.2
3.88
1.5
1.92
2.38
2.9
1.21
1.58
1.97
2.41
19.8
233 C r i t i c a l mass, k g o f U
24.51
32.19
39.62
48.03
29.49
37.75
46.79
57.01
35.51
46.37
57.82
70.73
72.0
u~~~( f i s s i o n s )
0.8805
0.8762
0.8741
0.8722
0.8843
0.8809
0.8784
0.8749
0.8880
0.8828
0.8809
0.8827
0.871
,,233
0.1195
0.1238
0.1259
0.1278
0.1157
0.1191
0.1216
0.1251
0.1120
0.1172
0.1191
0.1173
0.129
0.2427
0.1915
0.1604
0.1383
0.3209
0.2525
0.2062
0.1735
0.2407
0.305 1
0.2458
0.2073
0.070
Care v e s s e l
0.1891
0.1526
0.1288
0.1109
0.1858
0.1505
0.1258
0.1070
0.1756
0.1405
0.1168
0.1003
0.073
L i and F in blanket salt
0.0313
0.0272
0.0243
0.0221
0.0276
0.0238
0.021 1
0.0190
0.0247
0.0212
0.0187
0.0169
0.025
Leakage
0.0133
0.0111
0.0109
0.0108
0.0094
0.0081
0.0080
0.0078
0.0070
0.0069
0.0068
0.0068
0.031
Th in fuel s a l t
0.1901
0.3088
0.3902
0.4504
0.2182
0.3531
0.4455
0.5125
0.3952
0.3891
0.4903
0.5678
0.343
Th i n blanket s a l t
0.5454
0.5079
0.4794
0.4566
0.4589
0.4228
0.3983
0.3763
0.3859
0.3533
0.3325
0.3164
0.653
2.2100
2.1992
2.1940
2.1891
2.2197
2.21 10
2.2049
2.1960
2.2289
2.2160
2.21 10
2.2155
2.195
Median f i s s i o n energy, e v
0.721
1.575
2.475
3.685
0.1875
0.465
0.992
2.025
0.1223
0.230
0.676
1.345
147
Thermal f i s s i o n s , %
33.878
26.269
20.518
15.584
4 1.997
35.191
28.685
23.051
47.965
40.663
33.87
28.301
0.23
Regeneration r a t i o
0.7355
0.8167
0.8695
0.9071
0.6770
0.7760
0.8438
0.8887
0.6264
0.7424
0.8228
0.8842
0.996
fuel salt, mole %
Neutron absorption r a t i o s * *
Be, Li, and
Neutron yield,
F
in f u e l s o l t
7
atoms/cm
3
.
**Neutrons absorbed per neutron absorbed i n
U
233
.
1
.
J
Table
71
C a s e number
72
73
74
4 (continued)
75
76
77
ao
79
78
ai
a2
84
83
Core diameter, f t
4
4
6
6
6
8
a
a
10
10
10
12
12
12
ThF4 in fuel salt, mole %
4
7
2
4
7
2
4
7
2
4
7
2
4
7
0.856
1.247
0.236
0.450
0.762
0.152
0.316
0.603
0.121
0.262
0.528
0.101
0.222
0.477
27.4
39.9
7.55
14.4
24.4
4.88
10.1
19.3
3.86
8.30
16.9
3.24
7.39
15.25
100.5
146.5
94.2
177.8
30 1
143
299
566
22 1
48 1
970
320
732
1510
0.874
0.881
0.864
0.868
0.876
0.867
0.865
0.873
0.870
0.864
0.871
0.873
0.864
0.870
0.126
0.119
0.136
0.132
0.124
0.133
0.135
0.127
0.130
0.136
0.129
0.127
0.136
0.130
0.066
0.069
0.093
0.075
0.076
0.120
0.082
0.078
0.142
0.088
0.081
0.164
0.093
0.083
Core v e s s e l
0.059
0.048
0.068
0.049
0.035
0.057
0.037
0.025
0.046
0.030
0.018
0.033
0.022
0.012
L i and F in b l a n k e t s a l t
0.021
0.019
0.0 16
0.014
0.01 1
0.019
0.012
0.009
0.009
0.006
0.006
0.012
0.004
0.004
Leakage
0.031
0.028
0.017
0.017
0.015
0.010
0.010
0.009
0.007
0.008
0.007
0.004
0.006
0.004
Th i n fuel salt
0.426
0.517
0.581
0.650
0.740
0.716
0.785
0.865
0.800
0.865
0.938
0.872
0.922
0.998
Th i n blanket s a l t
0.600
0.330 0.382 2.187 2.2072
0.264
0.254
0.21 13
0.189
0.130
0.092
2.1860
2.180
2.200
2.1933
0.170 0.146 2.196 2.177
0.115
2.203
0.403 0.538 2.1770 2.219
2.1995
2.177
2.1931
Median f i s s i o n energy, e v
503
1085
24.2
64.0
443
8.42
51.3
243
3.64
45.9
193
2.45
41.4
178
Thermal fissions, %
0.11
0.084
4.3
0.35
0.076
11.0
1.5
0.091
17
1.a
0.12
21
2.1
0.15
Regeneration r a t i o
1.026
1.055
0.984
1.032
1.070
0.980
1.039
1.078
0.989
1.045
1.084
0.987
1.052
1.090
u~~~ i n fuel
U233
salt, mole %
atom density*
Critical mass, kg of
u
233
Neutron absorption r a t i o s * *
u~~~(fissions) ,233
hY)
Be, L i , and
Neutron yield,
F i n fuel s a l t
17
atams/cm 3 **Neutrons absorbed per neutron absorbed i n
U2 3 3
.
concentration, t h e regeneration r a t i o tends t o i n crease w i t h decreasing core size, and r a t i o s up t o
0.97
were observed i n t h i s series o f calculations, as shown i n F i g . 4. T h e data i n T a b l e 4 are for reactors using fuel and blanket s a l t s No. 2. I n t h i s series o f calculations, t h e diameter ranged up t o 12 ft and the thorium concentration i n the core up t o 7 mole %. It was necessary t o a l t e r progressively the comp o s i t i o n o f t h e base s a l t as the thorium concentrat i o n was increased i n order t o keep the l i q u i d u s temperature below 1000째F. There was a s l i g h t increase i n concentration o f L i F at the expense o f For cores h a v i n g thorium concentrations i n BeF,. the range from 0.25 t o 1 mole %, the r e s u l t s are about t h e same as those obtained w i t h cores u s i n g
No. 1 salts. T h e behavior w i t h No. 2 s a l t s a t higher concentrations o f ThF, i s shown i n F i g s . 5 and 6. It i s seen t h a t an i n i t i a l regeneration r a t i o o f about 1.0 can be achieved w i t h about 2.5 mole % thorium i n t h e fuel, regardless o f the diameter o f
4.4
Q
I
I
I
UNCLASSIFIED ORNL-LR-DWG 39526
I
1.0
5z 0
2
[L
w
z
sn W
CORE AND B L A N K E T SALTS NO. 2
1
I
8
IO
0.9
0.8
2
0
4
6
CORE DIAMETER
Fig.
6.
12
I n i t i a l Regeneration R a t i o in Two-Region,
Homogeneous, with
(f0
Mol ten-Fluor ide-Sal t
Reactors
Fueled
U233.
the core i n t h e range from 4 t o 12 ft. The corresponding c r i t i c a l masses range from about 80 t o 400 k g o f U233. NUCLEAR PERFORMANCEOF A R E F E R E N C E
DESIGN R E A C T O R
A conceptual d e s i g n study o f a 240-Mw ( e l e c t r i c a l ) c e n t r a l - s t a t i o n molten-salt-fueled reactor (MSR) was described by t h e Molten-Salt Reactor Group at
ORNL. l 6
T h e system employs a two-region homogeneous reactor having a core approximately 8 ft i n diameter and a blanket 2 ft thick.
The core, w i t h i t s volume of 113 ft3, i s capable o f generating 600 Mw o f heat a t a power density i n T h e general arrangement t h e core o f 187 w/cm3. o f t h e core and blanket i s shown i n F i g . 7. The I
I
I
b a s i c core s a l t i s a mixture o f lithium, beryllium, and thorium fluorides, together w i t h s u f f i c i e n t fluoride
5. C r i t i c a l Masses of Two-Region, Homogeneous, Molten-Fluoride-So I t Reactors Fueled with U233. Fig.
14
of
u235 or
U 2 3 3 t o make the system
16J. A. L a n e , H. G. MacPherson, and F. Maslan (eds.), F l u i d Fuel Reactors, p 569, Addison-Wesley, Reading, Mass., 1958.
UNCLASSIFIED ORNL-LR-DWG 286368
BLANKET
SECTION A - A FUEL PUMP MOTOR
MOTOR BLANKET EXPANSION
FUEL LINE TO HEAT EXCHANGER
-
2
0
2
F U E L EXPANSION
4
MOLTEN SALT
FEET
BREEDING BLANKET
FUEL RETURN
Fig. 7.
General Arrangement of Core and Blanket.
15
c r i t i c a l . T h e b l a n k e t c o n t a i n s ThF,, either as the or mixtures of i t w i t h eutectic o f L i F and ThF,, t h e b a s i c core salt. The l i q u i d u s temperature o f the fuel s a l t i s about 85OOF and that of the blanket i s 1O8O0F or lower. Both the core f u e l and t h e blanket s a l t are circulated t o external heat exchangers, s i x i n p a r a l l e l for t h e core and t w o i n p a r a l l e l for the blanket. T h e heat i s transferred by intermediate f l u i d s from these heat exchangers t o the boilers, superheaters, and reheaters. T h e heat transfer system i s designed so that, w i t h a fuel temperature o f 12OO0F, a steam temperature o f 1000째F a t 1800 psi can be achieved. The volume o f fuel s a l t external t o the core i n the transfer lines, pumps, and heat exchangers was estimated t o be 339 ft3. It i s t h i s external volume that largely determines t h e fuel inventory o f the system.. A parametric study of the regeneration r a t i o as a function of c r i t i c a l inventory i n t h i s
system was performed. With U 2 3 5i n reactors emp l o y i n g core and blanket s a l t s No. 1, the r e s u l t s are as shown i n F i g . 8, where regeneration r a t i o i s p l o t t e d vs c r i t i c a l inventory, w i t h thorium concentration i n the fuel as a parameter. The numbers associated w i t h t h e p l o t t e d p o i n t s are the diameters o f t h e cores. T h e curves are observed t o peak rather sharply, and these peaks define a locus o f maximum regeneration r a t i o for a given inventory. It may be
seen that, w i t h no thorium in the core, a regenerat i o n r a t i o o f 0.5 can be obtained w i t h an inventory o f 100 k g o f U 2 3 5i n a 7 - f t - d i a core. The a d d i t i o n o f 0.25 mole % thorium t o the core s a l t y i e l d s a regeneration o f about 0.6 for an inventory o f 200 k g i n a 7 - f t - d i a core. Optimum core s i z e increases hereafter. Also the r a t e o f increase of regeneration ratio f a l l s o f f substantially. With 0.75 mole % thorium, a regeneration of 0.67 i s obtained w i t h 400 k g o f fuel i n a 10-ft-dia core. A s mentioned above, it was f e l t that i t would be d i f f i c u l t t o
f a b r i c a t e r e l i a b l e core v e s s e l s having diameters greater than 10 ft, and, accordingly, larger cores were not investigated. However, an examination o f t h e curves i n d i c a t e s t h a t further increases in thorium loading and core diameter would probably not increase t h e regeneration r a t i o above 0.7. Reactors employing core and blanket s a l t s No.
o f 1 mole 7' 6 thorium, whereas w i t h the No. 1 salts, The regeneration about 850 k g was required. ratios, however, are about t h e same, ranging from
0.62 t o 0.64 for t h e 8 - f t - d i a cores. 065
050
w
I !
04
03
010
0
Fig.
1
I
8.
Initial
Homogeneous, with U235.
16
I
Fuel
with
I
400 600 800 (000 CRITICAL INVENTORY (kg OF UZ3?
(200
i400
Regeneration in Two-Region,
Molten-Fluoride-Salt
Reactors
I
Fueled
I
1
0
200
9.
1
1
600 800 CRITICAL INVENTGRY (kg OF u 400
Initial
Homogeneous,
CORE AN0 BLANKET SALTS NO 4 NUMBERS ON CURVE POINTS ARE CORE DIAMETERS IN F E E T
I
200
UNCLASSIFIED ORNL-LR-DWG 39528
I
Fig.
TOTAL POWER 600 Mw (th) EXTERNAL VOLUME OF SALT 339 i t 3
2
(see T a b l e 4) require somewhat lower inventories than t h e corresponding cores using s a l t s No. 1, as shown i n F i g . 9. T h e 8 - f t - d i a core, for instance, requires o n l y about 600 k g o f U235w i t h a loading
Fuel
1 ~
io00
~
~
)
(ZOO
Regeneration in Two-Region,
Molten-Fluoride-Salt
Reactors
Fueled
U235.
u233
With as t h e fuel, there i s a marked improvement i n the performance, and the inventories are much lower. T h e performance of cores using c u e and b l a n k e t s a l t s No. 1, w i t h thorium concentrations ranging up t o 1 mole %, i s shown i n Fig.
10. T h e regeneration r a t i o s range up to 0.95
.
a t inventories l e s s than 300 k g o f U 2 3 3 . T h e b e r y l l i u m - r i c h core and blanket s a l t s (No. 2) gave s u b s t a n t i a l l y t h e same results, a s shown i n Fig. 11.
I
I
UNCLASSIFIED ORNL-LR-OWG 39529 I I
09
08
0
5
07
0
G
06
0 W [L
A NO ThF4 IN FUEL SALT CORE AND BLANKET SALTS NO 1
03
0
100 200 300 CRITICAL INVENTORY (kg OF U233)
Fig. 10.
400
I n i t i a l Regeneration of F u e l in Two-Region,
Homogeneous,
Molten-Fluoride-Salt
Reactors
Fueled
Regeneration r a t i o s o f the order o f 0.6 can be obtained a t inventories o f about 100 kg o f U233. Increasing t h e thorium concentration up t o 7 mole % g i v e s a monotonically increasing regenerat i o n ratio, up t o about 1.09, but the fuel inventories become very high. The performance o f cores h a v i n g
diameters ranging from 4 t o 12 ft and thorium concentrations o f 1, 2, 4, and 7 mole % are shown i n F i g . 12. T h e dashed l i n e i s t h e estimated envelope o f t h e curves shown and represents the locus o f maximum regeneration for a given inventory. It i s seen t h a t regeneration r a t i o s above 1.0 can be obtained from fuel investments o f 400 k g or greater. Also, it appears t h a t t h e 8-ft-dia cores g i v e about t h e h i g h e s t regeneration a t a l l thorium concentrations. I n Fig. 13 the performances of 8 - f t - d i a cores u s i n g fuel and blanket s a l t s No. 2 and U235 and U 2 3 3 fuel, respectively, are compared. With U235, a maximum regeneration of about 0.65 i s obtained a t an inventory o f about 400 kg. The g i v e s a regeneration r a t i o o f same amount o f 1.0, and 1000 k g o f U 2 3 3 g i v e s a regeneration o f
u~~~
1.07.
with U233.
4.4
I
UNCLASSIFIED ORNL-LR-DWG 3 9 5 3 0
1
I
I
UNCLASSIFIED ORNL-LR-DWG 39
1.15
N U M B E R S ON DATA P O I N T S A R E CORE D I A M E T E R S I N F E E T
I
4.0
G
~
I
CORE A N D B L A N K E T S A L T S NO.
FUEL SALT: 37 mole % BeF2 + 63 mole % L I F + UF, + ThF,
4.10
2
1.05
0.3
2 n
0
0
0
Gn
z
t
n W z
z
t
0.0
2 W
0
n W
0
n W
1.00
[ W r
W
0.7
0.95
1
I
0.6
0.90
0.5
50
0
Fig. 11.
U233.
400
150
CRITICAL INVENTORY ( k g O F U 2 3 3 )
200
Molten-Fluoride-Salt
Reactors
0.85
0
400
800
f200
1600
2000
2400
TOTAL FUEL INVENTORY ( k g OF U233)
I n i t i a l Regeneration of F u e l in Two-Region,
Homogeneous, with
NUMBERS ON DATA POINTS ARE CORE DIAMETERS IN FEET
Fueled
Fig.
12.
I n i t i a l Regeneration
Homogeneous, with
U233.
Ratio in Two-Region,
Mol ten- Fluoride-Sa I t
Reactors
Fueled
17
COMPARATIVE PERFORMANCE In comparison w i t h converter reactors, the homoi s somewhat inferior geneous MSR fueled w i t h
u235
4.4
I
I
1
CORE AND BLANKET SALTS NO. 2 CORE DIAMETER: E f t EXTERNAL FUELVOLUME: 339 f t 3
4.2
1
4 .O
UNCLASSIFIED ORNL-LR-DWG 3953
L L I
I
0
tLz
z
0.8
0
5 W
06
8 0.4
i n respect t o regeneration ratio, but it i s capable o f matching or exceeding t h e other systems i n s p e c i f i c power, a s i n d i c a t e d i n T a b l e 5. However,
i t should be noted that t h e U 2 3 3 produced i n the MSR i s recycled, whereas the plutonium produced
by the other systems l i s t e d i s not recycled. T h e economics o f those systems depends strongly on t h e market v a l u e o f P u 2 3 9 , and t h i s isotope i s i n f e r i o r as a fuel r e l a t i v e to U233. Thus a regeneration o f 0.5 for U 2 3 3 ( a t $15 a gram) i s equivalent t o a regeneration 0.62 for plutonium ( a t $12 a gram). It i s concluded t h a t the molten-saltfueled system i s c o m p e t i t i v e w i t h s o l i d - f u e l converter reactors burning U 2 3 5 i n respect t o s p e c i f i c power and fuel regeneration. The choice between t h e t w o systems therefore l i e s i n other factors, such a s r e l i a b i l i t y , maintainability, and c o s t o f fuel reprocessing.
02
0
ACKNOWLEDGMENTS 200
0
Fig.
13.
18
1200
(400
I n i t i a l Regeneration of F u e l in Two-Region,
Homogeneous, Reactors.
400 600 800 1000 CRITICAL INVENTORY (kq OF FUEL)
Molten-Fluoride-Salt,
Reference
Design
T h e authors g r a t e f u l l y acknowledge t h e a d v i c e and a s s i s t a n c e o f R. Van Norton, I n s t i t u t e o f mathe-
matics, New York University, and o f D. Grimes, L. Dresner, W. E. Kinney, and R. H. F r a n k l i n o f Oak Ridge National Laboratory.
.
I
Table 5. Comparison of Converter Reactors
Reactor*
Power [Mw(e)l
Steam
Pressure (psig)
Steam
Temperature (OF)
Thermal Efficiency
Fuel Enrichment
Inventory
Fuel
Value of Fuel
Specific Power [Mw(e)/million
Average Conversion
(%I
(% U235)
(kg of U235)
($)
dol Iors]
Ratio
Isotope Produced
11.7
1.12
Pu239
30
7.5
0.74
Pu239
x lo6 EFR
94
600
740
31.3
28
485
GCR
225
950
950
32.1
2
2740
SGR
100
800
825
32.4
3.5
825
10.7
9.3
0.55
Pu239
OMR
150
415
550
26.1
1.5
1280
12.4
12.1
0.73
Pu239
DP R
180
950
540
28.7
1.5
78 2
7.6
23.6
?
Pâ&#x20AC;?239
ERR
22
600
825
30.2
148
2.5
8.8
0.52
u233
PWR
60
600
486
26.7
1.81
33.2
0.7
Pu239
YER
134
520
Saturated
26.2
550
6.6
20.4
0.64
Pu239
CER
140
600
485
34.7
> 90
275
4.7
29.8
0.47
u233
MSR
260
1800
1000
40.5
> 90
400
6.8
38.2
*EFR, GCR, SGR, OMR,
{
> 90 >90 0.72
2.6
75} 96
8.0
0.6-0.8**
Enrico Fermi Reactor, Solid Fuel Reactors, J. R. Dietrich and W. H. Zinn (eds.), Addison-Wesley, Reading, Mass., 1958. ORNL Gas Cooled Reactor, i b i d Sodium Graphite Reactor, Sodium Graphite Reactors, C. Starr and R. W. Dickinson (eds.), Addison-Wesley, 1958. Organic Moderated Reactor, Solid Fuel Reactors, loc. cit.
DPR, Dresden Nuclear Power Reoctor, Boiling Water Reactors, A. W. Kramer (ed.), Addison-Wesley, 1958. ERR, E l k River Reactor, ibid.
PWR. Pressurized Water Reactor, Shippingport P r e s s u r i z e d Water Reactor, R. T. Bayard et al., Addison-Wesley, 1958. Y ER, Yankee Atomic Electric Co. Reactor, Preliminary Hazards Summary Report, YAEC-60 (1957). CER, Consolidoted Edison Reoctor, Nuclear Reactor Data No. 2, Raytheon Mfg. Co., Walthom, Mass., 1956. MSR, Molten Salt Reactor, Fluid Fuel Reactors, J. A. Lane, H. G. MacPherson, and F. Maslan (eds.), Addison-Wesley, 1958. **Ratio depends on processing rate.
u233
a
ORNL-2751 Reactors-Power TID-4500 (14th ed.) 4
INTERNAL DlSTR IB UTION 1-10. 11. 12. 13. 14. 15. 16. 17. 18. 19. 20. 21. 22. 23. 24. 25. 26. 27. 28. 29. 30. 31. 32. 33. 34. 35. 36. 37. 38. 39. 40. 41. 42. 43. 44. 45. 46-47.
L. E. F. E. A. W. G. M. E. R.
G. S. F. P. L. F. E. A. J. B.
Alexander Bettis Blankenship Blizard Boch Boudreau Boyd Bredig Breeding Briggs W. E. Browning D. 0. Campbell W. H. Carr D. A. Carrison G. I. Cathers R. A. Charpie H. C. Claiborne F. L. Culler J. H. DeVan W. K. Ergen J. Y. Estabrook D. E. Ferguson A. P. Fraas W. R. Grimes E. Guth H. VI. Hoffman W. H. Jordan P. R. Kasten G. W. Keilholtz C. P. Keim M. T. Kelley F. Kertesz B. W. Kinyon M. E. Lackey J. A. L a n e R. N. Lyon H. G. MacPherson
48. 49. 50. 51. 52. 53. 54. 55. 56. 57. 58. 59. 60. 61. 62. 63. 64. 65. 66. 67. 68. 69. 70. 71. 72. 73. 74. 75. 76. 77. 78. 79-82.
W. D. Manly E. R. Mann L. A. Mann Vi. B. McDonald H. J. Metz R. P. Milford J. W. Miller K. Z. Morgan G. J. Nessle A. M. Perry P. M. Reyling J. T. Roberts M. T. Robinson M. W. Rosenthal
H. W. Savage A. W. Savolainen A. J. Shor M. J. Skinner J. A. Swartout A. Taboada R. E. Thoma M. Tobias D. B. Trauger F. C. VonderLage G. M. Watson A. M. Weinberg M. E. Whatley G. D. Whitman G. C. W i l l i a m s C. E. Winters J. Zasler ORNL - Y-12 Technical Library, Document Reference Section 83- 102. Laboratory Records Department 103. Laboratory Records, ORNL R.C. 104-105. Central Research Library
EXTERNAL DlSTRlBUTlON 106. F. C. Moesel, AEC, Washington 107. Division of Research and Development, AEC, OR0 108-695. Given distribution a s shown i n TID-4500 (14th ed.) under Reactors-Power category (75 copies - OTS)
21