Page 1

3 q45b 03bL387 3 T

.

4" ,

..

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.

N e i t h e r t h e U n i t e d States,

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

or usefulness of t h e informotion c o n t a i n e d i n t h i s report, or t h a t t h e u s e o f

informotion,

apparatus,

method, or process d i s c l o s e d i n t h i s report may n o t i n f r i n g e

p r i v a t e l y owned righ?s; or

5 . Assumes any l i a b i l i t i e s w i t h respect t o the use of, or for damages r e s u l t i n g from the use of any information, apparatus, method, or process d i s c l o s e d i n t h i s report. As

used i n t h e above, "person a c t i n g on b e h a l f of t h e Commission"

contractor or

of t h e Commission,

contractor

i n c l u d e s a n y employee or

or employee of s u c h contractor, t o t h e e x t e n t t h a t s u c h employee

of t h e Commission,

or employee of s u c h contractor prepares,

disseminates,

or

p r o v i d e s a c c e s s to, any information pursuont t o h i s employment or contract w i t h t h e Commission, or

h i s employment w i t h such contractor.


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

ORNL-2751  

http://www.energyfromthorium.com/pdf/ORNL-2751.pdf

Read more
Read more
Similar to
Popular now
Just for you