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ORNk/TM-7207 Dists Category uc-76

C o n t r a c t No. W-7405-eng-26

E n g i n e e r i n g Technology Division

CONCEPTUAL DESIGN CHARACTERISTICS OF A DENATURED MOLTEN-SALT REACTOR WITH ONCE-THROUGH FUELING

J. R. Engel H. F. Bauman J. F. Bearing

Date Published:

W. W. G r i m e s H. E. McCoy

W. A. Rhoades

July 1980

NOTICE This document contains information of a preliminary nature. It is subject to revision or correction and therefore docs not represent a final report.

P r e p a r e d by t h e OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 o p e r a t e d by UNION CARBIDE CORPORATION f o r the BEPARTNENT OF ENERGY

:.:.

.,:

...

iii

CONTENTS

.......................................................... 1. INTRODUCTION .................................................. 2 . CXNEML DESCRIPTION OF DESR 2.1 Fuel Circuit ............................................. 2.2 Coolant Circuit .......................................... 2 . 3 Balance-of-Plant ......................................... 2 . 4 Fuel Handling and Processing ............................. 3 . REFERENCE-CONCEPT DMSR ........................................ 3.1 Neutronic Properties ..................................... 3.1.1 Neutronics core model ............................. 3.1.2 Core design considerations ........................ 3 . 1 . 3 Neutronics calculation approach ................... 3.1.4 Once-through system Considerations ................ 3.1.5 Static neutronic results .......................... 3,1.6 Burnup r e s u l t s .................................... 3.1.7 Dynamic effects ................................... 3 . 2 Reactor Thermal Hydraulics ............................... 3.3 Fuel Behavior ............................................ 3.3.1 Basic considerations .............................. 3.3.2 Fission-product behavior .......................... 3 . 3 . 3 Fuel maintenance .................................. 3 . 4 Reactor Materials ........................................

ABSTMCT

B 2

5o s e e e .OO..O.eseB..O..OO.eeoeeeoe.oo

6 7

7 8

10 10

10 14 14 20

29 33 40

46 47

60 70 $1

.................................. ......................................... 3 . 5 Safety Considerations .................................... 3 . 6 Environmental Considerations ............................. 92 Antiproliferation Features ............................... 94 3 . 7 . 1 Potential sources of SSNM ......................... 94 Accessibility of ............................. 95 4 . ALTERNATIVE CONCEPTS ..................................... 4.1 Cycle Choices ....................................... Break-even breeding ............................... $8 4 . 1 . 2 Converter operation with fuel processfng .......... 100 4.1.3 Partial fissi~n-pr~duct~~IIIOVELI ................... 101 3.4,1

Structural a l l o y

3.4.2

Moderator

81 87 98

3.7

3.7.2

SSW

DMSR

Fuel

4.1.1

4.1.4

4.2

Salt replacement

F U ~ Pcycle performance

.................................. ...................................

104 105

iv Page 109 109 110 110 112 lb4 614 114

116 117 120 121 124

127 127 131 131 135 136 137

139 140 149

CONCEPTUAL DESIGN CHARACTERISTICS OF A DENATURED MOLTEN-SALT REACTOR WITH ONCE-THKOUGB FUELING J. 8. Fngel H. F. Rauman J. R e Dearing

W. R. Grimes PI. E. McCoy V. A. Rhoades

ABSTRACT A study was wade to examine the conceptual feasibility of a molten-salt power reactor fueled with denatured 235U and operated with a minimum of chemical processing. Because suck a reactor would n o t have a positive breeding gain, reductions in the fuel conversion ratio were allowed in the design t o achieve other potentially favorable characteristies for the reactor. A conceptual core design was developed in which the power density was low enough to allow a 30-year life expectancy of the moderator graphite with a fluence limit o f 3 x 1026 neutrons/rn* (E > 50 kev). mis reactor could be made critical with about 3450 kg of 20% enriched 2 3 5 U and operated for 30 years with routine additions of denatured 2 3 5 U and no chemical processing for removal of fission products. The lifetime requirement of natural U j O g f o r this once-through fuel cycle would be about 1810 Mg (-2000 short tons) for a 1-GWe plant operated at a 75% capacity factor. If the uranium in the fuel at the end of life were recovered ( 3 1 6 0 kg fissile uranium at -10% enrichment), the U3Og requirement could be further reduced by nearly a factor oâ&#x201A;Ź 2. The lifetime net plutonium production f o r this fuel cycle would be only 736 kg for all i s o topes ( 2 3 8 , 239, 2 4 0 , 2 4 1 , and 2 4 2 ) . A review of the chemical considerations associated with the conceptual fuel cycle indicates that no substantial dlfficulties would be expected if the soluble fission products and higher actinides were allowed t o remain in the fuel salt for the life sf the plant. Some salt treatment to counteract oxide eontamination and to maintain t h e oxidation potential of the melt probably would be necessary, but these would require only well-known and demonstrated te~hnobogy. Although substantial technology development would be required, the denatured molten-salt reactor concept apparently could be made commercial in about 30 years; if the costs of intermediate developmental reactors are included, the cost for development is estimated t o be $3750 million (1978 dollars). The resulting system would be approximately economically competitive with current-technology light-water reactor systems.

INTRODUCTION

Molten-salt reactors’ (PISRS) have been under study and development in the United States since about 1947, with most of the work since 1956 directed toward high-performance breeders for power production in the Th-233U fuel cycle.

The most recent development effort in this area was

terminated in September 1976 in response to guidance2 provided by the Energy Research and Development Administration (now Department of Energy) (ERDA/DOE)

in March 1976.

A brief s t u d y of alternative MSRs3 which em-

phasized their antiproliferation attributes was carried o u t in late 1376. This study concluded that MSRs without denatured fuel probably would n o t be sufficiently prs%iferation-resistant f o r unrestricted worldwide deploy-

Subsequently, a more extensive study was undertaken at Oak Ridge

ment.

National Laboratory (ORNL) t o identify and characterize denatured moltensalt r e a c t o r (B8fSR) concepts for p o s s i b l e a p p l i c a t i o n in antiproliferation

situations, This work began as p a r t of t h e effort initiated by ERDA in response to a nuclear policy statement by President Ford on October 28,

1 9 7 6 ; ~it was continued under the Nonproliferation ~lternativesystems ~ssessmentprogram (NASAP) $ 5 which was established in response t o the e

Nuclear Power P o l i c y Statement by President Carter on April 7, 1977,6 and

TIE National ~ n e r g y~lan.’ The DMSR is only one of a large number of reactors and associated

fuel cycles selected f o r study under NASAP.

However, it is also a member

of a smaller subgroup t h a t would operate primarily

cycle.

QII

the ~ h - 2 ”fuel ~

Molten-salt X - ~ ~ C ~ Qin T Sgeneral, , are particularly well suited t o

this fuel c y c l e because the fluid fuel and the associated core design tend

to enhance neutron economy, which is particularly important f o r effective res~urceutilization.

In addition, the ability of the ~ ~ ~ l tf uee nl to re-

~ in tain plutonium (produced from neutron captures in the 2 3 8 denaturant)

a relatively inaccessible form appears to contribute to t h e proliferation

resistance of the system.

The MSR concept also offers the possibility of

system operation within a sealed containment from which

IIQ

fissile mate-

ria% is removed and to which only denatured f u e l o r fertile material is

added during the life of the plant.

This combination of properties sug-

gests the possibility of a fuel. cycle with a %ow overall cost and significant resistance to p r ~ l i f e r a t i ~ n .

The primary purpose of this study w a s t o identify and characterize one o r more BXSR concepts with antiproliferation attributes at least equivalent to those of a "conventional" light-water reactor (EWR) operating on a once-through f u e l cycle.

The systens were a l s o required to

show an improvement over the LWR in terms of fissile and fertile resource Considerable effort was devoted t o characterizing features

utilfzation.

of the concept(s) that would be expected to affect the assessment of their basic technological feasibility. costs

These features included the estimated

and time schedule for developing and deploying the reactors and

their anticipated safety and environmental features. Although t h e older MSR studies were directed toward a high-performanse breeder [and a reference molten-salt breeder reactor (MSBR) design8

was developed], the b a s i c concept is adaptable to a broad range of fuel cycles, Aside from the breeder, these f u e l cycles range from a plutonium burner

~ Q 2P 3 3 ~ production,

through a DMSR with break-even breeding ana

complex on-site fission-product processing 9 9 to a denatured system with

a 30-year fuel cycle that is once-through with respect t o fission-product cleanup and fissile-material recycle.

Of these, the Past one currently

appears to offer t h e most advantages for development as a proliferationresistant power

SQUPC~.

Consequently, this report is concentrated on a

conceptual DMSR with a %@year fuel cycle and no special chemical processing for fission-product removal; other alternatives are considered only b r i e f l y -

Section

contains a general d e s c r i p t i o n of the

BMSR c o n c e p t , with

emphasis on those features that would be the same for a l l BMSR fuel cycles.

Section 3 presents a more detailed treatment of the reference-

concept DMSR covering the neutronic and thermal-hydraulic characteristics of the reactor core, fuel-salt chemistry, reactor materials, plant safety

considerations, and system-specific environmental considerations.

A gen-

eral treatment of the antipsoliferation attributes of the concept is a l s o included.

The next section (Sect. 4 ) addresses potential alternatives to

the reference concept and their perceived advantages and disadvantages. Section 5 addresses the commercialization considerations for DMSRs, including the perceived status, needs, and potential research, development,

4 and demonstration (RD&D) program; a p o s s i b l e s c h e d u l e

~ Q major P

e~n-

struetion projects; t h e e s t i m a t e d performance of commercial units; and any s p e c i a l licensing consideratbons.

Finally, Sect. 6 p r e s e n t s the gen-

eral cbnclusions of the study, along w i t h suggestions t h a t would a f f e c t any f u r t h e r work on t h i s concept.

5 2.

.. .... .

w..w

GENEMI, DESCRIPTION OF DMSR

The p l a n t concept f o r a DMSR is a d i r e c t outgrowth of t h e ORNZ r e f e r e n c e - d e s i g n MSBR, and, t h e r e f o r e , i t c o n t a i n s many favorable feat u r e s of the b r e e d e r d e s i g n .

However, t o comply w i t h t h e a n t f p ~ ~ l i f e ~ a -

t i o n g o d s , i t a l s o c o n t a i n s a number o f d i f f e r e n c e s , p r i n c i p a l l y i n the reactor

CQPQ

d e s i g n and t h e f u e l c y c l e .

F i g u r e 1 i s a s i m p l i f i e d sche-

matic diagram of t h e r e f e r e n c e - d e s i g n MSBR.

A t t h i s l e v e l of d e t a i l ,

t h e r e i s o n l y one d i f f e r e n c e from t h e DMSR concept:

t h e o n - l i n e chemical.

p r o c e s s i n g pliant (shown a t t h e l e f t of t h e c o r e ) would n o t be r e q u i r e d

f o r t h e DMSR.

ORNL-DWG 68-1185EA

PRIMARY SALT PUMP

,- \

r-' NaRF4

SECONDARY SALT PUMP

MODERATOR

r CHEMICAL PROCESSING PLANT

FUEL SALT

STEAM GENERATOR

Fig. 1.

...

Single-fluid,

two-region molten s a l t breeder r e a c t o r .

2. I

Fuel Circuit

The f u e l c i r c u i t f o r a DNSR would be v e r y similar t o t h a t f o r a n f.1SBR; o n l y t h e c o r e d e s i g n would be changed.

f o r t h i s redesign

The primary requirement

a realaction in the c o r e neutron faux (ana power

density) to

extend t h e life expectancy of t h e g r a p h i t e moderator t o t h e f u l l 30year p l a n t %ifetLaae,

limit n e u t r o n c a p t u r e s i n 2 3 3 ~ awhich, t o enhance p r o ~ ~ f e r a t i oren s i s t a n c e , would be r e t a i n e d i n t h e f u e l s a l t .

me lower power d e n s i t y would a l s o tend t o reduce t h e poisoning e f f e c t s

of s h o r t - l i v e d f i s s i o n p r o d u c t s and t o s i m p l i f y t h e thermal-hydraulic cons t r a i n t s on t h e d e s i g n of t h e moderator elements,

The p r i n c i p a l unfavor-

a b l e e f f e c t s would be t h e i n c r e a s e s i n i n v e n t o r y of t h e f u e l s a l t and fis-

sile fuel.

Reference-design features of t h e DNSR core a r e d e s c r i b e d i n

g r e a t e r detail in a later s e c t i o n . A t design power (h000 MWe),

dug

* temperature

the f u e l salt, which would have a l i q u i -

o f about 5OO"c, would enter t h e core at 5 6 6 " ~and leave

at 704'C t o t r a n s p o r t about 2250 PlWt ( i n f o u r p a r a l l e l loops) t o t h e secondary s a l t ,

The f l o w r a t e of salt i n each ~f the primary l o o p s ( i n c l u d -

ing the bypass f o r xenon s t r i p p i n g )

be a ~ ~ Iu &/s t

(a6,o00 gpm>.

IFfme primary salt would ~ ~ n t a 0.5 i n t o 1% (by volume) helium bubbles t o s e r v e as a s t r i p ing agent f o r xenon and o t h e r v o l a t i l e f i s s i o n p r o d u c t s , H e l i m W Q Ube I ~added t o and removed from bypass flows of -10% o f each

of the primary hoop flowse This g a s s t r i p p i n g would a l s o remove some of the tritium from t h e primary salt,t p a r t l y as ~ H F ;p~owever, nost of the t r i t i u m would d i f f u s e through t h e t u b e w a l l s of the primary heat exchang-

e r s i n t o t h e secondary salt.

Helium reraoved from t h e primary c i r c u i t

would be treated i n a s e r i e s of f i s s i o n - p r o d u c t t r a p p i n g and c l e a n u p s t e p s b e f o r e being r e c y c l e d f o r f u r t h e r gas s t r i p p i n g .

*%he t e m p e r a t u r e cooling

P ~ - o v i s i o n swould a l l s ~be

a t which t h e f i r s t c r y s t a l s appear on e q u i l i b r i u m

P~stimatesare that 18 to 19% 0% t h e t o t a l t r i t i w produced w m l ~ lbe removed i n t h i s gas,

made in the primary circuit t o remove and return fuel salt without opening ~ ~required t ~ to the primary containment and t o a d d fuel-salt c ~ n s t i t u as

maintain the chemical condition o f the salt,

The secondary, a r coolant-salt, circuits for the DMSR w ~ u l dbe identical to those developed for the reference-design MSBR.

The nominal flow

rate of the secondary salt (a eutectic mixture of NaBF4 and NaF) would be a b o u t 1.26 m 3 / s (20,000 gpm> in each of the four loops, w i t h a temperature

rise from 4 5 4 t o 621째C in t h e primary heat exchangers.

This salt would be

used t o generate supercritical steam at about 540째C and 25 MPa to drive the turbine-generator system. * In addition to its primary functi~nsof isolating the highly radis-

active primary circuit from t h e steam system and serving as an intemediate heat-transfer f l u i d , the socliuw flusroborate s a l t mix-eure t a ~ ~ l d play a major role in limiting the release of tritium from the DNSR s y s tem.

Engineering-scale tests in 2-9-16 (Ref. 1 0 ) demonstrated that this

salt is capable of trapping large quantities of tritium and transforming it to a less mobile, b u t s t i l l volatile, chemical form that transfers to

the cover-gas system rather than diffusing through the steam generators to the water system.

Consequently, the majority of the tritium (-80%)

wsuEd be trapped o r condensed out o f the secondary circuit cover gas, and less t h a n 0.2% of t h e total

2.3

releasea.

Balance-of-Plant

The balance-of-plant for a DNSR primarily would be identical to that f o r an MSBR.

Because the same salts and basic parameter values are Pn-

volved, there would be no basis f o r changing the normal auxiliary systems required for m~mnz~l plant operation. i n same o f the safety systems.

Differences, however, c o u l d appear

Because of the l ~ w e rpower density in the

ha he supercritical steam cycle appears to be particularly well suited to this concept because of the relatively h i g h melting temperature (385'6) of the secondary salt and the desire to avoid salt freezing in the steam generators. :.% :& ...... .,. .....

ii....

8 DMSR, t h e s h u t d ~ w nresidual-heat-remova1

vere t h a n i n the MSBR.

(RHR) probPem would be Less se-

Consequently, a less e l a b o r a t e RHR system t h a n

would be needed f o r a n MSBR might be a c c e p t a b l e f o r a DMSR.

However, f o r

p u r p o s e s of c h a r a c t e r i z i n g t h e DMSK, t h e assumption w a s that the balanceo f - p l a n t would be t h e same a s t h a t f o r a n EISBW, 2.4

F u e l Randline and P r o c e s s i n g

The performance of an MSBR wouEd be s t r o n g l y dependent

the avail-

a b i l i t y of a n o n - s i t e c o n t i n u o u s chemical-processing f a c i l i t y f o r removal of fission p r o d u c t s and i s o l a t i o n of p r o t a c t i n i u m on r e l a t i v e l y s h o r t t i m e These t r e a t m e n t s would make p o s s i b l e t h e achievement of a posi-

cycles,

t i v e 2 3 3 U b r e e d i n g g a i n i n a system with a low s p e c i f i c f i s s i l e inventory.

Because a DPfSR on a 30-year f u e l c y c l e would n o t r e q u i r e even nomi-

n a l break-even

breeding and because a significantly higher f i s s i l e inven-

t o r y could be t o l e r a t e d , t h e p r o c e s s i n g r e q u i r e m e n t s f o r a DMSR would be much l e s s s t r i n g e n t t h a n f o r an MSBR.

I s o l a t i o n of p r o t a c t i n i u m would be

avoided f o r p r o l i f e r a t i o n r e a s o n s , and chemical p r o c e s s i n g t o remove â&#x201A;Źiss i o n p r o d u c t s c o u l d b e avoided w i t h o u t s e v e r e performance p e n a l t i e s . D e s p i t e t h e s e c o n c e s s i o n s , some f i s s i o n - p r o d u c t removal would t a k e g l a c e i n any PISR.

Most of t h e r a r e g a s e s (and some o t h e r v o l a t i l e f i s -

s i o n p r o d u c t s ) would be removed by t h e gas-sparging s y s t e m i n t h e primary circuit.

~n addition, a s u b s t a n t i a l fraction of t h e noble-metal"

fis-

s i o n p r o d u c t s would be e x p e c t e d t o p l a t e o u t on metal s u r f a c e s where t h e y

would n o t a f f e c t t h e n e u t r o n i c performance.

However, t h e r e f e r e n c e - d e s i g n

r e d u c t i v e - e x t r a c t i o n / m e ~ ~ ~ - t process ~ ~ ~ s â&#x201A;Ź ~would ~ n o t be involved. Although t h e r e would be no chemical p r o c e s s i n g f o r f i s s i o n - p r o d u c t removaih, t h e DMSR l i k e l y would r e q u i r e a h y d r o f E u o r i n a t i o n system f o r o c c a s i o n a l (presumably b a t c h w i s e ) t r e a t m e n t of t h e s a l t t o r e m ~ v eoxygen contaminatione

Ira addition, because a BMSR would r e q u f r e routine addi-

t i o n s of f i s s i l e f u e l , a s w e l l as a d d i t i o n s of o t h e r materials n e c e s s a r y

t o keep t h e f u e l - s a l t chemical composition in p r o p e r b a l a n c e , a chemical

*N o b i l i t y

i s d e f i n e d h e r e i n r e l a t i o n t o t h e U4+/lJ3+ ( s e e Sect. 3 . 3 . 2 ) .

redox p o t e n t i a l

addition

StathQR

would be required.

The %eChnOlogy for b o t h Of these

operations is well established and was extensively demonstrated in the molten-salt reactor experiment (MSWE).

These and other aspects o f the

DMSR fuel chemistry are treated in greater detail in a later s ~ t i o n .

REFERENCE-CONCEPT DMSR

A preliminary conceptual design has been developed for a DMSR oper-

ating on a 30-year f u e l cycle.

The emphases to date have been on the re-

a c t o r core d e s i g n a n d fuel cycle, with less attention to other aspects of

the system.

Although t h f s design establishes t h e b a s i s concept and char-

acterizes its major properties, it is tentative and would be s u b j e c t t o major refimement and revtsion if a substantial d e s i g n effort were under-

taken. 3.1

Neutronic Properties

The basic features of this BMSR concept which distfnguish it from o t h e r MSRs are established primarily by the r e a c t o r care design and its a s s ~ c i a t e dneutronic properties.

'Phe d e s i g n described here represents

the results of a first-round effort to balance some of t h e many variables invo%ved in a r e a c t o r COT@, b u t it is by

3.1.1

PPO

means an o p t h i z e d design.

Weutronics core model

Prom a neutranics point of view, the core is simply designed as f ~ l -

EoWs ( F J g . 2 ) .

The core and r@f%e~tor fill a right circular cylinder that is

18 rn in diameter and 18 m high.

The core, which is a cylinder 8.3 m in

diameter and 8,3 m high and centered inside the larger volume, f s f i l l e d with cylindrical graphite l o g s in a triangular array of 8.254-m Approximately 95X of t h e

C Q F ~( c o r e B )

pitch.

has l o g diameter of 0.254 m, with

the fluid f u e l filling the interstftial volume to praduce a fuel volume

fraction of 9.31%.

An axial c y L i n d r i e a l h o l e of 0,851-m

diam in the een-

ter of each l o g admits another 3 . 6 3 % fuel for a t o t a l of 12.94 voP %.

achieve f l a t t e n i ~ gof t h e f a s t flux and thus maximize the lifetime of the

graphite moderator, the remaining 5% of t h e core (core A > , a cylinder 3 m in dfameter and 3 m h f g h , has a l o g diameter of 0.24 m, resultfng in a

total fuel volume fraction of 20.00% i n this zone.

ORML-DWG 80-4263 ETD

RADIAL GAP

CORE

MI BPLANE

BOTTOM PLENUM

BOTTOM REFLECTOR

-1.504

B).Q5=+

4.15

Fig. 2. DMSR core model f o r neutronic s t u d i e s Call dimensions in meters). ! I

.. .... .

. \.... .. ,/ .

- eylindrfcal

geomgtry

The r a d i a l r e f l e c t o r is g r a p h i t e 0.8 m t h i c k and i s a t t a c h e d t o

t h e r e a c t o r v e s s e l at t h e 18-m diam.

T h i s l e a v e s a gap of 0.05 m f i l l e d

with f u e l s a l t surrounding t h e c o r e l a t e r a l l y .

3. The i n l e t and o u t l e t p l e n a cover b o t h t h e c o r e and r a d i a l gap t o t h e i r full d i a m e t e r and a r e each 0.28 m t h i c k .

They c o n s i s t of 50% s t r u c -

t u r a l g r a p h i t e and 50X f u e l .

4. The axial r e f l e c t o r s are each 0.65 m thick and extend to t h e f u l l 1Q-mdiam.

A l l r e f l e c t o r r e g i o n s c o n t a i n a small amount of f u e l salt f o r

c o o l i n g , which i s e s t i m a t e d as 1 v o l

X a t o p e r a t i n g temperature.

A l l s t a t e d dimensions are assumed t o a p p l y a t nominal o p e r a t i n g

conditions.

During system h e a t u p , t h e l e n g t h and d i a m e t e r of t h e c o r e

v e s s e l are assumed t o i n c r e a s e a t t h e r a t e of expansion of Hastelhy-N. The r e f l e c t o r s are assumed t o expand a t t h e expansion rate of g r a p h i t e b u t to remain a t t a c h e d t o t h e v e s s e l .

Because g r a p h i t e expansion i s less

t h a n t h a t of t h e v e s s e l , t h i s w i l l r e s u l t i n admatting a d d i t i o n a l s a l t t o t h e r e f l e c t o r zones.

The c o r e and plenum r e g i o n s are assumed t o expand

r a d i a l l y o n l y a t t h e expansion r a t e of g r a p h i t e , which will e s t a b l i s h t h e t h i c k n e s s s f t h e r a d i a l gap.

The a x i a l c o n f i g u r a t i o n i s a f f e c t e d by t h e

l o g s f l o a t i n g upward i n t h e s a l t and by t h e lower plenum being c o n s t r u c t e d

s o t h a t i t always c o n t a i n s 58% s a l t .

The t h i c k n e s s e s of t h e c o r e and t h e

upper plenum, t h e n , i n c r e a s e a t t h e g r a p h i t e expansion r a t e , b u t t h e lower plenum grows a t such a r a t e as t o span t h e gap between t h e c o r e and t h e bottom r e f l e c t o r . Mechanical p r o p e r t i e s used f o r t h e p r i n c i p a l c o n s t i t u e n t s are sumThe s a l t i s taken to have t h e nominal chemical corn-

marized i n Table 1.

position shown i n Table 2.

The tern " a c t i n i d e s ' o i n t h i s s t u d y r e f e r s to

all elements of atomic numbers > 90 and n o t j u s t t o t r a n s p l u t o n i u m elemerits.

The a c t i n i d e p e r c e n t a g e i s s u b j e c t t o s m a l l v a r i a t i o n s depending

on t h e f u e l cycle and t h e h i s t o r y of t h e f u e l . The i n v e n t o r y of- f u e l s a l t , both i n and o u t of t h e c o r e , i s summar i z e d i n Table 3 .

This i s b e l i e v e d t o be a generous estimate s f t h e re-

q u i r e d inventory f o r a I-GWe system.

The thermal energy y i e l d p e r fission

i s assumed t o be 190 MeV f o r t r a n s l a t i o n of abss%ute f i s s i o n rates t o eff e c t i v e power l e v e l .

........... ........

Table 1.

Reference properties of f u e l s a l t and moderator f o r a DMSR

Value

Characteristic Graphite moderatsr density, M g l m 3

1.84 3, l o F u e l - s d t d e n s i t y , Mg/m 3 Graphite linear thermal expansion, x 10-6 ~ - 1 4.1 vessel linear thermal expansion, x K-1 17.1 Fuel voPumetrie thermal expansion, x K-â&#x20AC;&#x2122; 208

Table 2.

Nominal chemical composition Of DMsR f u e l salt Molar percentage

Mat er ia 1 k i p

74.0

BeF2

16.5

XFba

9.5

Fission products

ax refers

Table 3.

Trace

t o all actinides,

DMSR fuel-salt inventory

Lo cation

VoPume (m3>

core

59.4

Top and bottom plenums

11.1 10.9 3.0 20.0

Radial gap Re f 1e c tors External loop

104.0

The s i z e of t h e c o r e was determined s o as t o a l l o w a g r a p h i t e mode r a t o r l i f e t i m e e q u a l t o the d e s i g n l i f e t i m e of t h e p l a n t .

As compared

with a smaller c o r e , t h i s r e s u l t e d i n lower n e u t r o n l e a k a g e , h i g h e r invent o r y of f i s s i l e material, and lower loss of p r o t a c t i n i u m due t o n e u t r o n capture.

If h i g h e r l e v e l s of g r a p h i t e exposure were i n d i c a t e d by f u t u r e

d a t a o r d e c i s i o n s , a smaller c o r e would probably be chosen.

The c i r c u l a r c y l i n d e r moderator shape resists b i n d i n g e f f e c t s t h a t can occur wfth o t h e r shapes.

The h o l e i n t h e c e n t e r is s i z e d to p r o v i d e

d e s i r a b l e resonance s e l f - s h i e l d i n g without undue thermal flux d e p r e s s i o n .

The l a t t i c e p i t c h i s simply a convenient one from both t h e r m a l and neut r o n i c points of view.

me reduced d i a m e t e r of t h e central section of t h e

logs w a s a d j u s t e d t o g i v e t h e proper d e g r e e of n e u t r o n f l u x f l a t t e n i n g ,

There i s no doubt t h a t f l u x f l a t t e n i n g r e s u l t s i n more core l e a k a g e , s l i g h t l y degraded b r e e d i n g , and more f l u x i n the r e a c t o r v e s s e l a s compared with a n u n f l a t t e n e d c o r e .

The u n f l a t t e n e d c o r e , however, would

have a muck l a r g e r volume and much l a r g e r i n v e n t o r y of fissile material f o r the same maximum n e u t r o n damage flux. The thorium c o n c e n t r a t h n of t h e s a l t h a s been a d j u s t e d t o g i v e near-

optimum long-term c o n v e r s i o n and a l o w requirement f o r makeup f u e l .

approach l e a d s t o a r e l a t i v e l y h i g h i n - p l a n t

have economic d i s a d v a n t a g e s .

'Ibis

f i s s i l e i n v e n t o r y , which may

Thus, o v e r a l l o p t i m i z a t i o n might s u g g e s t

more f a v o r a b l e combinations of i n v e n t o r y and makeup.

The o t h e r actinide

C o n c e n t r a t i o n s a r e d e t e m i n e d by t h e v a r i o u s f u e l i n g p o l i c i e s c o n s i d e r e d and Bay the o p e r a t i n g h i s t o r y of t h e f u e l .

3.1.3

Neutro~iesc a l c u l a t i o n approach

3.1.3.1 The o v e r a l l approach was designed t o c o u p l e numerous computer runs of r e l a t i v e l y s h o r t d u r a t i o n .

The o b j e c t i v e s were good a c c u r a c y 3 rela-

t i v e l y quick computer r e s p o n s e , and t h e a b i l i t y t o r e p e a t and r e v i s e d i f f e r e n t p o r t i o n s of the procedure a s the d e s i g n evolved,

Initial scoping studies showed that the self-shielding of thorim and 2 3 8 ~has

a most critical effect on the system neutronic%, while that og

the other uranium nuclides was comparatively less.

Concentrations of

protactinium, neptunium, and plutonium remained small enough t o make selfshielding treatment of those nuclides necessary.

The effest of

P~SQ~BIEC~

overlap between 232Th and 238U was of particular interest and was studied in some depth using the ROLAIDS module of the AMPX code system. I l

'%he

conclusi~nswere that thPs effect c o u l d be ignored safely in the p r e s e n t s t u d y and that treatment of the effect would have been burdensome had it

been required. Statics.

A set sf cross sections for the more signifisant nuclides

(Table 4 ) was prepared based ow the ENDF/B Version 4 set of standard cross sections.12

A t o t a l o f 1 2 2 energy groups was rased, with boundaries as

listed in Table 5.

Downscatter from any group t o any other was allowed,

and upscatter between all groups below 1.86 e V w a s allowed.

The 123-gromp

set was t h e n repr~cessedto enforce strict neutron ~ o n s e r ~ a t i ~ nThis . was

especially important in the case of graphite.

Table 4 . Nuclides in library of 123 energy groups used for DMSR study

Self-shielding of thorium and uranium nuclides was treated using the NITAWL module of t h e AMPX socle system.

was selected in each case.

The Nordheim integral treatment

The geometric parameter applicable t~ the tsi-

cusp f u e l area between the l o g s was determined by a special Monte Carlo computer sode devised by J. R. Knight of ORNL.13

Figure 3 illustrates

XSDREU' 123-gKOug e n e r g y s t r u c t u r e

Table 5.

Boundaries

Group Energy 1 2

3 4 5 6 7 8 9 10 11

12 13 14 15 16

17 18

19 20 21 22 23 24 25 26 27 28 29 30 31 32 34 34 35 36 37

38 39 40 41 42

Boundaries Group Energy

Lethargy

1 e 491 8E07 --ae40 1.3499E07 a.30 1.22 14E07 -(le 20 4.IO B.1052E07 E. mQOE07 0.0 9 e 094 8E0 6 0.10 8.1873E06 0.20 7.4082E06 0.30 0.40 6.7032E06 6 065 3E06 0.50 0.60 5.4881E06 0.70 4.9659E06 0.80 4.4933E06 4 e 065 7E06 0.90 1.00 3.6788E.06 1.10 3.3287E06 3-01 1%06 1.20 1.30 2. 7253E06 1.40 2.4660E06 2.2313E06 1-50 2.0190E06 1.60 E e 8268E06 1.70 B o 6530E06 1.80 I .4957E06 1.90 2.00 1.3534E06 2.10 1 2246E06 2.20 1.1080E06 1 e 0026E06 2.30 9 e 0 7 18EO 5 2-40 2.50 8.2085E05 7 e 4274E05 2.60 6.7~06~05 2.70 2.80 6 08 1OEO 5 2.90 5.5024E05 3.00 4.9787E05 4,5049E05 3.10 4.0762E05 3.20 3.6883E05 3.30 3.33 7 3 ~50 3.40 3.50 3.0197E05 2.7324EO5 3.60 2.4724E05 3.70 I

43 44 45 46 47 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 34 75 76 77 78 79 80 81 82 83

Boundaries Group

2.2371E05 2.0242E05 1 8316E05 1,6573E05 1a 4996E05 1 0 3569E05 1 2277E05 I 1109E05 8.651 7E04 6.7379E04 5.247SE04 4.0868E04 3.1828E04 2 4788E04 1 9305E04 1 5034E04 1 1709E04 9.1188E03 7.1017E03 5.5308E03 4.3074E03 3.4546E03 2 e 6 126E03 2.0347E03 1 5846E03 1.2341E03 9.61 12E02 7.4852~202 5 a 829 5E02 4.5400E02 '3.5357E02 2 a 7 5 34E02 2.1145E02 1 6702E02 1 e 3O0SE02 1 e 01 3OE02 9.8893EOl 6.1442EOL 4.7851EOE 3.7267E01 2.9023E01

Lethargy 3.80 3.90 4.00 4. IO 4.20 4.30 4.40 4.50 4.75 5-00 5.25 5.50 5.75 6.00 6.25

84 85 86 87 88 89 90 91 92 9% 94 95 96 33

6-50

6.75 7.00 7.25 7.50 7.75 8.00 8.25 8.50 8.75 9.00 9.25 9.50 9.75 10.00 10.25 10.50 10.75 11.00 11.25 11.50 11.75 12.00 12.25 12.50 12.75

100 101 102 103 104 105 106

109 108 109 110 111 112 113 114 115 116 117

118 119

120 121 122 E23 124

Energy

Lethargy

2.2603EO1 1.7603E01 1 371OE01 E .0670E01 8.3153E-01 6 4 76m-0 1 5.0435E-01 3.9279E-01 3.0590E-01 2 e 382 4E-0 1 1a 8 5 54E-0 1 1.7090E-OE I o 5670E-01 I 432OE-Q 1 I. 285OE-01 I. 134OE-01 9,9920E-02 8.8100E-0% 7.6840E-02

13.00 13.25 13.50 13.75 16.30 16.55 16.80 17.50 17.30 17.55 17,80 17.88 17.97 18.06 18.17

4.5520~-02

5.4880E-02 4 e 4 85 0E -02 3 e 6 146E-0 2 2. W40E-O2 2.49303,-02 2.0 7 1OE-0 2 I e 798OE-02 1.5980E-02 1 398OE-02 1 1980E-02 9.9700E-03 8.2300E-03 6,990OE-03 5. W00E-03 4.99OQE-03 3.9800E-03 2.98OOE-03 2.1100E-03 1 - 4900E-03 9.8000E-04 4,7000E-04

18.29

18.42 18.55 18.68 18.84 19.02 19.22 19.44 19.63 19.81 20.00 20. 14 20.25 20.39 20.54 20.73 20.92 21.08 21.24 21-42 21 e 64 21.93 22.28 22.63 23.05 23. 7Bb

a ~ x xc o r r e s p o n d s t o 1 O ~ X . boundary o f g r o u p 123.

I..

.......

u . ; w

BWNL-DWG 804264 ETB I

0.45

E 6>

k0.50 B . I S N

8.25

0.80

0.85 0.98 0.95 ROD DIAMETERPITCH RATIO

1.00

Fig. 3. Mean chord l e n g t h of f u e l surrounding t r i a n g u l a r a r r a y s Circles i l l u s t r a t e p r e d i c t i o n s by t h e 4 8 l A r u l e . of moderator rods.

t h e r e s u l t s of t h i s t r e a t m e n t .

'Phe s a l t i n t h e plenum and r a d i a l gap re-

g i o n s w a s r e p r e s e n t e d a s a 0.05-m

p l a n e environment.

The r e s u l t i n g m u l t i g r o u p c r o s s s e c t i o n s were used w i t h t h e XSDRMPM module of AMPX t o accomplish a d i s c r e t e - o r d i n a t e s c e l l c a l c u l a t i o n i n t h e S-4 a p p r o x i m a t i o n and t o accomplish group r e d u c t i o n t o t h r e e e n e r g y g r o u p s , as shorn i n T a b l e 6.

A s e p a r a t e c e l l c a l c u l a t i o n was performed

f o r e a c h of t h e two l o g d i a m e t e r s .

Plenum and gap

C~QSS

s e c t i o n s were

weighted o v e r t h e spectrum of t h e smaller l o g d i a m e t e r because i t l i e s between t h e s t a n d a r d d i a m e t e r and t h e p u r e s a l t r e g i o n i n h a r d n e s s . The basic c o n c e n t r a t i o n s s f t h e n u c l i d e s were based on e s t i m a t e d midlife conditions.

A d d i t i o n a l cases of s e l f - s h i e l d i n g f o r t h e thorium

Energy range

Energy group Re so na nee

14.918 MeV to 52.475 eV 52.475 ev to 2.3824 eV

Thermal

'2.3824 eV to 0.88647 eV

Fast

and uranium nuclides were prepared for use in t h e depletion and reactiv-

ity coefficient studies.

These were weighted over the neutron spectra

calculated in the cell cal.culatPon.

The macrospatial effects were treated using the reduced cross secSeparate a x i a l and r a d i a l flux

tion set with t h e APC EI esmputer

proflies were found with mutually consistent flux and leakage resu1ts. Core heterogeneity was treated by transverse flux weighting of the de-

tailed geometry.

Reaction parameters necessary f o r burnup were deter-

mined from these results, with care taken t o combine all reactions representing a particular nuelear species regardless

cell or the identity of the c e l l involved.

Oâ&#x201A;Ź

positions in the

This is consistent with an as-

sumption of rapid fuel circulation and mixing. Burnup.

A simple burnup c o d e , QUAB, was devised to t r e a t the un-

usual requirements of this study.

Special features i n c l u d e the follow-

ing 1.

Sufficient 238U is added at d l times to maintain the denatured condition.

The %ho%-iuwl cQncentrat%Ql3 can be h e l d COnStaRt by automatic addition,

allowed to decline naturally, o r adjusted t o maintain constant total actinide concentration.

Periodic additions of enriched fissile material can be made.

P e s i o d f c withdrawals of f u e l can be made selectivePy by nuclide.

This

fuel can be held until the protactinium decays and then be reinserted selectively by nuclide into the machine.

The first removal is re-

placed with fuel identical to the initial loading.

Enriched material can be added activity margin.

QPI

demand to maintain a specified re-

The code calculates nuclide concentrations, total inventories, reactivity, and breeding r a t i o as a function of time. Treating the lengthy transplutonium and fission-product chains i n QUAB w a s not practical; multigroup data were not available far many of

Instead, the BRIGEN code15 w a s used with a library of cross sections" esthe required nuclides and were of dubious reliability for others.

pecially devised for its uses The BRIGEN results were then "patched into" the QUAB calculation directly. The burnup calculation allowed the cross sections of thorium and 2 3 8 ~

to vary continuously during the calculation; this was accomplished by in-

terpolation.

3 . 1 e 3.2

Evaluation

A s desired, the method provided relatively rapid response9 detailed

treatment of resonances, and a mulitigr~upspectrum and cell treatment. A l l details of the denatured fuel cycle were treated.

The expedient of

treating a range of thorium and 2 3 8 ~densities remevest the necessity of imbedding the expensive and tedious resenamce treatment inside t h e Loop f o r varying densities.

Deciding on the applicable range was not difficult

after a few initial tries. A system ~ouplingthe spatial calculation and depletion could be

used.

Kany such systems are available, although all would require exten-

s i v e modification f o r MSR u s e .

What of t h e cell calculation? T a b l e 7

shows the cell factors from our reference case which have been condensed to three energy groups.

This is clearly a heterogeneous core.

Further,

the actinide densities are continually changing, resulting in timedependent cell factors.

Studies beyond these would be required t~ prove

that a coupled system could be worthwhile without directly c o ~ p l e dcell. c a l culation s

The requirement to "umix" the revised nuclear densities after hav-

ing them lumped together during a depletion s t e p represents a complication that would thwart most existing codes.

However, this complication

must be coupled with logic to provide interpolation between cross-section sets representing various self-shielding situations. With o r without an

,... :.:..y w,

Cell material

Core zone

Energy

Fast Resonance The mal

A A A

Fast Resonance The ma P

Inner salt 1.14 0,97

0.94

1.2% 0.97

0.88

Mod era t o r

Interstitial salt

0.96 1.01 1.03

1.14 0 - 97 0*8%

Q,98 1.00 1.01

0.98 0.93

1.12

Average f l u x i n material d i v i d e d by a v e r a g e c e l l flux.

imbedded s e l l c a l c u l a t i o n , a n unusual code system c l e a r l y would be necess a r y to p r o v i d e

f d % y satisfying l e v e l of d e t a i l to t h i s problem,

Ob-

v i o u s l y , a true two-dimensional s p a t i a l t r e a t m e n t of t h e f l a t t e n e d core would be a p p r o p r i a t e , b u t imbedding s u c h a c a l c u l a t i o n i n s i d e a d e p l e t i o n l o o p i s expensive.

3.1.4

3 . 1 4.1

Once-through system c o n s i d e r a t i o n s

Fueling policy

For p u r p o s e s of n u c l e a r c a l c u l a t i o n s , t h e f u e l i n g p o l i c y for t h e once-through BMSR i s a s fohlows.

Thorium i s added t o a n i n i t i a l l o a d i n g of s a l t i n a s p e c i f i e d concentration.

During o p e r a t i o n , t h e c o n c e n t r a t i o n i s allowed t o d e c l i n e

v i a burnup. Near t h e end of p l a n t l i f e , small amounts a r e removed a s requ-ired t o keep t h e t o t a l a c t i n i d e c o n t e n t below t h e s t a r t u p value. 2.

Uranium is added a t t h e maximum a l l o w a b l e enrichment in t h e necessary t o maintain c r i t i c a l i t y .

ZLIIIQU~~

A d d i t i o n a l 238Y is added as r e q u i r e d t o m a i n t a i n t h e denaturing inequality

Removal of c e r t a i n fission products P s accomplished a c c o r d i n g t o Table 8.

T a b l e 8.

Removal times f o r f i s s i o n p r o d u c t s i n once-through c y c l e s

Fission-product group

Element

Removal time

Noble gases

Kr, X e

50 s

Seminoble and noble metals

Zn, G a , Ge, A s , N b , Mo, Tc, Ru, Rh, P d ,

2.4 h

A g , Cd, I n , Sn, Sb

3.1.4.2

Fission-product buildup

A s t u d y of 30-year f i s s i o n - p r o d u c t b u i l d u p was made a s a f ~ r t c t i ~o fn

v a r i o u s c o n t i n u o u s removal r a t e s f o r t h o s e p r o d u c t s n o t l i s t e d i n T a b l e 8. The r e a c t i v i t y e f f e c t may be s a t i s f a c t o r i l y r e p r e s e n t e d by

where p = fission-product

r e a c t i v i t y e f f e c t (%>,

t = t i m e (year),

= yield

(0.93%/year),

X = burnout r a t e (6.8 year)-', R = removal r a t e (year-1

"Other n o n f i s s i l e uranium wuclides f u r t h e r d i l u t e t h e 233U.; d i l u tion to 1 2 % 2 3 % may r e q u i r e additional 23%.

The fit t o d a t a r e p r e s e n t i n g f o u r E - ~ K W V ~ tEi m e s from f i v e y e a r s t o i n f i n i t y had a n a b s o l u t e s t a n d a r d d e v i a t i o n of 6.28%, which w a s c o n s i d e r e d S t u d i e s u s i n g t h i s model i n d i c a t e d t h a t removal times of a few

adequate.

years b u t s h o r t e r t h a n i n f i n i t y were not worthwhile, and t h e r e s u l t s of t h i s s e c t i o n assume R = 0.

3-1 e 4.3

T r a n s ~ l u t o n i u me f f e c t s

A d e t a i l e d s t u d y of t r a n s p l u t o n i u m effects9 was made, and t h e con-

c h s i o n was reached t h a t t h e r e s u l t i n g f i s s i l e p r o d u c t i o n only p a r t i a l l y o f f s e t s t h e capture.

The b a l a n c e i s l e s s f a v o r a b l e t h a n i n r e a c t o r s of

h i g h e r power d e n s i t y because of t h e partial ~ ~ e c aofy 2 4 4 ~ 1 to~ 2 4 0 which ~ ~ ~

has c o m p a r a t i v e l y less v a l u e .

me s t u d y showed t h a t e a c h atom of 2 4 0 ~ ~

produced from 2 3 % ~ i s j o i n e d b y 0.11 a d d i t i o n a l atoms from t h e decay of 244~me

FOP

each n e u t r ~ na b s o r p t i o n i n 2 4 2 ~ uc a ~ c u l a t e dw i t h o u t t h e trans-

plutonium e f f e c t , 4.0 a d d i t i o n a l a b s o s p t i o ~ ~and s 3 . 2 additional. f i s s i o n neutrons u l t i m a t e l y r e s u l t . Although t h e a c t u a l t i m e e f f e c t s a r e c o m p l i c a t e d , the net e f f e c t w a s a p p r o x i m a t e l y r e p r e s e n t e d a s a n a d d i t i o n a l f i c t i t i o u s ~ u c l i d e ,which w a s produced by c a p t u r e i n 2â&#x20AC;&#x2DC;â&#x20AC;&#x2122;~u and had t h e a b s o r p t i o n cross s e c t i o n o f 2 4 2 ~ u and no progeny.

This would be s l i g h t l y c o n s e r v a t i v e a t e q u i l i b r i u m and

probably a t e a r l i e r times a l s o .

3. E. 5

S t a t i c n e u t r o n i c results

T a b l e 9 i n d i c a t e s t h e i n v e n t o r y of a c t i n i d e s a t t h e b e g i n n i n g , mid-

d l e , and end of t h e 30-year o p e r a t i n g p e r i o d , assuming a 752 c a p a c i t y factor.

The h i g h i n i t i a l l o a d i n g of 235U i s l a r g e l y replaced by 2 3 3 U b r e d by

t h e system i n t h e f i r s t h a l f of t h e l i f e t i m e .

Toward t h e end of lifetime,

e n r i c h e d uranium a d d i t i o n s r e q u i r e d t o o v e r r i d e f i s s i o n - p r o d u c t b u i l d u p c a u s e a final i n c r e a s e i n b o t h the 23% and 238U c o n t e n t .

The plutonium

i n v e n t o r y i s n e v e r Barge because of i t s h i g h c r o s s s e c t i o n i n t h i s spec-

trum. T a b l e 10 shows t h e m i d l i f e n e u t r o n u t i l i z a t i o n i n f o m a t i o n .

low c a p t u r e r a t e i n nonfuel s a l t c o n s t i t u @ n t s (O.Ol53)

Y s t e the

and t h e f i s s i o n -

Table 9.

A c t i n i d e i n v e n t o r i e s i n DMSR f u e l s a l t

Inventory ( k g )

110,800 0 0 8

3,450 0 0 %4,000 0 0 0

Total actinides F i s s i l e uranium Total f i s s i l e

103,000 45 1,970 372 1,020 661 75

92 * 900 38 1,910

19,608

28 9 608 23 1

0 0

179 IO2 76 99 36

127 9 000 3,450 3 9 450

127 000 2 990 3 440

596 1 9

250 978

136

B 33 100 179 93

127 9 000 3,160 3 ,490

A t o t a l of 22.2% of t h e f i s s i o n t a k e

product c a p t u r e r a t e (0.0563).

p l a c e i n 2 3 8 ~and i t s progeny, even though they comprise o n l y 9.8% sf t h e fissile

~ I I V ~ I I ~ QI n~ ~s p . ite

of t h e h i g h v a l u e of v f o r these n u c l i d e s ,

t h e y would n o t be a s u f f i c i e n t f u e l without the thorium c h a i n .

The s l i g h t

c o n t r i b u t i o n of t h e t r a n s p l u t o n i u m n u c l i d e s to t o t a l mass h a s been i g n o r e d , and t h e a b s o r p t i o n v a l u e shown f o r t h i s n u s l i d e group i s a n e t of absorp-

tions less fissions. 2 4 1 ~ m ,a poor fuel.

A ~ O U4% ~

of the 2 b l ~ uis lost through decay to

me c a p t u r e i n 2 3 4 ~ a i s p a r t i c u l a r l y expensive be-

c a u s e e a c h such atom o t h e r w i s e would r e s u l t i n a h i g h l y p r o f i t a b l e 2 3 3 U

fission.

...z,y .. . c .A,

24 T a b l e 10.

Nuclide concentrations and n e u t r o n u t i l i z a t i o n after 1 5 years s f DNSR operation Concentration

Nus 1i d e

( x 1024)

Neutron absorption'

Fission fraction

vcrf/aa

Transplutoniram 8Pu

2 56% 1.13 49.0 9.21 25-1 16.2 1.83 474 4.34 2.46 1.84 2.38 0.882

Total actinfdes Fluorine Lithium Beryl Piurn

48,000 24,500

0,0017 0*0000 0.5480 0 e 0002

0,2561 0.0018 0 s 2483 8.0120 0,1161 0.0075 0.0047 0.0901 0.0896 0.0324 0.0293 0.0039 0.0014 0.0024

0,0001

0.0090 0.0033 2.2427 0,0143 l e 9894 0.0168 0.0182 0.0194 7905 0.0032 2. â&#x201A;Ź 7 5 4 0.0136

0.9803

0. E245

0.8956

0.2292

0.0001 0 s 0000 0.0017 0.31578 0.0001 0.0628

310

0000

0,0079 0 0062 0.0012 0

5 440

0.9109 Graphi t e Fission p r o d u c t s

92 9 270

0.0172 0 e 0563

0.9844

Total

Nuclei per c u b i c meter of s a l t o r moderator. h AbsorptTon per neutron horn; leakage is 0.0156.

3.1.5.2

Flux and power distributions and graphite lifetime

e he relative fast flux (E g i v e n i n Table 11.

52.4 k e ~ )and power-peaking factors are

These factors include t h e effects of flattening.

For

comparison, the o v e r a l l fast f l u x peaking in an unflattened core would be

-2.3;

the neutron leakage, however, wsbald be only 0.8% vs 1.56% for t h i s

core

25 ';x.w' ....

R e l a t i v e power d i s t r i b u t i o n s ( F i g . 4 ) show no s e r f o u s problems. peak o c c u r s i n t h e well-cooled

A power peak p e r u n i t

i n n e r zone.

Oâ&#x201A;Ź

The core

volume o c c u r s i n t h e gap between t h e c o r e and t h e r e f l e c t o r , b u t t h e power p e r unit volume of s a l t i s a c t u a l l y r e l a t i v e l y low i n that r e g i o n .

Table E l . Neutron flux and power-peaking f a c t o r s

Fast flux

Radial Axial Overall

Power

1.32 1.15

1.36 1.15 1.56

1.52

OHNL UWG80 4265 F T 3

AXIAL

RACIAL-

(50% SALT)

CORE B (12 9% SALT)

CORE A (20% SALT)

DISTANCE F R O M CENTER OF CORE (mi

Fig. 4 . DMSR r e l a t i v e power-density d i s t r i b u t i o n . p r o f i l e s a t e s e p a r a t e l y and a r b i t r a r i l y norwa%ized.

... .......... w:w

Axial and r a d i a l

26 The a b s o l u t e maximum f a s t f l u x i s of s p e c i a l i n t e r e s t because of its

effect in Limiting t h e g r a p h i t e l i f e t i m e and, t h u s , i n d e f i n i n g t h e c o r e The maximum danage f l u x c a l c u l a t e d i n t h i s s t u d y o c c u r s n e a r t h e

size.

edge of t h e i n n e r core and Fs, a t full powerrr

In 3 0 y e a r s a t 75% c a p a c i t y f a c t o r , t h i s l e a d s t o a f l u e n c e of 2.7 m-’

(2.7

loz2 ~ m - ~ )which , i s well below t h e nominal g r a p h i t e

damage limit of 3 x lo2‘ m-2 ( 3 x 10”

3.1.5.3

cm-21,

S p e c t r a l and cross-section e f f e c t s

A summary of r e l a t i v e a b s o r p t i o n s , f i s s i o n n e u t r o n p r o d u c t i o n s , and

n e u t r o n f l u x by energy group i s shown i n T a b l e 1 2 . are i n t h e resonance range, l a r g e l y i n thorium and

l a r g e r absorption f r a c t i o n i n t h a t group.

Many of the captures

’”u,

which l e a d s t o a

I n c o n t r a s t , most ~f t h e f i s -

s i o n s are caused by t h e r m a l ~ e u t r ~ n s .

N ~ U ~ ~ Q I IR e l a t i v e energy neut koa flux group

F r a c t i o n of neutron absorptions

F r a c t i o n of f i s s i o n neutrons produced

a A B B

Inner &Iter Inner Outer

2.44 2.51 2.42 3,14 2. I 4

2.548 2.032 2.540 1.022 5.8

7.86 7.96 7.96 18.6 6.6

3*22 3.17 3.29 3.38 3.08

of zone A , a s judged by t h e two zones with 1 = 25.4 m e To estimate the

e f f e c t on n e u t r o n y i e l d , t h e 2 3 2 ~ hc h a i n laas a n u l t i m a t e y m a i n a par-

ticular situation of 1.06 neutrons p e r capture in 2 3 2 ~ h ,while the yield of 23%

f s only 0.84

With -40% of t h e f e r t i l e c a p t u r e i n 2%!,

(Ref. 9).

as it is for the present system, a 10% increase tn the 238-to-232

ratis would reduce n e u t r o n yield by -8.5%. Table 12 is not a l a r g e e f f e c t . cross sections significantly

capture

Accordingly, t h e v a r i a t i o n i n

Even though cell geometry changes t h e

the 2 3 2 ~ hana

2 3 8 ~ changes approximately

cancel each o t h e r . h o t h e r v a r i a b l e of i n t e r e s t i s t h e d e n s i % y of the fuel-salt heavy nucIides.

Table 3 shows that 238U d e n s i t y apprsx%mately doubles d u r i n g

t h e l i f e of t h e system, and t h i s i s n o t accompanied by a corresponding change in 2 3 2 . ~ h . varying

Q V ~ Pa

range of r e a s o n a b l e i n t e r e s t , resonance

d a t a vs d e n s i t y a r e shown i n Table 14 f o r t h e case of c o r e zone A w i t h 25.4 mw.

While the 232Th d e n s i t y i n c r e a s e s by 51%, t h e product of den-

s i t y and cross section i n c r e a s e s by only 28%.

F o r 238U, the d e n s i t y in-

creases by 129%, and the product increases by only 57%, t h u s i l l u s t r a t i n g t h a t n u c l i d e d e n s i t y and i t s e f f e c t on resonance c r o s s s e c t i o n are both l a r g e , but p a r t i a l l y c a n e e l l i n g , e f f e c t s .

A similar t a b l e % ~ or t h e r nu-

c l i d e s f o r which resonances w e r e c a l c u l a t e d shows r e l a t i v e l y less influe n c e of n u c l i d e d e n s i t y on c r o s s s e c t i o n (Table 1 5 ) .

. .:. ..i ........

Table 14. E f f e c t of n u c l i d e density on key resonance cross s e c t i o n s

232Th

2288 2408 2582 2 800 3318

2.6% 2.52 2.44 2.37 2.21

0.5742 0.6048 0.6300 0.6636 8.7332

350

9.03 7.86

8.31160 0.3781 0.4452 0.4563 0.4976

238U

48% 658 722 800

47.2

6.85 6.32 6.22

33.3 32. a

15.8

39.9 45.6

25.9 40.8

27. 5 27.3

16.4 26*3

22.9 21.6

56,O 8.62

Spectral. e f f e c t s a r e a l s o i m p o r t a n t , because a more thermal spectrum improves the n e u t r o n y i e l d of b o t h 2 3 3 U and 2 3 5 U b u t a l s o results i n more p a r a s i t i c capture i n these f i s s i l e nuclides.

To i l l u s t r a t e t h P s , Table 16

shows the e f f e c t i v e n e u t ~ y~i ei l~d f o r t h e hard spectrum of core A vs the

s o f t spectrum of core B.

while u a f / o a within each ner~trsn group shows

r e l a t i v e l y l i t t l e change, t h e o v e r a l l r a t i o shows a 3% i n c r e a s e in y i e l d because of t h e s o f t e r spectrum.

Table 16. E f f e c t of n e u t r o n spectrum OR n e u t r o n y i e l d f o r homogenized c e l l material

A A

Resonance The rma 1

6.338

Overall Re s m a n c e

1.050 8.330 1e 432

B B B

Thermal Overall

1* 442

1e 880

a See F i g . 2 f o r i d e n t i f i c a t i o n of c o r e zones.

3.1.6

3.1.6-1

Burnup r e s u l t s

Reactor f u e l c y c l e

The t i m e h i s t o r y of t h e f u e l cycle i n t h e DMSR provides some i n s i g h t i n t o t h e uranium r e s o u r c e u t i l i z a t i o n i n t h i s concept.

The a v a i l a b l e re-

a c t i v i t y i n t h e c o r e (Fig. 5) shows an i n c r e a s e d u r i n g t h e f i r s t year as

the i n v e n t o r y Of

233u,

mQ%e e f f i c i e n t f u e l than

235u,b u i l d s .

'PkiS

r i s e would have to be controlled so that f u e l consumption was ~ i n h i ~ e d ,

Thus, a temporary removal of some denatured f u e l o r a d d i t i o n s of f e r t i l e

material might be more e f f e c t i v e than i n s e r t i o n of simple n e u t r o n p~lsons. After t h e f i r s t y e a r , t h e r e a c t i v i t y b e g i n s t o d e c l i n e as f i s s i o n - p r o d u c t

poisoning i n c r e a s e s ana overcomes t h e 2 3 4 ~e f f e c t .

R e a c t i v i t y i s subse-

q u e n t l y k e p t above 1.0 by p e r i o d i c a d d i t i o n s of makeup f u e l , c o n t a i n i n g 20% e n r i c h e d 2 3 5 ~ .

The n e t conversion r a t i o of the system ( f i s s i l e p r o d u c t i o n d i v i d e d by f i s s i l e consumption), which i s shown i n Fig. 6 , undergoes a much more p e r s i s t e n t rise t h a t l a s t s about f i v e y e a r s b e f o r e a g r a d u a l d e c l i n e sets i n t h a t Basts u n t i l t h e end of t h e %@=-yearcycle.

Much of t h i s decline

i s a t t r i b u t a b l e to n e u t r o n p ~ i s o n i n gby 2 3 8 U 9 which i s added w i t h t h e

30 ORNL-DWG 804266

ETD

FIRST FUEL

o= $5 c

< 2.5

4 4 5

0.0 0

OPERATING TIME (yri

Fig. 5. T i m e v a r i a t i o n of core r e a c t i v i t y i n a once-through BMSR operating at 75% c a p a c i t y factor.

OHNL - D V G 80-4267

FVD

OPERATING TIME ( v r )

Fig. 4 .

makeup f u e l .

Conversion r a t i o vs t i m e ,

The l i f e s t m e average c o n v e r s i o n r a t i o f o r t h e 30-year f u e l

cycle is ekose to 0.8. The s c h e d u l e of f u e l a d d i t i o n s , i n c l u d i n g t h e i n i t i a l c r i t i c a l load-

ing f o r a 1-GWe p l a n t o p e r a t i n g a t a 75% c a p a c i t y f a c t o r i s shown i n Table 17.

This t a b l e also i n c l u d e s the quantities of the U308 and separative

work r e q u i r e d to s u p p l y the f i s s i l e materialie ~ h u s ,the l i f e t i m e

re-

quirement would be about 2688 t o n s of U308 i f m c r e d i t were allowed f o r

t h e end-of-life

fissile i n ~ e n t o r y . However, uranium is r e a d i l y r&coV~B"-

able from t h i s f u e l i n a p u r e and r e u s a b l e f o m as LEge The recovered uranium would have t o be reenriched, e i t h e r by i s o t o p i c separation or by

a d d i t i o n of high-enrichment f u e l , before i t could be r e u s e d i n a n o t h e r

DMSR, b u t reuse i n some manner might be p r e f e r a b l e t o discarding t h e

Table 17.

ob E

Fuel addition schedule for once-thrsugh BMSR

14,800 0 174

105 890

4 5 6 7 8 9 10 11

3,450 0 0 0 203 0 203 0 203 263 0 2 03 203 2 03 203 2 03 0

0 822 6 82 2 822 6 822 82 2 822 82 2 822 0 822 822 822 82 2 82 2 822 822 822 822 822 822 82 2 822

13 14 15 16 17 1% 19 20 21 22 2% 24 25 26 29

28 29 Total

32,400

283 203 203 203 283 203

283 2 03 203 203 203

203 2 03

7 920 -~

~~~~

a 6 Eag

= 1.182

short tons.

bInitial l o a d i n g .

788 0 0 0

46.4 0 46*4 0 46.4 46.4 6 46.4 46.4 46.4 46.4 46.4 0 46.4 46.4 46.4 46.4 46.4 46.4 46.4 46. 4 46.4 46.4 46.4 46.4 46.4 1,810.6

789 0 0 0

46.4 0 46.4 0 46.4 46.4 0 46.4 46.4 46.4 46.4 46.4 0

46.4 46.4 46.4 46.4 46.4 46.4 46.4 46.4 46.4 46.4 46.4 46. Pa 46.4 1,810.0

"spent" f u e l .

If c r e d i t were allowed f o r t h e r e s i d u a l f i s s i l e uranium

i n t h e s a l t (plutonium presumably would n o t be r e c o v e r e d ) , t h e n e t U308 requirement would be reduced by almost one-half,

The temporal d i s t r i b u t i o n of f u e l r e q u i r e m e n t s i n a DMSR i s a l s o significant.

The d a t a i n Table 1 7 show t h a t o n l y about 36% of t h e makeup

f u e l i s r e q u i r e d d u r i n g t h e f i r s t 1 5 y e a r s of t h e c y c l e ; t h e major demand o c c u r s toward t h e end-of-life.

Thus, i f t h e r e a c t o r were o p e r a t e d a t a

lower c a p a c i t v f a c t o r i n l a t e r y e a r s ,

t h e U308 requirement could be re-

duced f u r t h e r o r t h e p l a n t c a l e n d a r l i f e t i m e could be extended.

The ad-

v a n t a g e a s s o c i a t e d w i t h t h e time d i s t r i b u t i o n of t h e makeup f u e l r e q u i r e ment i s p a r t l y o f f s e t by the l a r g e i n i t i a l f u e l l o a d i n g and t h e h i g h i n plant f i s s i l e inventory.

T h e r e f o r e , a n optimum f u e l c y c l e might coneeiv-

a b l y b a l a n c e a Power i n i t i a l l o a d i n g (and i n v e n t o r y ) having a lower n e t c o n v e r s i o n r a t i o a g a i n s t a h i g h e r requirement f o r makeup f u e l .

There ap-

p e a r s t o be some l a t i t u d e f o r o p t i m i z a t i o n o f t h e f u e l c y c l e i n t h i s area.

3.1.6.2

P o t e n t i a l f o r imDrsvement

mile

t h e f u e l u t i l i z a t i o n of t h i s c o n c e p t u a l system compares f a v o r -

a b l y with t h a t of o t h e r r e a c t o r systems, some f u r t h e r improvements may be possible.

Only a l i m i t e d range of f u e l volume f r a c t i o n s and c o r e zone

s i z e s h a s been c o n s i d e ~ e d lf o r t h i s c o r e , and o t h e r v a l u e s could l e a d to

higher performance.

However, t h e r e a p p e a r s t o be l i t t l e p o t e n t i a l b e n e f i t

i n u s i n g more t h a n two core zones. The a c t i n i d e c o n t e n t of t h e s a l t i s thought t o be near optimum f o r long-term, high-performance c o n v e r s i o n , b u t , as implied p r e v i o u s l y , ano t h e r c o n c e n t r a t i o n might be b e t t e r f o r t h e 30-year c y c l e .

Certainly,

some improvement i n f u e l u t i l i z a t i o n would come from r e l a x i n g t h e requirement f o r 2 3 8 ~c o n t e n t e i t h e r of t h e system i n o p e r a t i o n o r of t h e makeup material being added.

Table 18 shows t h e approximate e f f e c t of

t h e s e c o n s t r a i n t s on 2 3 5 ~requirements.

Removing t h e requirement t o

f u l l y d e n a t u r e t h e makeup f e e d material h a s o n l y a small e f f e c t on t h e f u e l requirement. dk

S i m i l a r l y , i n c r e a s i n g t h e allowed enrichment of 234U

T h i s i s f r e q u e n t l y done i n e l e c t r i c power stations a s newer and cheaper p l a n t s are b u i l t .

Table 18. Effects of denaturing on 30-year cycle performance

0.1 0.2 0.4

0.2

0.142828 0.142828 0.142828

3450 3450 3450

5120 4470 4260

$ma

0.2 0.2

0.2

0.071414 0.142828 0.285656

3450 3450 3450

4470

0.2 0.2

0.2 0.2 8.2

4470 4470

8120a 7920 7920

8.2 0.2 0.2

0.2 0.4

0.142828 0.142828 0.142828

3980 3450 3040

4470 4670

7710

0.2

2800

4040

6860

2800

5600

%enotes

5120

9928

7710

9100a 7920

reference conditions.

in the coke has little effect.

Only complete removal of all enrichment

constraints would achieve an important fuel saving of 29%.

Thus, the re-

quirement for d e n a t u r i ~ g cannot be regarded as an overriding limitation on the potential performance of this f u e l cycle.

3.1.7

Dynamic effects

The dynamics of the DMSR would be dominated by the following factors:

a slow, positive moderator-temperature coefficient ~f reactivity;

a negative fuel-salt density coefficient of reactivity;

an interaction between fuel-salt flow rate and the neutronic response

prompt, negative fuel-temperature coefficient of reactivity;

of the core caused by the sweeping of delayed neutron emitters out of the core;

a long f u e l - s a l t residence time in the core (relative to the average

neutron lifetime).

.. .....a .... # '

34 3.1.7.1

Material r e a c t i v i t y w s r t h s

The r e a c t i v i t y worths sf t h e major f u e l c o ~ s t i t u e n t sare shown in Table 19.

The t o t a l worth is n e g a t i v e because t h e e f f e c t s of f e r t i l e

thorium and 2 3 8 ~overcone t h e p o s i t i v e e f f e c t s sf t h e f i s s i l e materials. T h i s means t h a t t h e r e a c t i v i t y c o u l d be made s i g n i f i c a n t l y h i g h e r by removing f u e l s a l t , a l t h o u g h t h e b r e e d i n g performance would be reduced.

Table 19.

Material concentration coeffielent of reactivity

0.00026 -0*08691

-0.14 0.15 0. 010

-9- 00011 -0.0023

4.29 -0.0626

0.0048

0.20

-0.8015 0.0025 -0 e 00050 -0.00019

4.014 4-063 4.0079 4.092

Removing lX of t h e uranium would have a r e a c t i v i t y e f f e c t of -0.0015

~ k / k ,and r e i n s e r t i n g i t would have a comparable p o s i t i v e e f f e c t . p a r a b l e r e s u l t f o r plutonium would be o n l y 8.0001 Akbk.

A com-

If 1%of t h e

f u e l s a l t could be r e p l a c e d s u d d e n l y by bubbles, t h e e f f e c t would be an i n c r e a s e of 0.00114 i n r e a c t i v i t y , which i s s u f f i c i e n t t o induce a signific a n t system transient.

I n p r a c t i c e , no l i k e l y mechanism e x i s t s t h a t could

b r i n g about such a n e f f e c t suddenly. ~ ~ a s pee c i f i c c o e f f i c i e n t s S ~ Q Wt h a t , atom f o r atom, 2 3 3 ~i s a much

more r e a c t i v e f u e l t h a n 2 3 5 ~o r plutonium i n t h e r e f e r e n c e i s o t o p i c mix

and t h a t 2 3 % i s a g r e a t e r d e p r e s s a n t t h a n thorium.

3.1.7.2

Temperature effects on r e a c t i v i t y

Temperature a f f e c t s t h e r e a c t i v i t y of t h e core by (1) broadening narrow ~ r o s ~ - s e ~ t iroe sno n a n c e s , t h u s i n c r e a s i n g t h e i r c a p t u r e rate

35 (Doppler e f f e c t ) , (2) changing t h e energy d i s t r i b u t i o n of the thermal n e u t r o n spectrum, and ( 3 ) causing expansion of t h e c o n s t i t u e n t materials.

The expansfon changes b o t h t h e s i z e and density of t h e c o r e , as discussed

earlier.

Table 28 shows t h e v a r i o u s c m p o n e n t s of t h e t o t a l temperature

coefficient.

The f u e l c o e f f i c i e n t i s dominated by both a l a r g e , n e g a t i v e

Doppler component and a similar s p e c t r a l component. p l e , t h a t an i n c r e a s e of " I s C

This means, f o r exam-

i n f u e l temperature would reduce r e a c t i v -

ity by 6.009 e s s e n t i a l l y i n s t a n t a n e o u s l y .

Table 20.

Temperature c o e f f i c i e n t s of r e a c t i v i t y f o r DMSR

F u e l - s a l t Doppler Fuel-salt d e n s i t y F u e l - s a l t thermal spectrum Total f u e l salt

Moderator d e n s i t y Moderator expansion Moderator thermal spectrum Total moderator T o t a l core Reflector density R e f l e c t o r thermal spectrum R e f l e c t o r and v e s s e l expansion Total r e f l e c t o r Total reactor

57 30 4 0

4 7 -2.2 7.2 14

4 8 0.1 1.2 -4.9 -%. 8

-7 2

"he moderator e f f e c t i s dominated by p o s i t i v e s p e c t r a l and expansion

effects.

T h i s e f f e c t i s r e l a t i v e l y slow t o appearo however, because t h e

t i m e c o n s t a n t f o r conduction h e a t i n g of t h e g r a p h i t e f s on t h e o r d e r of 146 s.

If t h e temperature change were caused by a r a p i d power i n c r e a s e ,

a small p o r t i o n (-5%) of t h e excess power would appear i n the m ~ d e ~ a t o r

36 immediately because of d e p o s i t i o n of energy by fast n e u t r o n s and prompt

gammas,

Because the h e a t c a p a c i t y of t h e moderator i n a zone i s always

a t l e a s t five times that of the f u e l , t h e e f f e c t of d i r e c t t r a n s i e n t mad-

erator heating would be n e g l i g i b l e . The r e f l e c t o r and vessel c o e f f i c i e n t s would probably b e v e r y slow i n t a k i n g effect because of l a r g e h e a t c a p a c i t i e s and Pow fuel flow rates. T h e i r t o t a l i s dominated by a n e g a t i v e expansion tern.

3 1e 7 3 0

Belayed-neutron effects

T a b l e 21.

Contributor

Delayed-neutron fraction, 6

Fissiom fraction

2 q J 2 3 5 ~

0,55

B 1u e. oniurn

0.22

0.23

Total B

Contribution t o B 0.0014 Q.OOP5 0 * 00046 0.0034

An unusual aspect of MSRs 1 s t h a t the fuel c i r c u l a t e s f a s t enough t o remove s i g n i f i c a n t numbers of delayed-neutron p r e c u r s o r s from t h e core be-

f o r e t h e neutrons are e m i t t e d .

The Lmped-parmeter kinetics equations

a r e taken a s

and

,...".IW ........ ..,

where

= relative

delayed-neutron precursor concentration,

t = time, B = reactor power,

delayed-neutron fraction,

muPtipLieation factor,

h = prompt-neutron generation time,

aeiayea-neutPon p r e c ~ r s o rdecay constant, delayed-neutron fractional yield,

R = coolant flow constant, T =

mean salt transit time in external loop.

These equations then show that, where dollars of reactivity are defined as $i= ( k

- l)/kB, t h e steady balance c o n d i t i o n requires a nonzero value

Thus,

where E5 is equal to 1 - e

'Ai'

Defining a new effective reactivity as

we can write the inhour equation relating asymptotic inverse period w t o A , the amount of reactivity in excess of that required t o maintain steady

state under the given flow rate.

where Fi is equal t o 1 - e~p[--CAi 9

U)T].

The prompt-neutron generation

t i m e was calculated by the boron-poison method to be 362 ps.

For now,

we w i l l ignore t h e difference between B and Beff, which i s expected t o be small fn Bow-leakage systems.

Table 22 was compiled u s i n g standard

Table 22.

FIQW constant,

Kinetic response of BMSR

Flow reactivity %oss, $ - A

reactivity, A

IieactQr period, I / @

( d o l l a r s1

%SI

Net

(dollbar S)

(8-9

0 0 0 6 0 0 6 0 0.0515 0.0515 0.0515 6.8515 0.051 5 0.0515 0.851 5 0,0515

6 8.11 0.26 6.45 8*69 Q. 92 1.28 2,04

0 0

0 0 0 0

delayed-neutron data.

’’

0 0.23 0.23 0. 23 0.23 0.23

0 0.87 0.16 0.29 0*40

8*23 0.23 0.23

0.78 1.05 1.82

re flow

100

30 110

8.3 0.

31 80 20 10 3 1 0.3 0.6

c ~ ~ l s t a n t 0.0515

s-l corresponds t o

residence time ~f 19.4 s in t h e full-power ope ratio^ with a r r ~ e afuel ~ CQ%@

‘ ~ h e s edata sthow that less excess ~ e a ~ t i v i tis y r e q u i r e d ~ O Pa g i v e n

small power response when the s a l t is flowing.

becomes constant at higher reactivftfes.

T h i s reactivity difference

Because of the v e r y PSI^ genera-

t i o n time, t h e response to net reactivity changes of more than l dollar

would be much smaller than that of many reactor types, which is characteristic of over-moderated graphite assemblies.

The overall result would be

a system with a power l e v e l that fluctuates more than u s u a l because of in-

herent operating noise b u t that would be relatively easy to shut down by CoIltPOl rad aC%iOn

an Unplanlled

eVeIlt.

Power f%uCtUatiQRs Would ’be ex-

pected t o have l i t t l e effect on the e x t e r ~ a lsystem because of t h e large heat Capacity

the Cor@ atad t h e low flow %pate.

39 .... ...., , c.y..p

3.1.7.4

Control requirements f o r normal operation

Assuming a core preheated to 775 K

and near critical, a reactivity

increase of 0.07% must be supplfed by t h e contrp~lsystem

fuel flow

through the core is started t o compensate f o r the l o s s sf delayed neu-

trons.

An increase of about 1.1% must be supplied to bring t h e fuel9

IIIOderatQr, and Vessel t o t h e average OpePatiIlg tempE?ratU%e

925 R e

Xenon concentration is kept to a negligible level by the salt cleanup

system and, therefore, has little effect on the control requirements. A Illore sE?rfoUS %eqUiPement is the lOIlgE?le-telXl pOS%tiV@ reactivity peaking

caused by e a r l y production of 233U; this 1 s approximately a 3% effect.

'En addition, some shutdown margin (perhaps. 2 % ) would be required for sELfe%y, T h i s Would be SuffiefeIlt to

OVE?PCQITIe i 3

12% cha.l%gei n salt dell-

sityr for example, o r a comparable l o s s in actinide content of the salt. The t o t a l reactivity control §pan r e q u i r e d W i t h respect t o a 7 7 5 R , T - I O - ~ ~ Q W , just-critical

core thus would be from +%,2% t o -5.O%,

Of this,

only about -2% must be rapid in nature, inserted by an a c t i v e contr~l device.

The remainder csu%d be partially supplied by adjusting t h e com-

position of t h e fuel s a l t .

3.1.7.5

Stability and transient safety

The

i s stable to a l l frequen€!ieS Of Q§cilbatiOIl because t h e

negative prompt

of the temperature coefficient of reactivity

CQITI~QH~~I-I~

dominates the positive delayed component.

A t frequencies below that as-

sociated with the graphite thermal conduction process, the delayed component could subtract from the prompt component, but the effective coef-

ficient wsuld be no less negative than the t o t a l core coefficient.

These

frequencies would be on the order of inverse minutes and should pose no problem f o r c o n t r o l . The response to sudden changes in the fuel-salt inlet temperature is relatively

S ~ O Wdue

t o a s a l t residence time of almost 20 s .

For example,

the reactivity response t o an abrupt change in inlet fuel-salt temperature

...

"<....srJ

"This would be an absolute minimum temperature because the salt ~ begin l t od freeze at lower temperatures.

40 of-50"C

would be +0,004 ( a b o u t 1 . 9 d ~ l l a r s )i f t h e e n t i r e core c o u l d be

filled.

Wowever, t h e e x t e r n a l l o o p s contain enough s a l t t o fill o n l y

about o n e - t h i r d of t h e c o r e ; t h u s , t h e actual r e a c t i v i t y e f f e c t would be

much smaller, and i t would be i n s e r t e d o v e r a p e r i o d of s e v e r a l seconds. The control system could r e a d i l y compensate f o r such a activity.

S ~ S Wchange

i n re-

With no c o n t r o l r e s p o n s e , a new power l e v e l would be g r a d u a l l y

approached to c o u n t e r a c t t h e cooling e f f e c t of t h e i n l e t c o n d i t i o n , and t h e n t h e c o n t i n u e d h e a t i n g e f f e c t of t h e h i g h e r power level would c a u s e a r e d u c t t o n i n r e a c t i v i t y and a r e t u r n t o a s t a b l e condition.

3.2

R e a c t o r Thermal H y d r a u l i c s

The purpose of t h e t h e r m a l - h y d r a u l i c a n a l y s i s o f t h e DMSR i s t o demo n s t r a t e t h a t t h e concept i s v i a b l e and n o t t o p r o v i d e a d e t a i l e d d e s i g n . N e i t h e r t h e funding n5r t h e n e c e s s a r y t h e r m a l - h y d r a u l i c p r o p e r t i e s o f m o l t e n s a l t f l o w i n a g r a p h i t e c o r e a r e p r e s e n t l y a v a i l a b l e t o perform t h e latter.

C o n s e r v a t i v e estimates of i m p o r t a n t p a r a m e t e r s are t a k e n wherever

p o s s i b l e ; even i f some of t h e s e should be n o n c o n s e r v a t i v e , simple modifications Qâ&#x201A;Ź the core design apparently could lead t o acceptable r e s u l t s . Thermal-hydraulic

b e h a v i o r d o e s n o t a p p e a r t o be a l i m i t i n g d e s i g n con-

s t r a i n t on t h e DMSR r e f e r e n c e c o r e . Because of t h e r e l a t i v e l y low power d e n s i t y of t h i s concepts s i m p l e

c o r e c o n f i g u r a t i o n s t h a t were n o t p o s s i b l e i n t h e MSWR r e f e r e n c e d e s i g n 8 may b e ~ ~ n s i d e r e d ,Three s i m p l e d e s i g n s were c o n s i d e r e d : 1.

a c o r e made up of s p a c e d g r a p h i t e s l a b s ,

a c o r e made up of s t a c k e d hexagonal g r a p h i t e b l o c k s w i t h c i r c u l a r

coolant channels,

a c o r e c o n s i s t i n g of a t r i a n g u l a r a r r a y of g r a p h i t e c y l i n d e r s w i t h c e n t r a l coolant channels. C o n s t r a i n t s t h a t must be c o n s i d e r e d i n s e l e c t i n g a c o r e d e s i g n i n -

e l u d e maximum g r a p h i t e element t e m p e r a t u r e , l o c a l s a l t volume f r a c t i o n , and t h e d e s i r e d 2 3 8 ~s e l f - s h i e l d i n g e f f e c t , which imposes a m i n i m u m

l i m i t a t i o n on the c o o l a n t c h a n n e l dimensions.

The t e m p e r a t u r e rise be-

tween t h e coolant ~ f n a n n e land t h e h o t s p o t i n t h e g r a p h i t e ~ ~ ~ ~ d e releator

m e m t i s e s p e c i a l l y important because of t h e s t r o n g dependence of g r a p h i t e

dimensional change on temperature.

he salt volume fraction ana the

self-shielding effect strongly couple the thermal-hydraulic and the

2 3 8 ~

R~U-

These combined constraints appear to rule o u t the

tronic core designs.

possibility oâ&#x201A;Ź a graphite slab core configuration. Nechanical problems, especially the loss of coolant channel geometry caused by shifting of stacked hexagonal blocks (thus creating stagnant or low flow zones), rude o u t the second o p t i o n .

In addition, that o p t i o n w o d d leave an un-

desirably large fraction of the fuel salt hexagonal elements.

irt

narrow passages between the

The third design seems to f i l l all the requirements

and is a l s o very appealing because of its structural simplicity, which i s important in a core expected to last the life of the plant. The outer diameter of the cylindrical graphite elements i~ the reference BMSR is 254 mm (40 i n . ) , which is machined down to 244 m ( 9 - 1 6 i n e ) in the central region (core zone A , P i g . 2 ) .

ner coolant channels i s c o n s t a n t at 51 mm (2 i n . ) .

The diameter s f t h e in-

This design provides

a salt fraction sf 20.0% in the central region and 12.9% in the remainder of the core. a

The motivation behind this two-region design is t o provide

â&#x201A;Źirst estimate of a flux-flattened core.

Flux flattening is crucial t o

the d e s i g n objective of reactor-lifetime graphite because both the. waximum graphfte damage and the maximum graphite temperature are reduced. Figure 7 shows the arrangement of the graphite moderator elements in the outer region, zone B, which occupies most of the core volume, the 51-m-dim

(2-in.)

Note

interior salt channels and the e x t e r i o r salt chan-

nels formed between the moderator elements, In zone B , the exterior chann e l s have a uniform cross section along their entire length, except for p o s s i b l e orificing provisions at the ends.

The arrangement i s the same

in the interior s f the core (zone A ) , but the outer diameter of the moderator elements in the interior is reduced.

This provides the higher salt

fraction and allows the exterior channels to interconnect in that region. Figure 7 also shows t h e l o e a t i o n of a 30' segment of a graphite element used in the analysis.

The film heat-transfer coefficient at the

graphite-salt interface is not well known, primarily because a helium f i l m may exist on the graphite surface, This film would increase the thermal resistance b u t also would g i v e a no-drag wall boundary condition to the salt velocity profile.

....:...w ..., ..,: %.

%at

addition, near the moderator element contact

WJDEL SECTION FOR T! i E R M A L_ AN A L.Y SIS

Fig. 7.

Arrangement of moderator elements in c o r e B,

p o i n t s (which o n l y e x i s t in t h e o u t e r c o r e r e g i o n ) , h e a t transfer would be g r e a t l y diminished.

The foPPowfng s i m p l i f y i n g assumption w a s macle

(probably c o n s e r v a t i v e ) concerning t h e s a l t - f i l m h e a t - t r a n s f e r c o e f f i cient:

within

of a c o n t a c t p o i n t , t h e h e a t - t r a n s f e r c o e f f i c i e n t i s

z e r o ; elsewhere, i t i s e q u a l t o 80% of t h e v a l u e o b t a i n e d by use. of the

~ i t t u s - ~ s e ~ t ceorr r e l a t i o n . 18

TMS assurnption i n c r e a s e d the c a l c u l a t e d

t o t a l temperature r i s e i n t h e moderator element by -50% over t h a t obt a i n e d by u s i n g 8OX of t h e D i t t u s - B ~ e l t e r v a l u e f o r t h e e n t i r e s u r f a c e , With t h e s e boundary c o n d i t i o n s , the h e a t c o n d l ~ e t i ~enq u a t f o n i n cyl i n d r i c a l finite d i f f e r e n c e f o m was solved i n t h e 30" s e c t i o n a t each a x i a l node ( 4 0 nodes t o t a l ) using t h e method of s u c c e s s i v e ~ v e ~ ~ e % a ~ a t i o n .

Starting a t t h e core i n l e t , t h e temperatures of an i n t e r i o r and an e x t e r i o r s a l t channel are advanced i n an axial marching-type

s o l u t i o n through t h e

coreo M ~ P and r a a i a i k0mogeneoras power p r o f i l e s gsee Fig. 4 )

used,

along w i t h t h e following assumptions, t o g i v e s a l t - c h a n n e l axial l i n e a r power p r o f i l e s (Fig. 8) and

IbOcal

UIode~atOH VOlUElet9Pic power.

A t a given

l o c a t i o n , t h e vsPumetrie powers i n an interior and an e x t e r i o r channel

were assumed t o be t h e same, and t h e volumetric power w i t h i n t h e g r a p h i t e

was assumed t o be 1% of t h a t in the s a l t and c o n s t a n t over t h e 30" section,

~ e u t r o n i canalyses of t h e previous MSBR design8 provided the basis

Fig. 8.

4 5 A X I A L POSITION (mi

Core hot-channel linear power profiles.

for the latter a ~ s u ~ ~ p t i othe n ; present analysis is n o t detailed enough t o yield a better estimate. A s n ~ t g dpreviously, t h e hot-ckannd axial linear power profiles of

an interior and exterior channel are shown in Fig. 8.

The h o t channel oc-

curs (radially) at t h e boundary between zones A and B.

The central loca-

t i o n of zone A (2.65 t~ 5.65 an) can be seen by the discontinuities in the

linear power curves.

The

curves reflect b o t h heat generated within the

salt and heat transferred from t h e graphite to the salt. Figure 9

S ~ Q W Scalculated

temperatures in the interior and exterior

channels and the maximum temperature in t h e moderator as functions of axial position for the core hot channel.

The highest moderator temperature

occurs at a position 3.0 rn above the core midplane.

Isotherms in the 30"

moderator segment at this l ~ c a t i ~are n shown in Pig. 10.

The assumption

of RQ heat transfer within 15" ~f the contact point causes substantial

distortion of the isotherms near the o u t e r surface.

The calculated maxi-

mum graphite temperature is 741째C ( 1 3 6 6 " F ) , which i s close t o the maximum

allowable temperature (-720째C.) total fluence of 3

.... ,.::.y ..,, ~.

x 1026 m-2-

for zero positive irradiation growth at a

3. EXTERIOR SALT CHANNEL TEMPERATURE

Fig. 9.

4 5 AXIAL POSITION (rn)

Axial temperature p r o f i l e s :

graphite and f u e l salt for core

h o t channel.

Table 23 g i v e s f l o w areas, salt velocities, Reynolds numbers, and heat-transfer c~efficientsf o r t h e interior and exterior channels ( i n b o t h core zones A and R ) for two cases:

core average channel.

equal 139'C

The salt velocities are those necessary f o r an

(25Q째F) temperature r i s e

exterior channels.

t h e core h o t channel and the

~ C P Q S S the

core in the interior and

The hydraulic diameters are not e q u a l ; t h u s , orificing

of the interior channels would be necessary.

This could be accomplished

easily by reducing the diameter of the interior channel by -50% f o r a short interval near core inlet a n d / o r outlet, Overall core orificing would a l s o be neeessary t o equalize core exit temperatures.

tional p r e s s u r e drop across the core

[-a

The fric-

kPa (1 p s i ) ] is insignificant

when compared with the pressure d r o p across the primary heat exchanger

45 O R k L -DWG 80-4270

ETD

EXTE R IO w

CHANNEL TSALT =

Fig. 10. Isotherms in graphite at location of maximum temperature Film heat-transfer coefficient h equals 0 where shown and equals ("C). 80% of Dittus-BoePter correlation value elsewhere.

Table 23.

Thermal-hydraulic d a t a f o r DMSR c o r e Channel Hot

FLOW

area,

Average

I n t e r i o r channel E x t e r f o r core A Exterior core R

2.025 4.580 2.601

x x

10-3 10-3 10-3

2.025 4.580 2.601

10-3

x x

10-3 10-3

S a l t v e l o c i t y , m/s I n t e r i o r channel Exterior core A Exterior core B

0.601 0.418 0.735

0.441 0.307 0.540

1.034 x i o 5 6.959 x l o 4 6.490 x l o 4

7.592 4.962 4.765

173 I30 232

7% 59 145

Reynolds number I n t e r i o r channel Exterior core A Exterior core R Heat-transfer c o e f f i c i e n t , W rnW2 K-l I n t e r i o r channel Exterior core A Exterior core R

x 104 x

io4

lo4

~~~

Maximmdaverage power i s 1 . 3 6 2 over c u r e s A a n d B.

'Obtained value.

,.. .%.....+.P...

..,.,...,.

by using 8 0 2 of Dittus-Boelter c o r r e l a t i o n

[-SO0

kPa (130 p s i ) ] .

Graphite and f u e l - s a l t p r o p e r t i e s used i n t h e

a n a l y s i s were o b t a i n e d from Ref. 8. The r e f e r e n c e DMSR

d e s i g n s a t i s f i e s t h e two most i m p o r t a n t

thermal-hydraulic c o n s i d e r a t i o n s :

(a>

t h e maximum g r a p h i t e t e m p e r a t u r e

i s ISW enough t o a l l o w i t t o last t h e l i f e of t h e p l a n t (24-full power y e a r s ) and ( 2 ) r e g i o n s sf s t a g n a n t o r l a m i n a r flow are avoided.

Variations

t h i s design

W i l l

be possible i n a c h i e v i n g

but t h e second c o n s i d e r a t i o n must a l w a y s be noted.

iipl

Halay

Optimum c o r e ,

Because of the EQW

thermal c o n d u c t i v i t y of t h e f u e l s a l t , e x c e s s i v e t e m p e r a t u r e s can o c c u r i n very small s t a g n a n t o r l a m i n a r flow r e g i o n s .

The g r a p h i t e elements

must r e t a i n t h e i r g e o m e t r i c i n t e g r i t y and must n o t c r e a t e f l o w blockages. E x t e n s i v e in-pile t e s t i n g would be n e c e s s a r y t o e n s u r e t h a t b o t h of t h e s e cOnsideratiOnS Were met b e f o r e Const?XC%iQnSf

dâ&#x201A;ŹSXlonStratiOn p l a n t could

be undertaken.

3.3

F u e l Behavior

E x c e l l e n t n e u t r o n economy i s a n a b s o l u t e requirement f o r a thermal b r e e d e r ( s u c h as a n MSBR), and fuel c o ~ l p o n e n t sw i t h a c c e p t a b l y l o w n e u t r o n CFQSS

sections a r e few.

We recognized v e r y e a r l y i n t h e MSBR development

e f f o r t t h a t ( 1 ) only f l u o r i d e s need t o be c o n s i d e r e d , ( 2 ) o n l y L i F and BeP2 would prove a c c e p t a b l e a s f u e l s o l v e n t s ( d i l u e n t s ) f o r t h e f i s s i l e

and f e r t i l e f l u o r i d e s , and ( 3 ) t h e L i F must be h i g h l y e n r i c h e d in 7Li, For a break-even r e a c t o r o r f o r one t h a t , though i t r e t a i n s most of t h e

fission p r ~ d u c t sw i t h i n t h e f u e l , i s t o be a n e f f e c t i v e c o n v e r t e r , some s a c r i f i c e i n meutron economy may be p e r m i s s i b l e .

However, no l i k e l i h o o d

f o r s u c c e s s seems p o s s i b l e w i t h d i l u e n t s o t h e r than BeF2 and L i F h i g h l y

e n r i c h e d ( p r o b a b l y t o 99.99%) i n 'Li, A c c o r d i ~ g l y , t h e f u e l system f o r a DMSR n e c e s s a r i l y w i l l be v e r y

s i m i l a r t o t h e system t h a t r e c e i v e d i n t e n s i v e s t u d y f o r many y e a r s i n

MSBR development.

A c o n s i d e r a b l e fund of i n f o r m a t i o n e x i s t s about ckem-

i c a l p r o p e r t i e s , p h y s i c a l p r o p e r t i e s , and expected i n - r e a c t o r b e h a v i o r of suck materials.

47 ........ ....... . .%..W

3.3-1 3.3.1.1

Basic considerations

Composition of DMSR fuel

Choice of initial composition.

A DMSR will derive some of i t s fis-

sion energy from plutonium isotopes, but

2 3 3 ~and '3%

will be the p r i m a r y

f i s s i l e isotopes, while '32~h, with important assistance from 2 4 8 ~ , is the

fertile material

Clearly ( s e e previous neutronics discussion) the 2 3 2 ~ h

concentration will need t o be markedly higher t h a n the total concentration of uranium isotopes.

The only stable fluoride of thorium is ThF4; thus, it must be used i n such fuels.

Puke UF3 is appreciably disproportionated at high tempera-

tures by the reaction

Generally in molten fluoride s~lutions,this reaction proceeds appreciably at lower temperatures. A small amount of UF3 will be formed within the f u e l by reaction (reduction) of UP4 with species within the container

metal and,

explained below, a small quantity of UF3 deliberately main-

tained in the fuel serves as a very useful reduction-oxidation (redox) buffer in the fuel.

a large excess of UP,,

Such UF3 is sufficiently stable in the presence s f but UF4 must be the major uranium species in the

fuel. 19320

~onversely,P U F ~is reduced by the metallic container (and

a l s o by LE,),

and PuF3 is the stable fluoride of this element in DMSR

fuels. Phase equilibria among the pertinent fluorides have been defined in detail and are well documented.

19-21

Because the concentration of ThF4

is much higher than that of UF4, the phase behavior of the fuel is dic-

tated by that of the LiP-BeF2-ThF4

system shown in Pig. 11.

The compound

3LIP*ThF4 can incorporate Be2* ions i n both interstitial and substitutional sites t o form solid solutions whose compositional extremes a r e represented by the shaded triangular region near that compound.

The maximum

ThF4 concentration available with the l i q u i d u s temperature below 500째C is

just above 14 mole %.Replacement of a moderate amount of ThP4 by UF4

scarcely changes the phase behavior.

48 QRML-LR-BWG

37420AR9

ThG 4 1 7 4

TEMPERQTIIRE IN OC COMPOSITION IN mole To

526 IF Bei=<'500/450

P 458

Fig. 11.

400

400 450

500

BeF, 5 55

E 360

System EiF-BeP2-ThF4.

The MSBR p r o p ~ ~ et od use a n initial f u e l mixture c o n t a i n i n g 71.7

mole % E f F , 16 mole X BeP2, 12 mole 2 ThF4, and 8 . 3 mole % ( h i g h l y en-

riched) UFq.

The o p t i m a l initial concentration of fuel f o r a DMSR is not

yet p r e c i s e l y defined.

The f n i t i a l f u e l l i k e l y will need t o c ~ r t % a i ben

tween 9 . 5 and 12.5 nobe % of heavy metal (uranium plus thori~m), w i t h

uraIliW~I (enriched to 20% a3581> CQrreSpOnding to about 12% O f the t o t a l . The composition range of interest to BEufSRs, therefore, is likely to be bounded by ( a l l c o ~ ~ ~ ~ t ~ ~i nt mole i o nXI: s ThF4; 1.15

78.8 LFF; 19.7 BeF2; 8.35

UF4 and 71.5 L i F ; 16 BeF2; 11 ThF4; and 1.5 UF4.

For such

~ o m p ~ s i t i o n s the , l i q u i d u s would range from about 480 t o 5 0 0 ~~ o~ s ~t shernical and physical properties of the chosen composition can be i n f e r r e d r e a s o n a b l y well from existing data f o r the MSBR reference composition.

Variation of fuel composition with time.

Fission s f 235U in the

operating reactor will result in a decrease of that isotope and an ingrowth of fission products and in the generation of ( 1 ) 234U from 232Th, ( 2 ) 2 3 9 ~ ufrom 2 3 8 ~ ,and ( 3 ) numerous transuranium isotopes.

~ u r t h e r ,a

once-through DMSR will require additions of uranium at intervals during its lifetime.

F o r example, a DMSR with a fuel containing 9.5 mole X heavy

metal ~ ~ u contain l d about 110,000 kg of 232Thf 3,456 leg of 2 3 5 U 9 and 1 4 , 0 0 0 kg of 2 3 8 ~at startup.

~uring30 years of operation at 75% plant

factor, it would require the addition of 4 , 4 7 0 leg of 235Y and bEi9400 kg Of

238u

I f Such a r e a c t o r r e c e i v e d only additions

UF4 and UF3 and i f

no f u e l were removed,* t h e final quantities and concentrations of heavy

metals in the f u e l would be those indicated in T a b l e 24.

The end-of-

life f u e l would a l s o contain about 1.4 mole % of soluble fission-product

species and would have a t o t a l of nearly 2.4 mole Z of uranium isotopes, about 6.053 mole % of plutonium isotopes (about 32% of which is 2 3 9 P u ) 9 and less ThFk than the original fuel.

Thus, the concentration of heavy

metal in the f u e l changes very little although the species do change; total heavy metals in the end-of-life f u e l equal about 9.3 mole X compared with an initial 9 - 5 mole X.

Therefore, the physical properties of the

f u e l would not be likely to change appreciably during reactor life a l -

though a gradual change in some chemical properties would be expected. Additions of uranium can be made conveniently a s a liquid EiF-UFb mixture19 (liquidus 4 9 0 " ~ )while the reactor is operating, as was done many times during operation o f the MSRE.22 To keep a proper concentration of UF3 in the fuel and possibly to remove tramp oxide-ion contamination from the fuel, some fuel maintenance operations will be necessary.

The

combination of these relatively simple operations likely will result in sufficient addition of LiF and BeF2 to require on-site removal and stsrage of a s m a l l fraction of the fuel before reactor end-of-life.

These

operations and the resulting fuel management options are described in a later section after the chemical basis f o r them has been presented.

*T h i s

type of speration may be p ~ s s i b l ealthough subsequent discussion will show other more likely modes of operation.

50 .

Table 24.

Approximate heavy meta.. i?IV@ntOPy Of end-of-life fuel in hypothetical DMsg with no f u e l removal Inventory

Specie kf2 92 9 000

Mole % 6,84 2.8

x 10-3

1,910

0.140

59s

0.043

250

0.091

97 8

0.671

1 9

28 600 93

2.05 6.7

x 10-3

23 1

6.0165

133

9.5 x 10-3

100

7.1

179

0.0126

136

9.8

10-3

a Operated at 1 GWe f o r 30 years at 75% plant f a c t o r .

3.3,1.2

Physical properties of DMSR fuels

Table 25 shows key physical properties of the compositions identified previously to represent the likely limits for DMSR use. d e t a i l elsewhere,1 9 9 2 0 9 '39

A s described in

several of these properties - particularly

those of the s a l t with 9.5 mole % heavy metal- are interpolated from measurements on similar salt mixtures, Prom careful consideration of v e r y similar mixtures f o r use in the MSBR, the properties clearly are adequate f o r the proposed service.

However, because estimates rather than measured

values are presented in several cases, an experimental program would be

,..'w.'..

T a b l e 25.

Physical properties for probable range of DMSR f u e l compositions Heavy metal content

Properties composition, mole

L i P 71.5 BeFa 16 m 12.5

Liquidus,

508

Properties at 600째C Density, Mg/& Heat capacity, kJ/kg*K

3.10 1.46

3.35 1.36

vi scssity Paes

Centipoise Thermal csnductfvity, W/K% Vapor pressure

0.012

1,2

1.2

<10

<o.

Torr ~~

0.012 12

<10 1

<o.

heavy metal fluorides.

r e q u i r e d to firm up the physical properties of the cOITIpQSitiQn(%)

f o r service.

chosen

Careful reevaluation of the properties would not be likely

t o disqualify these compositions from DMSR use.

3.3.1.3

Chenieal properties of BMSR fuels

A msltem-salt reactor such as a DMSR makes a n m b e r of stringent

demands on its circulating fuel.

Some of these demands have been im-

plicit in the foregoing discussion of f u e l behavior; examples include the obvious need to accommodate moderate concentrations of UF,+ and large ~ ~ n ~ e n t r a t i oof n sThF, in relatively low-melting mixtures of materials with s m a l l cross sections for parasitic neutron capture and the need for adequate heat-transfer capability.

. .*,p.

........

The fuel must be capable of convenient

preparation in a pure homogeneous form for introduction i n t o the reactor. In addition, t h e fuel must (1) be compatible with t h e structural and the moderator materials during normal operation, ( 2 ) be stable to i n t e n s e

radiation fields, and ( 3 ) tolerate â&#x201A;Źissfon of uranium and plutonium and the development of significant concentrations of fission products and p%tttomium and other actinides.

Also, without dangerous consequences, it

must be able to withstand a variety of off-design situations such as heat-exchanger leaks o r possible ingress of air.

Finally, although n o t

presently necessary, it is desirable that at end-of-life the f u e l be amenable t o recovery o f fissile, fertile, and other valuable materials.

The

ability of the f u e l t o meet o r n o t to meet these diverse and conflicting demands l a r g e l y depends on the chemical properties sf the f u e l .

~ o s tof

the details of fission-prsduet behavior are deferred to a subsequent section, but much of the basis f o r expected fuel performance is presented in t h e foBkowing discussion.

Thermodynamics of molten fluoride solutions.

The thermodynamic

properties of many pertinent species in molten LIF-BeF2 solutions and a smaller number in LiF-BeF2-ThFq solutions have been studied in a long-

continued experimental program. employed.

A variety o f experimental techniques w a s

Much of the d a t a was obtained by direct measurement of equilib-

rium concentrations and partial pressures for reactions such as

and

(where g , c , I, and d represent and solute respectively) us%ng dium.

Maray studies of solubility behavior of sparingly s ~ l ~ b lf leu o r i d e

species have a l s o been made,

13aes25926 has reviewed a l l these studies

amd has tabulated themodynamic data f o r many species i n molten Li2BeFq. Table 26 shows values for standard free energy of fomstfsm of major

constituents and s ~ m epossible c o r r o s i o n products.

has a l s o

,.x ..; ... .;* ... &

53 Table 26, Standard free e n e r g i e s of f o r m a t i o n f o r compounds dfsssPved i n m o l t e n Ei2WeP4a

-141.79

16.58

-243.86

30.01

-338.04

40.26

180.4

-445.92

57.85

98,5

491.19

62.50

108.8

-452.96

65.05

98.6

-146.87

36.27

57.1

-154.69

21.78

67.5

-171.82

21.41

76.3

e v a l u a t e d t h e e f f e c t of s o l v e n t composition i n t h e EiF-BeF2

system ow

a c t i v i t y c o e f f i c i e n t s of a v a r i e t y of s o l u t e s . Using a s o p h i s t i c a t e d s p e c t r o p h o t o m e t r i c a n a l y s i s f o r UF,

and U F 3 $

more r e c e n t s t u d y by G i l p a t r i c k and Tsth t o e v a l u a t e t h e e q u i l i b r i u m 1

UF,(d)

W(d) +

i n s e v e r a l LiF-BeF2 and LIF-BeF2-ThF4

;:c.... x..w ..i

cd s o l v e n t s e s s e n t i a l l y confirmed t h e

54 U F ~v a l u e of T a b l e 26 ( i f the UF+ value is a c c e p t e d ) and sl-aowed t h a t t h e

difference between UF3 and UP, (72-16-12

X > i s virtually

mole

s t a n d a r d free e n e r g i e s i n LiF-BeF2-ThFG identical t o t h a t of T a b l e 26,

Balnberger e t a1.27 have shown t h a t

i s -1358 t 10.9 kJ/moPe (-325.4 (--324,6

2 2.6

~3f o r

pup3 in moltera E ~ ~ B ~ F Q

2 . 6 kcal/rnole) and -1357

k ~ d / r a s l e )at 888 and 988 K, r e s p e c t i v e l y .

P 10.9 kJ/mole From t h e s e d a t a

a d from t h e s o l ~ k i l i t yof P u F 3 , they have estimated f o r pure c r y s t a l l i n e PuF3 t h e following v a l u e s :

AGf

-1392

t 10.9 kJ/mole (-333.l

2 2.6

kcal/rnole) a t 988 K

combining t h e s e v a l u e s w i t h t h o s e of Dawaon e t a1.28

f o r the r e a c t i o n

y i e l d s t h e expression

f o r c r y s t a l l i n e BuPk NO

d e f i n i t i v e s t u d y sf

been made.

A G ~f o r

P U F ~i n molten f ~ u o r i d es o l u t i o n has

I t s s o l u b i l i t y (by analogy w i t h t h o s e of ZrF4,, UFh, and ThP4)

i s r e l a t i v e l y high.

A l s o , u s i n g t h e h y p o t h e t i c a l s t a n d a r d s t a t e of u n i t

mole f r a c t i o n , i t i s more s t a b l e [ p e r h a p s by 6 3 kJ/moIe (15 kcal/aaole)] i n solution t h a n a s t h e c r y s t a l l i n e s o l i d .

80,

the r e a c t i o n

The s t a n d a r d s t a t e of PUP i s t h e hypothetical u n i t mole fraction 3(d) used f o r s o l u t e s i n T a b l e 26,

55 would be expected t o have a n e q u i l i b r i u m q u o t i e n t (0) of

where the brackets i n d i c a t e mole fractions of the aissoivea s p e c i e s ,

o n l y 2% of t h e uranium W e r e prese1I8: as UF39 t h e Patio of PuF3bPuP4 Would be n e a r 2 - 5

IO4, and, t o s u s t a i n a PuF3/PuFk r a t i o of 1, o n l y about b

p a r t TJF2 p e r million p a r t s UF4 c o u l d be t o l e r a t e d .

Clearly, unless very

oxidizing c o n d i t i o n s are maintained i n t h e melt, t h e plutonium i s essent i a l l y a l l PU3+. (92-16-12

The s o l u b i l i t y of PuF3 i n LiP-BeFz-ThFq measured i n two l a b o r a t o r i e s .

29330

mole

has been

Bamberges e t a 1 2 9 show t h e s o l u b i l i t y

o f Pup3 in m ~ l eX ( % p u F 3 ) t o be g i v e n by

l o g SpuQ = (3.01

0.06) - ( 2 . 4 1 -%- 0.05)

w i t h a h e a t of s o l ~ t i ~(AH,) n

mole).

of 46.8

1,O

(IOOO/T)

kJ/mole (k1.008

I f s o , t h e solubility of PuF3 a t 565"C,

0.237 k e d /

t h e l i k e l y minimum t e m -

p e r a t u r e w i t h i n t h e BMSW c i r c u i t , s h o u l d b e n e a r 1.36 mole %.T h i s v a l u e

i s n e a r l y i d e n t i c a l ( a s i s t h e heat of s o l u t i o n ) t o t h a t o b t a i n e d by Barton e t

for ~

e i~n t3h e same

solvent.

ow ever,

the I n d i a n study30

gave s i g n i f i c a n t l y lower v a l u e s ; f o r example, at: 565째C t h a t s t u d y would suggest t h a t S p U ~ 3should be n e a r 1.1 ~ l ~ 2.l e The s o l u b i l i t y of PuF3 d o n e is undoubtedly much h i g h e r than is r e q u i r e d for i t s rase in a DMSW,

However, as d e s c r i b e d i n a l a t e r s e c t i o n , d i f f i c u l t i e s p o s s i b l y c o u l d u l t i m a t e l y r e s u l t from t h e combined solubflities of a number of t r i f l u o r i d e s that f ~ m s o l i d solutions.

Given r e a s o n a b l e W , / ~ P , r a t i o s , americium, curium, e a l i f o m f m , and probably neptunium also e x i s t as trifluorides in t h e w e l t .

No d e f i n i t i v e

s t u d i e s of t h e i r s o l u b i l i t i e s i n EiF-BeF2-ThF4 melts have been made.

Such

s t u d i e s are needed, b u t t h e i r i n d i v i d u a l s o l u b i l i t i e s c e r t a i n l y will p r ~ v e

t o be far h i g h e r t h a n t h e i r c o n c e n t r a t i o n s i n DMSB f u e l . Our knowledge of t h e thermodynamics of molten LiF-BeF2-ThF4

soiu-

t i o n s appears a d e q u a t e to g u i d e the n e c e s s a r y development s t u d i e s , b u t

c o n s i d e r a b l e research ana development (as w e l l 8s d a t a analysis) renain

56 t o b r i n g t h a t u n d e r s t a n d i n g t o the l e v e l that exists f o r LiP-BeF2

solu-

tions. O x i d e - f l u o r i d e behavior.

The b e h a v i o r of molten f l u ~ r i d i esystems

such a s t h e DMSB f u e l m i x t u r e c a n be a f f e c t e d markedly by a d d i t i o n of s i g n i f i c a n t concentrations s f oxide

ispa,

For example, we know t h a t c r y s -

t a l s of UO2 p r e c i p i t a t e when w e l t s of EiF2-UF4

a r e t r e a t e d w i t h a reac-

t f v e o x i d e s u c h as water v a ~ o r ~ ~ 9 ~ ~ ~ 9 3 3

The s o l u b i l i t i e s of t h e a c t i n i d e d i o x i d e s i n L ~ F - B ~ F ~ - T ~ F ~ - . U mixFL+ t u r e s a r e low, and they d e c r e a s e i n t h e o r d e r Tho2, Pa02, U0z9 and

oreo over^ t h e s e d i o x i d e s all psssess t h e same ( f l u o r i t e )

~ ~ 0 2 . 216-24 9

c r y s t a l s t r u c t u r e and can a l l form s o l i d s o l u t i o n s w i t h one a n o t h e r . Sokmbility p r o d u c t s and their t e m p e r a t u r e dependence have been meas u r e d ; 34-43

t h e i r behavior i s g e n e r a l l y we11 understood.

T r i v a l e n t plutonium shows l i t t l e or no tendency t o p r e c i p i t a t e as o x i d e from LiF-BeF2-ThPq-UFq

mixtures

Because r e l a t i v e l y l a r g e so%u-

b i l i t y seems t o be g e n e r a l f o r t r i v a l e n t o x i d e s , i t is h i g h l y l f k e k y

t h a t p r e c i p i t a t i o n of h 2 0 3 and o t h e r t r i v a l e n t a c t i n i d e s would b e d i f f i c u l t t o achieve.

If Pa4+ i s oxfdized to Pa5+ (which can b e d o ~ er e a d i l y i n LiF-BeF2ThF4-UF4 (QP

by t r e a t m e n t w i t h anhydrous and hydrogen-free

KF g a s ) , t h e n Pa205

an a d d i t i o n compound of i t > can be p r e c i p i t a t e d s e l e c t i v e l y . 4 l 3 4 4

such o x i d a t i o n t o pas* can be avoided by m a i n t a i n i n g a small f r a c t i o n of t h e uranium as UF3 i n t h e f u e l mixture. he relaiively ~ s oxide w t o l e r a n c e of DMSR f u e l will r e q u i r e reason-

a b l e care t o a v o i d i n a d v e r t e n t precipitation of a c t i n i d e o x i d e s w i t h i n t h e r e a c t o r system.

However, t r e a t m e n t o f melts w i t h anhydrous I-iF ( e v e n when

s u b s t a n t i a l l y d i l u t e d w i t h ii2> s e r v e s t o Lower t h e o x i d e c o n c e n t r a t i o n t o t o l e r a b l e l e v e l s . 18,45

C o m p a t i b i l i t y of f u e l w i t h r e a c t o r materials, e x c e l l e n t f l u x e s f o r many materials.

Molten f l u o r i d e s a r e

Though some oxides are r e l a t i v e l y

i n s o k u b l e , most are readily d i s s o l v e d , and d l are r a p i d l y recrystall i z e d ; c o n s e q u e n t l y , p r o t e c t i v e c o a t i n g s a r e n ~ uts e f u l , and t h e bare

c l e a n metal must w i t h s t a n d

loy-N,

described

Prt

C Q P ~ Q S ~ aVt t~a c k .

The r e a c t o r metal (Hastel-

d e t a i l i n S e c t , 3 . 4 ) was chosen and t a i l o r e d t o be

t h e g m ~ d p ~ i ~ i e a sl tl ayb l e t o t h e f u e l components, a s much as p o s s i b l e ,

._. . ...... ,.. w

Corrosion of Hastelloy-N by MSRE and MSBR flue1 mixtures without irradiation and without the consequences of fission has been studied in sophisticated equipment for many years.

It has been thoroughly de-

Table 26 clearly indicates that chromium is the most easily o x i d i z e d of the major Hastelloy-N components.

C ~ r r o s i o nof the alloy, therefure,

is essentially by selective leaching of chromium from the a l l o y .

A rapid

initial attack can result from reactions such as

and

if the fuel s a l t is impure o r if the metal system is p o o r l y cleaned. These reactions proceed t o completion at all temperatures within the reactor circuit and do not afford a basis f o r continued attack. The most oxidizing of the major constituents sf t h e fuel is UF4, and the reaction

has an equilibrium constant with a small temperature dependence. When the salt i s forced to circulate very rapidly through a large ( 1 4 0 째 C ) temperature gradient,

as i s

the case within the reactor circuit, a mechanism

exists for mass transfer u f chromium and f o r continued attack.

The re-

sult is that chromium is selectively removed from t h e alloy in higktemperature regions and deposited on the alloy in low-temperature regions

*Molybdenum

f l u o r i d e s are somewhat less s t a b l e t h a n NdF2. M [ S P S ( ~ ) at 906 K has a standard free energy of formatfon25 of about 21% kJ (-51.4 kcal) per gram-atom of F--

.... *.& , s....

: i

58 The r a t e of t r a n s f e r of chromium i s l i m i t e d by t h e r a t e

of t h e r e a c t o r .

a t which t h e t r a n s f e r r e d

~ ~ ~ Q I I I ~ LaK InI

temperature r e g i ~ n "~9 . 2 0 $ 56

d i f f u s e i n t o t h e a l l o y i n t h e low-

The r e s u l t s of two decades of s o p h i s t i c a t e d

c o r r o s i o n t e s t i n g have dem~nstratedt h e v a l i d i t y of t h i s mechanism and

have shsm t h a t s u c h f o r MSBW.

w i l l p r ~ t~o ebe o n l y a t r i f l i n g problem

C O ~ ~ O S ~ O - ~ "

Appreciable chromlium d e p l e t i o n W Q U $be ~ expected t o a d e p t h of

less than 0.13 mm/year (0.5

m i l / y e a r ) i n metal a t TO~~C.''~~'

The i n i t i a l a t t a c k , which i s n o t s e r i s u s i f p r o p e r p u r i f i c a t i o n of t h e s a l t and c l e a n i n g of t h e system have o c c u r r e d , c a n be m i t i g a t e d by

t h e presence of a s m a l l q u a n t i t y of UFq a l o n g w i t h UF4 i n t h e s a l t .

Csn-

t r o l of t h e o x i d a t i o n s t a t e s of plutonium and p r o t a c t i n i u m and of c e r t a i n f i s s i o n p r o d u c t s , a l o n g w i t h c o n t r o l o f t h e o x i d a t i v e e f f e c t s of t h e f i s s i o n process, f u r n i s h more cogent r e a s o n s f o r m a i n t a i n i n g %iF3 i n t h e f u e l mixture.

S l i g h t c o n t i n u i n g c o r r o s i o n i s a f f e c t e d very l i t t l e ( i f a t a l l )

by t h e p r e s e n c e of s m a l l q u a n t i t i e s of UF3.

The unclad moderator g r a p h i t e i s not wetted by o r c h e m i c a l l y r e a c t i v e t o t h e Standard MSRE Q F HSBR f u e l COmpQSit%QnS,and t h e s e facts a p p e a r Un9 %203 57,58 changed by i n t e n s e i r r a d i a t i o n and t h e e o ~ s e ~ u e n c eofs f i s s i ~1~ ~.

Estimates22 a r e that t h e MSRE g r a p h i t e moderator s t a c k (3700 kg) a c q u i r e d

less t h a n 2 g of uranium d u r i n g ~ p e r a t i of ~ ~ tlh e r e a c t o r .

ObviousPy, no

a p p r e c i a b l e i n t e r a c t i o n of g r a p h i t e w i t h the f u e l (whose UF3/&YFb, ratio w a s

never above 0.02) o c c u r r e d i n t h a t r e a c t o r .

Mowever, g i ~ ae s~u f f i c i e n t l y

h i g h UF3/UP4 r a t i o , formation of uranium c a r b i d e s must be e x p e c t e d , and t h i s should be avoided,

TQth a d G i l p a t r f c k , W h o used SpectHOphQtOmetrg

i n a g r a p h i t e c e l l w i t h diamond windows t o a s s a y e q u i l i b r i u m UF3 and UF4

c o n c e n t r a t i o n s , have c a r e f u l l y s t u d i e d u r a n i m c a r b i d e f o m a t i ~ nu s i n g Li2BeF4 ( R e f s .

59 and 60), o t h e r LiF-BeF2 mixtures,6Q ,61 and LiF-BeP2-TkP4

(72-16-12

X ) (Ref. 4 1 ) as solvents.

mole

A s u r p r i s i n g f i n d i n g of t h e s e

s t u d i e s i s t h a t , c o n t r a r y t o g e n e r a l l y a c c e p t e d themodynamic d a t a , "

U C ~

i s t h e s t a b l e c a r b i d e phase over t h e t e m p e r a t u r e i n t e r v a l 550 to 700째C. t h e r e s u l t s 6 ' of equilibration experiments i n MSBR f u e l

F i g u r e I12

S~QWS

solvent.

A p p a r e n t l y , a t t h e l o ~ e s t e m p e r a t u r e (565째C) w i t h i n a BNSR,

"~srrosion i n the p r e s e n c e of f i s s i o n and f i s s i o n p r o d u c t s i s more See Sects, 3 . 3 . 2 and 3.4 f o r a d d i t i o n a l d e t a i l s .

complex.

TEMPERATURE

500

550

600

cue,

OWNL-DWG 72-42324

656

766

-4

-2

-3

-4

m 0 -

-5

-6

-7

-8 1.38

.;;..,..I ."LW .....

f.25

2.28

f.f5

1.40

1.85

1.OQ

60 3.3.1.4

Operational c~nstraintsand uncertainties

Most of the fuel behavior described or implied above may be consid-

ered well authenticated, Several constraints and at least minor uncertainties are obvious. 1.

F u e l for a DMSR must be prepared from LiF containing a very high

The fuel mixture must be managed and maintained so that an appre-

ciable fraction of the uranium is present as UP3. 3.

Additional experiments are necessary t o establish exactly what

fraction of the uranium may b e present as UFcj without deleterious chemical reactions of the UP3 with graphite or possibly with other materials within the primary reactor system.

Direct measurement of the physical and heat-transfer properties

of the BMSR fuel mixture must be made.

Further study of the fundamental thermodynamic properties of so-

lutes in the LiF-BeF2-ThF4-UF4

mixture are needed t o ensure that basic

understanding of the chemical behavior is accurate. 3.3.2

Fission-product behavior

Fragments produced on fission of a heavy atom originate in energy and with i ~ l s i z a t i ~levels n far from those normally considered in chemical reactions. When the fission occurs in a well-mixed mo%ten-salt states

liquid medium, these fragments must come t o a steady state as comonHy encountered chemical entities because they quicldy l o s e energy through eolLisions with t h e medium.

The valence states that these chemical spe-

cies assume are presumably defined by the requirements that (1) cationanion equivalence b e maintained in the molten-salt medium and ( 2 ) redox e~uilibriabe established between the melt and the surface layers of the he fission-product cations must satisfy the container metals1 9 9 6 3 9 6 4 fission-product anions plus the fluoride i o n s released by disappearance

*The

rapid radioactive decay of many species further complicates an already complex situation.

ea of the f i s s i o n e d atom.

~ a r l yassessment"

i n d i c a t e d t h a t the c a t i o n s

would prove adequate o n l y if some o f them assumed oxidation s t a t e s corr o s i v e t o H%%telPQy-N. A

FBIOPB

%â&#x201A;Ź!Cent eXEU'ilin21tion25 strongly SUppOrted

t h i s view; t h e s e studies i n d i c a t e d t h a t t h e summation of t h e p r o d u c t s of f i s s i o n y i e l d and stable v a l e n c e f o r each s p e c i e s wight be as l o w a s t h r e e p e r f i s s i o n event.

( ~ r -+

1-1

Accordingly, fission of UP4 [ r e l e a s i n g 4 I?-+

0.815

p e r f i s s i o n ] would^ be i n t r i n s i c a ~ ~oxidizing y t o ~ a s t e l l o y - ~*.

Maintenance of a small f r a c t i o n of t h e uranium in the f u e l as UF3 w a s s u c c e s s f u l l y adopted t o p r e c l u d e c o r r o s i o n from f i s s i o n of 2 3 5 ~ ~i n4 t h e

MSWE.~A ~ p r o p e r l y maintained redox p o t e n t i a l i n t h e f u e l s a l t a p p a r e n t l y w i l l p r e v e n t any untoward immediate c o ~ s e q ~ e n c es sf t h e f i s s i o n e v e n t and

will p e r n i t grow-in of t h e f i s s i o n p r o d u c t s in v a l e n c e s t a t e s d e f i n e d by t h e redox p o t e n t i a l .

3.3.2.2

E f f e c t s of r a d i a t i o n

When f i s s i o n o c c u r s i n a molten f l u o r i d e solution, b o t h eleetromagnetic r a d f a t i o ~and ~ p a r t i c l e s o f v e r y h i g h energy arnd i n t e n s i t y originate within the fluid.

Local o v e r h e a t i n g i s almost c e r t a i n l y n o t

important i n a DMSR where t u r b u l e n t f l o w causes r a p i d i n t i m a t e mixing. Moreover, t h e bonding i n molten f l u o r i d e s i s completely i o n i c .

Such a

m i x t u r e , w i t h n e i t h e r covabernt bonds t o r u p t u r e n o r a l a t t i c e t o disruptp should be q u i t e r e s f s t a n t to r a d i a t i o n .

N e v e r t h e l e s s , because t h e r e pbau-

s i b l y e x i s t s a r a d i a t i o n l e v e l s u f f i c i e n t l y h i g h t o d i s s o c i a t e a molten f l u o r i d e i n t o metal and f l u o r i n e , + a number of tests of t h e p o s s i b i l i t y

were wade. 19,2 0 9 6 3 9 65 Nany i r r a d i a t i o n t e s t s were conducted p r i o r t o 1959 w i t h NaF-ZrP4-UP4 m i x t u r e s i n I n e o n e l a t t e m p e r a t u r e s a t o r above 8 1 5 O C (Refs. 2 4 and 2 5 )

and a t q u i t e high fission power d e n s i t i e s from 80 t o 1080 ~ ~ / of m 3f u e l . No i n s t a b i l i t y o f t h e f u e l system was a p p a r e n t , and t h e c o r r o s i o n d i d n o t

exceed t h e c o n s i d e r a b l e anmount expected from l a b o r a t o r y - s c a l e

tests.

" F i s s i o n sf PuF3 probably would be n e a r l y n e u t r a l i n t h i s regardp and fission of a m i x t u r e of UF, and P u F 3 $ a s in t h e DMSR, would be i n t e r m e d i a t e between t h e s e extremes.

?This would occur even though the r a t e of recombination o f LiO and Fo should be extremely r a p i d .

62 These e a r l y t e s t s seemed t o show t h a t r a d i a t i o n posed no t h r e a t even

a t v e r y h i g h power l e v e l s , b u t f u r t h e r studies66-69 were conducted primarily t o t e s t t h e w e t t i n g of g r a p h i t e by LiF-BeP2-ZrFq-ThF4-UPq under i r r a d i a t i o n .

mixtures

Examination of t h e s e c a p s u l e s a f t e r storage a t m b i -

ent t e m p e r a t u r e s f o r many weeks r e v e a l e d a p p r e c i a b l e q u a n t i t i e s of CF4 and, i n most cases, c o n s i d e r a b l e q u a n t i t i e s of f l u o r i n e i n the c o v e r gas. C a r e f u l e ~ m i n a t i o n 2 0 ~ 7 0 ~strongly 71 s u g g e s t e d t h a t the p2 g e n e r a t i o n had n o t occurred a t t h e h i g h t e m p e r a t u r e b u t had occurred by r a d i o l y s i s of t h e mixture i n the s o l i d state.

This s u g g e s t i o n was confirmed by i r r a d i a t i o n of two a r r a y s sf

Nastelloy-M c a p s u l e s , a l l c o n t a i n i n g graphite and LiF-BeF2-ZrF4-UF4 tures.

mix-

Two of t h e c a p s u l e s i n e a c h a r r a y had gas i n l e t and e x i t l i n e s t o

p e r m i t sampling of t h e cover g a s as d e s i r e d .

Gas samples d r a m from t h e

t e s t c a p s u l e s a t o p e r a t i n g t e m p e r a t u r e s and a t v a r i o u s power l e v e l s up t o 80 m / m 3 showed

H ~ QF~

( t h o u g h an occasional sample from the f i r s t array

showed d e t e c t a b l e traces o f CF,).

However, d u r i n g reactor shutdowns w i t h

t h e c a p s u l e s a t about 3 % " 6 , p r e s s u r e rises were observed ( u s u a l l y a f t e r a n i n d u c t i o n p e r i o d of a few Bmours), and F2 w a s evolved.

In t h e second

array, t h e c a p s u l e s were k e p t h o t d u r i n g r e a c t o r shutdown as w e l l as Buri ~ l go p e r a t i o n ;

evidence of F~ o r CF, w a s observed.

such

-17~

generation

at ambient t e m p e r a t u r e s was s u b s e q u e n t l y followed f o r s e v e r a l months i n OREaE h o t c e l l s .

The g e n e r a t i o n diminished with time in a manner c o r r e s -

ponding c l o s e l y w i t h decay of f i s s i o n - p r o d u e t a c t i v i t y ; F 2 e v o l u t i o n a t 35째C corresponded t~ about 0,02 molecule p e r 188 e V a b s o r b e d , e o d d be Completely stopped by h e a t i n g t o I.OO"C o r above, and c o u l d be r e s f u ~ e d

markedly by c h i l i l i n g to -7O"C.

The F2 e ~ s l u t i o nresumed, u s u a l I ~a f t e r

a few hours, when t e m p e r a t u r e was r e t u r n e d to 35 t o 5 0 " C e These and subsequent e x p e r i e n c e s , i n c l u d i n g o p e r a t i o n of t h e MSRE, s t r o n g l y i n d i c a t e t h a t r a d f o l y s f s of the molten f u e l a t r e a s o n a b l e power d e n s i t i e s i s n o t a problem.

It seems u n l i k e l y , though it i s p o s s i b l e ,

t h a t DMSR f u e l s will e v o l v e F2 on c o o l i t ~ g , If t h e y d o , arrangements must be made for t h e i r s t o r a g e a t e l e v a t e d temperature until a f ~ a ~ t i osf n the decay energy i s d i s s i p a t e d .

n e results of a program of s o l u t i o n themodynamics,25926 a ~ o n g - t e m

program of i n - p i l e i r r a d i a t i o n s , 5 ’

atad a number of s p e c i a l experiments

g e m m e d g e n e l ~ a i i yaccurate p r e a i e t i o n s , 63-65 b u t much of our d e t a i l e d -

and s t i l l incomplete

- understanding

from o p e r a t i o n of t h e ~

of f i s s i o n - p r o d u c t

behavior comes

m ~ e ~ability 9 2 of2 t h~e f~i s s2f s n produets

to form s t a b l e compounds and t o d i s s o l v e i n t h e molten f u e l serves t o d i v i d e them i n t o t h e t h r e e distinct groups d e s c r i b e d i n the f o l l o w i n g

dissussisn. e

Krypton and xenon (which

i s an important n e u t r o n absorber) f o m no eorrapounds under c o n d i t i o n s exi s t i n g i n a DMSR o r o t h e r molten-salt reactor.19,73

Noreover, t h e s e

g a s e s are only v e r y s p a r i n g l y s o l u b l e i n molten f l u o r i d e mixtures. 74-76 A s w i t h a l l n o b l e gases ( s e e Fig.

l a ) , 76 t h e i r s o l u b i l i t y i n c r e a s e s w i t h

t e m p e r a t u r e and w i t h diminishfng size of t h e gaseous atom, while t h e h e a t

of s o l u t i o n i n c r e a s e s w i t h i n c r e a s i n g atomic s i z e .

T h i s low s o l u b i l i t y

i s a d i s t i n c t advantage because i t e n a b l e s the r e a d y removal sf krypton

and xenon from t h e r e a c t o r by sparging w i t h helium.

The r e l a t i v e l y sim-

p l e s p a r g i n g system of the MSRE served t o remove more than 80% of t h e 13%e, and f a r more efficient spargbng w a s proposed f o r the MSBR.”

S t r i p p i n g of t h e n o b l e g a s e s from t h e r e a c t o r a f t e r a s h o r t r e s i d e n c e

time a v o i d s t h e p r e s e n c e of t h e i r r a d i o a c t f v e d a u g h t e r s i n t h e f u e l . Tritium qualifies as a f i s s i o n product because mall quantities of i t are produced i n t e r n a r y f i s s i o n s .

However, e s s e n t i a l l y all of t h e

tritium a n t i c i p a t e d i n M S B R ’ Q , ~r~e s u l t s from other s o u r c e s , a s shown i n Table 27. A DMSW a t similar power l e v e l and w i t h a g e n e r a l l y similar f u e l must be expected t o g e n e r a t e tritium a t approximately t h i s same rate.

tritium w i l l o r i g i n a t e i n p r i n c i p l e a s 3 e s n e e n t r a t i o n s of ~3 A

This

~ however, ~ ; with appreciable

p r e s e n t , this ’HF w i l l be reduced l a r g e l y t o %2.

However, note t h a t t h e p o r e s i n t h e moderator g r a p h i t e can o f f e r a haven far &heSE? gases.

64 QRNL-68-BWG 44908

TE M PE RATU R E ( O 6 1

800

500

700 650 680

-+--

Fig. 13, S ~ P u b i l i t k sof f o u r noble gases as function of temperature in LiF-BeFz ( 6 4 - 3 6 mole

65 Table 27. Sources and r a t e s p r o d u c t i o n of tritium i n a 1000-m~e M S B P

Production rate

Soharce MBq/s

ci/a

6 L i ( * , a ) 3H

518

1210

%i(n 9 na>3H

501

1170

1036

2420

Ternary f i s s i o n

1 9 ~ ~ ~ , 1 7 ~ ) 3 ~

Total “From ~ e f .7 7 .

through h o t metals w i l l p e r m i t a Parge f r a c t i o n of the ’82 t o p e n e t r a t e t h e primary h e a t exchanger t o e n t e r t h e secondary c o o l a n t .

T h i s phenome-

non and i t s consequences are d e s c r i b e d b r i e f l y i n S e c t . 3 . 3 . 3 . 2 , F i s s i o n p r o d u c t s w i t h s o l u b l e s t a b l e compounds.

Rubidium, cesium,

s t r o n t i u m , barium, yttrium, the l a n t h a n i d e s , and zirconium a l l form q u i r e s t a b l e f l u o r i d e s t h a t are r e l a t i v e l y s o l u b l e i n molten f l u o r i d e m i x t u r e s such as MSBR and DMSR f u e l s .

I s o t o p e s of t h e s e e l e m e n t s t h a t have no

noble-gas p r e c u r s o r s , as e x p e c t e d , appeared almost e n t i r e l y i n t h e c i r c u lating f u e l s € the MSRE.19,2Q,22*72

Very s m a l l q u a n t i t i e s appeared a t ~r

n e a r the s u r f a c e o f exposed g r a p h i t e spec5mens; most of t h i s d e p o s i t i o n e v i d e n t l y r e s u l t e d from f i s s i o n r e c o i l .

I s o t o p e s such as 89Sr and 14’Ba,

whose v o l a t i l e p r e c u r s o r s have a p p r e c i a b l e h a l f - l i v e s and which were part i a l l y s t r i p p e d from t h e r e a c t o r , w e r e found i n samples o f t h e c o v e r g a s and w i t h i n specimens o f moderator g r a p h i t e as w e l l a s i n the f u e l of t h e MSRE.

Along w i t h b e h a v i o r of o t h e r i s o t o p e s , Fig. 14 shows t h e p r o f i l e s

observed for 1 3 . p ~and ~ 1 4 o ~ ai n g r a p h i t e specimens through d i f f u s i o n of t h e i r r e s p e c t i v e (137xe, 3 . 9 min; 140xe, 16 s > p r e c u r s o r s . Bromine and i o d i n e would be expected t o a p p e a r i n t h e f u e l a s s o l u b l e B r - and I-, p a r t i c u l a r l y i n t h e ease where the f u e l contains a n apgre-

ciable c o n c e n t r a t i o n of UP^.

a n a l y s e s f o r ~ r were performed d u r i n g

OWL-DWG 68-10155R

4 Q‘7

40” 0

90 100 440 420 130 DEPTH IN FQAPHITE (mils)

140

(50

460

470

480

490 200

216

Fig. 14. Fission product distribution in CGB ($55) graphite specimen 3sed i n MSWE core d u r i n g 32,000 sf power o p e r a t i o n . 79

67 o p e r a t i o n of t h e PISRE.

a n a l y s e s f o r 1311 showed t h a t a l a r g e f r a c t i o n of

the i o d i n e w a s p r e s e n t i n t h e f u e 1 2 0 ~ 7 2and t h a t 1 3 1 1 d e p o s i t e d on metal o r g r a p h i t e s u r f a c e s i n t h e c o r e region.

were g e n e r a l l y low.

However, material b a l a n c e s f o r

~t i s p o s s i b l e t h a t some of t h e p r e c u r s o r , 1 3 1 ~ e

(25 m i n ) , w a s v o l a t i l i z e d and sparged w i t h t h e krypton and

X~H~QR.

Fur-

t h e r , 1 3 1 ~produced by decay of E ~ I Ti n~ complex m e t a l l i c d e p o s i t s (as in t h e h e a t exchanger) may n o t have been a b l e t o r e t u r n t o t h e s a l t . Noble and seminoble f i s s i o n products. ( ~ e AS, ,

Nb,

RU,

Some f i s s i o n - p r o d u c t metals

~ h P, a , Ag, cd, SW, and s b ) have f l u o r i d e s t h a t itre

u n s t a b l e toward r e d u e t i o n by fuel m i x t u r e s w i t h a p p r e c i a b l e eoncentra-

tions of UPQ; t h u s , t h e y must b e expected t o e x i s t e n t i r e l y %n t h e elemental s t a t e in the r e a c t o r .

Selenium and t e l l u r i u m were a l s o expected

t o be p r e s e n t as elements w i t h i n t h e r e a c t o r c i r c u i t , and t h i s behavior

w a s g e n e r a l l y c s n f i r n e d during o p e r a t i o n of t h e MSRE. 19920 $ 7 2

he MSRE

f u e l samples u s u a l l y c o n t a i n e d far less t h a n the g e n e r a t e d q u a n t i t i e s of t h e s e elements.

P o r t i o n s of the MSRE samples were found ( p r o b a b l y as

metalPic p a r t i c u l a t e s ) i n t h e helium s p a r g e gas,* d e p o s i t e d on metal surf a c e s , and (a r e a s o n a b l y small f r a c t i o n ) d e p o s i t e d on g r a p h i t e specimens. However, the d i s t r i b u t i o n and e s p e c i a l l y t h e i n v e n t o r y i n t h e f u e l a t t h e sampling p o i n t i n t h e pump bowl showed major v a r i a t i o n s .

Further study

w i l l be n e c e s s a r y b e f o r e d e t a i l s of t h e i r behavior can be p r e d i c t e d w i t h c o n f i d e n c e f o r a DMSR. ~n g e n e r a l , t h e r e s u l t s from MSRE o p e r a t i o n s suggest t h e following,20

The bulk o f the n o b l e metals remain a c c e s s i b l e i n t h e c i r c u l a t i n g l o o p

but w i t h w i d e l y v a r y i n g amounts i n c i r c u l a t i o n a t any p a r t i c u l a r t i m e .

I n s p i t e of t h i s wide v a r i a t i o n i n t h e t o t a l amount found i n a part i c u l a r sample, t h e p r o p o r t i o n a l composition i s r e l a t i v e l y c o n s t a n t , i n d i c a t i n g that t h e e n t i r e i n v e n t o r y i s i n s u b s t a n t i a l e q u i l i b r i u m w i t h t h e new material being produced.

Much-improved gas-sampling t e c h n i q u e s used i n l a t e r s t a g e s of MSRE o p e r a t i o n showed t h e f r a c t i o n c a r r i e d i n t h e g a s t o b e less t h a n 2% o f t h e q u a n t i t y producedâ&#x20AC;? ( w i t h t h e p o s s i b l e e x c e p t t o n of i l i: 3% of k o 3 ~ u ) .

The m o b i l i t y of t h e p o d of noble-metal material s u g g e s t s t h a t dep o s i t s occur as an a c c m u l a t f s n of f i n e l y d i v i d e d well-mixed material r a t h e r than as a Suck p r e c i p i t a t i o n w i t h i n t h e r e a c t o r , though e x p e c t e d , i s a dis-

adVantag@. P r e c f p f t a t a o n

OIE

t h e metal surface (most Sf Which iS in t h e

heat exchanger) w i l l be q u i t e i n s u f f i c i e n t t o impede f u e l f l o w , b u t r a d i o a c t i v e decay SS t h e d e p o s i t e d material contributes t o h e a t generation BuriRg

?X?actor SfiutdoWII.

Precipitation

OFI

the mOdePatOlP g r a p h i t e , Which ap-

peared t o be c o n s i d e r a b l y smaller than on t h e metal, would maximize t h e i r opportunities t o absorb valuable neutrons. Operation Of t h e MSaE d i d prOdUCe One products,

WktOWZ3sd

e f f e c t Sf f i s s i o n

Metal s u r f a c e s exposed t o t h e f u e l in the MSRE showed g r a i n

boundaries that were e m b r i t t l e d to d e p t h s of 0 - 1 t o O.?

IILH~

( 5 t o 10 mils).

I n t h e h e a t exchanger, the e m b r i t t l e d boundaries opened t o form metallsg r a p h i c a l l y v i s i $ % @c r a c k s ; i n o t h e r r e g f s n s such cracks formed

t h e specimens were d e l i b e r a t e l y s t r a i n e d .

QES%~ when

E a r l y studies*0 i m p l i c a t e d

f i s s i o n - p r o d u c t t e l l u r i m a s r e s p o n s i b l e f o r t h i s e m b r i t t l e m e n t , and subSeqUE?Ist WQPk

has CQnfi9fmed thiS."6a4*

However,

mOPe

Irecent S%Udfes46,81982

strongly s u g g e s t that (1) i f t h e molten fuel i s made t o c o n t a i n as much

a s 5% of the u r a n i m 8s U

F t~h e ~t e ~ ~ u r iwould m be p r e s e n t as

( 2 ) in t h a t f o m , tellurium is much less a g g r e s s i v e .

~ e 2 - and

Much f u r t h e r s t u d y

w i l l be n e c e s s a r y , b u t use s f t h i s h i g h e r but s t i l l moderate UF,/IJP,

ratlo

a p p a r e n t l y will. markedly a l l e v i a t e , and probably c o n t r o l , the t e l l u r i m embrittlemerat problem.

3.302.4

Operational ConSthaintS

Avoiding the d e t r i m e n t a l e f f e c t s s f f i s s i o n - p r o d u e t t e l l u r i u m (des c r i b e d immediately preceding) may wake n e c e s s a r y t h e o p e r a t i o n of t h e DMSR w i t h as mush as 5% of t h e uranium f l u o r i d e present as UF3"

C~nsfd-

e r a t i o n of p o s s i b l e r e a c t i ~ n sof ~ T F t o~ produce uranium c a r b i d e s (des c r i b e d preaiousPy) s u g g e s t s t h a t o p e r a t i o n w i t h a c o n s i d e r a b l y h i g h e r A

T h i s subject i s addressed further in Sect, 3 . 4 ,

& .& !-.,.

UP3/UF4 r a t i o would be p o s s i b l e .

knm.rer, a more s u b t l e c o n s t r a i n t on

DMSR o p e r a t i o n p o s s i b l y may r e s u l t . The l a n t h a n i d e t r i f l u o r i d e s are o n l y moderately s o l u b l e i n w ~ l t e n LiF-BeP2-ThP4-UF4

m i x t u r e s ; i f more than one such t r i f l u o r i d e i s p r e s e n t ,

t h e y c r y s t a l l i z e as a s o l i d s o l u t i o n of all t h e t r i f l u o r i d e s on c o o l i n g

of a s a t u r a t e d m e l t .

Though n o t d e f i n i t i v e , t h e r e is evidence t h a t t h e

a c t i n i d e t r i f l u s r i d e s ( i n c l u d i n g W 3 > might a l s o j o i n in such s o l i d solutions.

I f s o , the t o t a l ( l a n t h a n i d e p l u s a c t i n i d e ) t r i f l u o r i d e s i n t h e

end-of-life

r e a c t o r might p o s s i b l y exceed t h e i r combined s o l u b i l i t y .

The solubilities of PuFq ( R e f .

29) and CeF3 ( R e f .

f u l l y determined i m % ~ F - B ~ F ~ - (72-16-12 T ~ F ~

3 1 ) have been cape-

mole % > and e m be e o ~ ~ i h f e r e d

t o be moderately w e l l known i n DMSR f u e l .

According t o Baes e t al 9 2 9

t h e s o l u b i l i t y of PuF3 i n t h e LiF-BeF2-ThFq

melt at 565째C ( t h e minimum

temperature a n t i c i p a t e d within the BMSR f u e l c i r c u i t ) i s 1.35 mole %.The s o l u b i l i t y of CeP3 umder t h e same c o n d i t i o n s appears t o be v e r y s l i g h t l y

smaller ( 1.3 m o l e % ) .29 a r e n o t w e l l known.

9 3

S o l u b i l i t i e s of the o t h e r p e r t i n e n t f l u o r i d e s

I n L12BeF4, t h e s o l u b i l i t i e s of s e v e r a l l a n t h a n i d e

t r i f l u o r i d e s ( i n c l u d i n g CeF3 ) have been shown83 t o be c o n s i d e r a b l y smaller and t o v a r y w i t h some more and some l e s s s o l u b l e t h a n CeF3.

As a reason-

a b l e approximation (~bviously,many a d d i t i o n a l d a t a are needed), t h e s o l u b i l i t y of t h e l a n t h a n i d e - a c t i n i d e t r i f l u o r i d e s o l i d s o l u t i ~ nmay b e

assumed t o be n e a r 1 . 3 mole 2.

From T a b l e 9 , t h e BMSR f u e l a t e n d - o f - l i f e

will c o n t a i n some 11.404

185 moles of uranium i s o t o p e s and 3 . 6 4 x PO3 moles of transuranium i s o topes.

The end-of-life

b e n e a r 4.7

x 104

moles.

i n v e n t o r y of l a n t h a n i d e p l u s y t t r i u m i s o t o p e s w i l l I f 5X of t h e uranium i s p r e s e n t a s U F and ~ if

a l l t r a n s u r a n i c and l a n t h a n i d e s p e c i e s are assumed t o be t r i v a l e n t , the end-of-life rides.

r e a c t o r f u e l w i l l c o n t a i n about 5.77

x 104

moles of t r i f l u o -

'Fke DMSR system ( w i t h about 5.3 x IO6 moles of f l u o r i d e s ) , t h e r e -

f o r e , would c o n t a i n about 1.1 mole % of t r i f l u o r i d e s .

The s o l u b i l i t y o f

the combined t r i f l u o r i d e s l i k e l y would not b e exceeded w i t h i n t h e r e a c t o r c i r c u i t , b u t a d d i t i o n a l s o l u b i l i t y data are needed t o make t h i s p o i n t

certain.

3,3,2.5

U n c e r t a i n features

Prom t h e f o r e g o i n g d i s c u s s i o n , s e v e r a l u n c e r t a i n t i e s are a p p a r e n t .

Details of b e h a v i o r of t h e n o b l e and seminoble f i s s i o n p r o d u c t s are s t i l l p o o r l y known.

The f r a c t i o n s o f each i s o t o p e t h a t w i l l appear in t h e o f f -

g a s , d e p o s i t on t h e m o d e r a t s t g r a p h i t e , and d e p o s i t

metal u% t h e I>r/nยงR c a n o n l y be c r u d e l y e s t i m a t e d .

t h e heat-exchanger

That f r a c t i o n which a p -

p e a r s i n t h e r e a c t o r o f f - g a s would seem t o cause no insurmountable prsb-

l e m s though t h a t s y s t e m - a t our p r e s e n t s t a t e of knowledge - w o u l d

t o be overdesigned s u b s t a n t i a l l y .

need

T h e i r d e p o s i t i o n i n t h e h e a t exchanger

is a recognized d i s a d v a n t a g e and w i l l be quite i n s u f f i c i e n t to impede f u e l flow, b u t r a d i o a c t i v e decay of t h e d e p o s i t e d material c o n t r i b u t e s h e a t that must be removed d u r i n g r e a c t o r shutdown.

Precipztation on the

g r a p h i t e , which a p p e a r s t o be smaller t h a n on heat-exchanger metal, maxi-

mizes t h e i r o p p o r t u n i t i e s t o a b s o r b neutrons.

C l e a r l y , a b e t t e r knowl-

edge of t h i s s i t u a t i o n i s weeded,

While t h e r e s u l t s o b t a i n e d i n t h e r e c e n t p a s t are highly encouraging, additional d a t a - e s p e c i a l l y on a l a r g e r scale - are needed t o e s t a b l i s h the redox p o t e n t i a l ( U F ~ / U Pr~a t i o ) required t o keep t h e tebburium cracki n g problem t o t o l e r a b l e l e v e l s . s u b s t a n t i a l l y above 0.85,

Should t h e r e q u i r e d UF3/UF4 r a t i o rise

t h e p r o b a b i l i t y of p r e c i p i t a t i o n of t r i f l u s r i d e

solid s o l u t i o n s would be i n c r e a s e d . F i n a l l y , a d d i t i o n a l i n f o m a t i o n abseat t h e c o l l e c t i v e s o l u b i l i t y beh a v i o r of t h e l a n t h a n i d e - a c t i n i d e t r f f l u o r i d e s i s r e q u i r e d .

Should t h e y

prove a p p r e c i a b l y less s o l u b l e than wow b e l i e v e d l i k e l y , some replacement

of f u e l might b e r e q u i r e d l a t e i n t h e l i f e of t h e e s s e n t i a l l y unprocessed DMSR e

3.3.3

Fuel. maintenance

To a c h i e v e f u e l maintenance, ( 1 ) t h e f u e l must be d e l i v e r e d t o and

i n t o t h e r e a c t o r i n a proper s t a t e of p u r i t y and homogeneity, ( 2 ) the f u e l must be s u f f i c i e n t l y p r o t e c t e d from e x t r a n e o u s i m p u r i t i e s , and ( 3 ) sound p r o c e d u r e s must e x i s t f o r a d d i t i o n of t h e r e q u i r e d uranium and p r o v i s i o n

of the required BTF3/UF,

ratio,

$ . . . . . . p 9

3.3.3.1

Preparation of initial fuel

Initial purification procedures for the DNSR present no f o r m i d a b l e problems.

Nuclear poisoms (e.g.,

boron, cadmium, or lanthanides) are not

common contaminants of the constituent r a w materials.

All the pertinent

cmpsunds contain at least small amounts of water, and all are r e a d i l y hyarsliyzea to oxides ana oxyfluorides at elevated temperatures.

he C O W -

pounds LiF and BeF2 generally contain a small quantity of sulfur as SUI-

fate ion.

Uranium tetrafluoride commonly contains small amonts of U 0 z p

UFg, and U82F2.

purification prseeaures63 ,*4 %85 uses to prepare materials f o r the

aircraft reactor experiment (ARE), the MSRE, and many laboratory and engineering experiments have treated the mixed materials at. high temperat u r e (usually at 600Â°C) with gaseous H ~ - H P mixtures and then with pure H 2

in equipment of nickel or copper. duce t h e 6T5*

and U6+ to

u4+,

me HF-HZ

treatment serves to (1) re-

( 2 ) reduce SU%fat@ ts sulfide and BIâ&#x201A;Ź3SlOVe it

HCP, and ( 4 ) ~ s n v e r tt h e oxides and oxyfluorides Final treatment with H2 serves to reduce PeF3 and FeF2 to

as H ~ S , ( 3 ) remove CP- as

to fluorides.

insoluble iron and to remove NiF2 that may have been produced during hydrofluorination, To date, all preparations have been performed in batch equipment but eontiwuous equipment has been partially developed. 86 8 87 F o r a DMSR, a s for the MSRE,*% purificatfsn of the bulk of t h e f u e l

would presumably b e conducted on L ~ F - B ~ F ~ - T ~ F L + mixtures -UP~ containing perhaps 85 to 90% sf the required WL, and on molten Li3UF7 to provide the additional uranium necessary to bring the fuel to the critical and opera t i n g Concentration.

Such a purification procedure can provide a sufficiently pure and eompletely homogeneous fuel material f o r initial operation of the reactor. 3.3.3.2

Contamination possibilities

Though the fuel material can be supplied and introduced into the reactor in sufficiently pure fom, contamination of the fuel is possible ESOre

SeQePal

SOUlllC@S e

Other reactor materials.

a%e

moderator graphite can contain a large

quantity of 602, C6, and H28 by virtue of its porosity and internal

surface.

Outgassing sf the moderator by pumping a t reduces p r e s s u r e and

e l e v a t e d t e m p e r a t u r e i s n e c e s s a r y and s u f f i c i e n t to p r e v e n t c o n t a m i n a t i o n

of t h e f u e l by oxide i o n from r e a c t i v e g a s e s from t h i s source. Oxide f i l m s on t h e s t r u c t u r a l metal can a l s o contaminate t h e f u e l by

o x i d e i o n , and, as d e s c r i b e d p r e v i o u s l y , t h e d i s s o l v e d ~ e 3 + $~ e 2 + $and N i 2 + can be r e s p o n s i b l e f o r subsequent metal c o r r ~ s i o n . I n o p e r a t i o n s f

the MSRE, the s y s t e m w a s flushed with an LiF-BeF2 m i x t u r e f o r c l e a n i n g a t

s t a r t - u p and a f t e r each shutdown b e f o r e i n t r o d u c t i o n of t h e f u e l mixture. This p r e c a u t i o n might be unnecessa%y, b u t i t d i d s u f f i c e t o keep oxide

contamination caused by s u r f a c e o x i d a t i o n of t h e metal to a m i n i m u m . A s m a l l (-1OO-ppm)

Of r e a c t i o n

c o n c e n t r a t i o n of ~ r 2 +i n the f u e l as a consequence

t h e metal w i t h t h e f u e l Cartnot b e aVOid@di. However, in the

absence of e x t r a n e o u s o x i d a n t s , t h e r e a c t i ~ ni s v e r y s l i g h t , and t h e presence

Cr2'

is CQIIIpbet@%y illlX3CuOuS.

Grow-in of the fission p r o d u c t s i s a l s o unavoidable, a s i s t h e pres-

ence of a r e l a t i v e l y small s t e a d y - s t a t e c o n c e n t r a t i o n of 3

Atmsspherfc contamination. oxygen is r e l a t i v e l y

* slow,

Reaction of t h e DMSK f u e l m i x t u r e w i t h

b u t r e a c t i o n with water vapor i s more r a p i d ,

F u r t h e r , contaminati~nof t h e f u e l with 40 t o 50 ppm ( b y w e i g h t ) sf oxide

i o n could r e s u l t i n p r e c i p i t a t i o n of a uranium-rich (UTB%)02 s o l i d

SO~U-

tion,

A l a r g e i n g r e s s of contaminant a i r would be r e ~ t a i r e dt o produce

40 ppm

Of 02-

i n the fuel, and the DMSR would be designed amd o p e r a t e d

as t o minimize t h e chances sf such contamination. d u r i n g much

O p e r a t i o n of PISRE

a fQuK'-yeaaP p e r i o d With Illany ShUtdOWTls and several

lrnimpor

r e p a i r o p e r a % i o n s s h o ~ e dno evidence sf a n i n c r e a s e i n oxide contamina-

t i o n level.22

~reatment

the i n i t i a l f u e l charge w i t h anhydrous H F - B ~

m i x t u r e d u r i n g its p r e p r a t t o n reataces the 02- c o n c e n t r a t i o n t o innocuous

Peve%s, and s i m i l a r treatment of contaminated f u e l would s e r v e t o r e ~ ~ v e the

o%-.

such t r e a t m e n t might never be r e q u i r e d , but i n the DMSR, s i m p l e

equipment should b e i n c l u d e d t h a t i s c a p a b l e of t r e a t m e n t t o ~ e m o v eo x i d e fOn should BnadVePtent

C o R t P A R i n a t i O n BCCUre

K Q W ~ V ~ P t, h e r e a c t o r metal a t h i g h t e m p e r a t u r e can r e a d i l y react w i t h oxygenB and t h e f u e l can react with the o x i d e s formed in this mannero

.... ...:.. wx

The o n l y secondary cool-

Contamination of f u e l by secondary c o o l a n t . a n t t h a t has been demonstrated i n a m o l t e n - s a l t

r e a c t o r i s L i 2 B e F 4 , which

i s prepared from 7%iF and p u r i f i e d through procedures d e s c r i b e d p r e v i o u s l y ~ Q I -t h e

fuel

i n the M S R E . ~ ~mis material i s n o t considered s u i t a b l e

f o r an MSBR o r a DMSR because i t i s expensive and its l i q u i d u s i s t o o high.

Many s u b s t i t u t e s have been considered, bur none have p r o p e r t i e s

t h a t a r e all n e a r the ideal.

a m i x t u r e of 8 mole

Bat b a l a n c e , t h e b e s t choice a p p e a r s t o be

X NaF and 92 mole 2

%9aBF4 (Refs. 19, 20, and 88).

These compounds are r e a d i l y a v a i l a b l e a t Isw c o s t .

The l i q ~ i d u s( s~e e~

F%g. 151% s t a b i l i t y toward gamma r a d i a t i o n i n the primary h e a t exchanger,9Q heat-transfer

p r o p e r t i e s $23$ 2 4 3 9 1 and compatibility47 948 w i t h modified

HasteEBoy-W a l l a p p e a r adequate. I n t e r m i x i n g of the f u e l and t h e secondary c o o I a n t s a l t s , a s caused

by leaks i n the p r i m a r y heat exchanger, would be an important c o n s i d e r a tion.

~ f a eMSBR design8 and p r e s u w a ~ ythe DMSR d e s i g n assured a s l i g h t l y

h i g h e r p r e s s u r e on the c o o l a n t s i d e so t h a t most l e a k s would be of cool-

ant into fuel.

Such a Leak, however small, should be recognized at. once

Fig, 15.

System NaF-NaBF4.

because of t h e marked r e a c t i v i t y l o s s caused by admission of boron i n t o the fuel, A small q u a n t i t y of NaF-NaBF4

added t o t h e DMSR f u e l would a l l o w

d i s s o c i a t i o n of t h e NaBF4 i n t o NaF and BP3.

The NaF would d i s s o l v e i n

t h e fuel and remain as a minor p a r a s i t i c neutron a b s o r b e r .

The BF3 i s

r e l a t i v e l y i n s s l i u b ~ ei n t h e f u e l 9 2 9 9 3 and w o p a ~ c be ~ r e a d i l y spargec~w i t h

the krypton and xenon i n t o t h e off-gas system,

A s u f f i c i e n t l y small con-

t i n u i n g leak could p o s s i b l y b e tolerated w i t h some impairment i n system performance.

Given t h a t t h e l e a k i n g t u b e could be plugged, i n f r e q u e n t

s m a l l leaks almost c e r t a i n l y ~ u l ndo t pose s a f e t y problems.

A d d i t i o n of

a s u f f i c i e n t l y l a r g e q u a n t i t y of WaF-MaBF4 c o u l d lead t o f o r m a t i o n of two immiscible l i q u i d phases.93

Such a l e a k (one c a p a b l e of adding a few t e n s

of p e r c e n t s o f c o o l a n t t o f u e l ) seems i n c r e d i b l e ; the p r e s e n c e of the large q u a n t i t y of boron should c e r t a i n l y p r e c l u d e r e a c t i v i t y a c c i d e n t s , b u t the f u e l would be r u i n e d .

~ e t u r n i n gt h e fuel m i x t u r e t o some secure

s i t e f o r r e c o v e r y would be n e c e s s a r y , and a most d i f f i c u l t c l e a n u p and r e p a i r of t h e r e a c t o r would be n e c e s s a r y , I f p o s s i b l e .

Small leaks of c o o l a n t i n t o the f u e l system p r o b a b l y pose no s a f e t y problems.

However, a d d i t i o n a l s t u d y of t h e mixing of t h e s e f l u i d s i n

r e a l i s t i c g e o m e t r i e s and i n flowing systems i s needed b e f o r e we c a n be c e r t a i n t h a t no p o t e n t i a l l y damaging s i t u a t i o n eouEd a r i s e as a consequence of a sudden major f a i l u r e of t h e h e a t exchanger. 20 The f l u o r o b o r a t e secondary c o o l a n t a p p a r e n t l y w i l l c o n t a i n m a l l

q u a n t i t i e s of oxygenated species and some species c o n t a i n i n g hydroxyl

ions.2o

These would be c a p a b l e of p r e c i p i t a t i n g oxides from the f u e l if

t h e c o o l a n t were m i x e d w i t h f u e l i n l a r g e amounts, b u t t h e e f f e c t s would be t r i v i a l Compared With o t h e r e f f e c t s noted p r e v i o u s l y .

These SUbStanceS

i n t h e secondary c o o l a n t , however, appear t o have a n o n t r i v i a l and benef i c i a l e f f e c t on DMSR p e r f o m a n c e . 2

This b e n e f i c i a l e f f e c t i s t h e appar-

e n t a b i l i t y of t h e secondary c o o l a n t t o scavenge tritium and c o n v e r t i t

t o a recoverable water-soluble f o m . As noted e a r l i e r , a DMSR must be expected t o g e n e r a t e about I G B q / s (2500 C i / d ) ~6 t r i t i m , and most of t h i s must be expected t o d i f f u s e

*However,

t h e o f f - g a s system would have t o be d e s i g n e d t o accomo-

d a t e t h e c ~ n s e q u e n c e sof BF3 admission.

... ... .. *.%

*w-

through t h e walls of t h e primary h e a t exchamger i n t o t h e coolant.

estimates9'+suggested that

Early

unless o t h e r mechanisms f o r tritium r e t e n t i o n

were p r s v f d e d , a s muck as 60% of t h e tritium generated would be lost through the coolant p i p i n g to t h e s t e m system, from which i t would be presumed t o e s c a p e t o t h e environment.

Such a l o s s r a t e t o t h e environ-

ment would be i n t o l e r a b l y high.

small-scale s t u d i e s ' s $ 9 6 suggested that oxide-bearing and p r o t o n a t e d

( e S g e pBF30H-I s p e c i e s were p r e s e n t i n t h e molten NaF-NaBP4 mixture proposed as t h e secondary c o o l a n t ; the h y p o t h e s i s was t h a t exchange r e a c t i o n s might o f f e r a mechanism f o r holdup of tritium i n t h i s

scale eXpePhents9B

mixture.^^

Smab%-

seemed t o show t h a t df2ueeriUltl diffused through

thin metal tube i n t o such m i x t u r e s was r e t a i n e d by t h e m e l t b u t that exchange w i t h OH- w a s n o t t h e r e s p o n s i b ~ emechanism. ~ h o ~ gthe h t r a p p i n g mechanism remains obscure, more r e c e n t t e s t s 1 0 have confirmed t h e a b i l i t y of WaF-NaBFb m i x t u r e s t o hold up the t r i t i m .

h e n g i n e e r f n g - s c a l e Poop, through which the salt c o u l d be pumped at 8.85 m 3 / s (850 gp), was used.

~zmisl o o p w a s arranged so t h a t tritium c o u l d

be i n t r o d u c e d by d i f f u s i o n through thin-wa%led t u b e s w i t h i n t h e s a l t ;

a l s o , t h e q u a n t i t i e s of tritium w i t h i n t h e salt, t h e q u a n t i t i e s removed i n t h e g a s f l o w above the f r e e s a l t s u r f a c e w i t h i n t h e pump bowl, and t h e q u a n t i t y d i f f u s i n g through t h e loop w a l l s i n t o t h e cooling a i r c o u l d be determined.

During s t e a d y - s t a t e o p e r a t i o n of t h i s d e v i c e i n two tests,

e a c h P a s t i n g about 68 d a y s , m a t e r i a l balance accounted f o r about 99% of t h e added tritium."

About 98% of t h e added tritium appeared i n t h e ef-

f l u e n t gas system of the pump, w i t h more t h a n 90% of t h i s i n a c h e m i c a l l y cmbfned ( w a t e r - s o l u b l e )

form.

The tritium w i t h i n the s a l t w a s essen-

t i a l l y aP% c h e m i c a l l y combined; the ratio of f r e e tritium to combined

tritium was less t h a n 1:4000.

E x t r a p o l a t i o n of t h e s e d a t a t o t h e MSBR

coolant system s u g g e s t s t h a t tritium Posses t o the MSBR steam g e n e r a t o r could be k e p t t o less t h a n -4 ~ ~ q (10 / s

c i / $ ) . 118

F u r t h e r s t u d i e s are c l e a r l y n e c e s s a r y ; once t h e mechanism i s establ i s h e d , t h e performance of t h e system might b e improved.

Means f o r re-

plenishment of the a c t i v e a g e n t must be established, and improved means

f o r r e c o v e r y and u l t i m a t e d i s p o s a l sf the tritium must be developed.

3-3.3.3

F u e l maintenance ol3tions and methods

The i n i t i a l f u e l c h a r g e for a INSR c a n be prepared i n a high s t a t e of p u r i t y and i n t r o d u c e d i n t o t h e r e a c t o r by minor variamts of t h e

methods* used f o r t h e MSRE4 and proposed f o r the MSBR.8

F o r a once-

through BMSR t h a t proposes no chemical repr~cessing t o remove f i s s i o n p r o d u c t s , t h e r e q u i r e d f u e l maintenance o l p e r a t i ~ n sare r e l a t i v e l y fewe They i n c l u d e (1) c o n t i n u ~ t ~removal s ( b y t h e s p a r g i n g and s t r i p p i n g sect i o n of t h e r e a c t o r ) o f f i s s i o n - p r o d u c t krypton and xenon, (2) a d d i t i o n

of 2 3 5 ~and '3%

to r e p l a c e that lost by burnup and t o keep t h e f u e l suf-

f i c i e n t l y d e n a t u r e d , and ( 3 ) i n s i t u p r ~ d u ~ t i oofn UF3 t o keep t h e redox p o t e n t i a l of t h e f u e l a t t h e d e s i r e d l e v e l ; t h e y probably a l s o i n c l u d e

( 4 ) removal o f i n a d v e r t e n t oxide contaminants from t h e f u e l ; i n a d d i t i o n , t h e y may i n c l u d e ( 5 ) a d d i t i o n of "4

t o r e p l a c e t h a t l o s t by t r a n m u t a -

tion o r stored with f u e l removed from t h e o p e r a t i n g c i r c u i t and ( 6 ) removal 05 a p o r t i o n of t h e i n s o l ~ b k en o b l e and seminoble f i s s i o n products. Each of t h e s e i s d i s c u s s e d b r i e f l y i n t h e f o l l o w i n g s e c t i o n s .

Continuous removal of f i s s i o n - p r o d u c t krypton and

XC~OEB.

Stripping

of k r y p t o n and xenon makes p o s s i b l e t h e i r c o n t i n u o u s removal from t h e rea c t o r c i r c u i t by t h e p u r e l y p h y s i c a l means of stripping w i t h helium.

For

the reference-design MSBR,* h e l i m flowing a t 0.005 m 3 / s (10 c f m ) w a s to b e i n j e c t e d c o n t i n u o u s l y i n t o and w i t h d r a m from f u e l - s a l t

c a r r y i n g a total of 0.35 m'/s

s i d e streams

o r about 10%of the t o t a l f u e l flow r a t e ,

g e n e r a l l y similar o p e r a t i o n should prove o p t i m a l f o r t h e DMSR.

Such

a s t r i p p i n g c i r c u i t would remove an a p p r e c i a b l e ( b u t not a m a j o r ) f r a c t i o n of t h e tritium and a small ( p e r h a p s v e r y small) f r a c t i o n of t h e n o b l e and seminoble fissiom p r o d u c t s as gas-bor~ne p a r t i c u l a t e s . s t r i p p e r would remove BF3 i f l e a k s s f s e c o d a r y c o o l a n t were t o occur.

In addition, t h e iRtO

the f u e l

None of these removals ( e x c e p t p o s s i b l y t h e l a s t ) appre-

c i a b l y a f f e c t the chemical behavior of t h e f u e l system. A d d i t i o n sf fissionable and f e r t i l e uranium.

u'"

and -18,400

*FWI

kg of 23%

Adding -4,470

kg of

d u r i n g t h e l i f e t i m e of t h e once-through DNSR

for the MSRE was prepared i n r e l a t i v e l y s m a l l ~atches.22s85 If DKSRs were t o b e of commercia% consequence, c o n t i n u o u s p u r i f i c a t i o n systems would c e r t a i n l y be d e v i s e d f o r i n i t i a l f u e l p r e p a r a t i o n .

77 4

w i l l a p p a r e n t l y b e n e c e s s a r y (see Table .l7Is assuming the f u e l volume CfpaRgeS from these a d d i t i o n s any f u e l t o s t o r a g e ,

O t h e r Causes do

T3Ot

require

3?@lTkOVa% Sf

Such additions, which are made over- t h e 30-year

l i f e t i m e , would comprise some 38,190 kg (96,332 moles) o f UP4 added at a n a v e r a g e o f 1,840 kg (3,320 rneles)/year.

~ u r i n go p e r a t i o n , 22 many on-stream a d d i t i o n s a s molten ~

fissionable natertal

~ were ~ u made P ~t o the MSRE, and t h i s method of a d d i t i o n can

o b v i o u s l y b e used as a c l e a n , c o n v e n i e n t way to add t h e uranium t o a DMSW, Using t h i s method of a d d i t i o n would require use of 2.89

kg) of ? L i F c o n t a i n i n g 2013 kg o f 7 L i .

IO5 moles (7.514

This r e p r e s e n t s about 6.8% of t h e

ki i n t h e o r i g i n a l f u e l i n v e n t o r y and would r e s u l t i n a p p r e c i a b l e vo~upne i n c r e a s e ( e s p e c i a l l y i f BeF2 were added p r o p o r t i o n a l l y ) i n the f u e l . t During t h e c o u r s e of r e a c t o r ope ratio^, removing some f u e l t u s t o r a g e w i t h i n t h e r e a c t o r complex would probably b e n e c e s s a r y i f t h i s a d d i t i o n procedure were used. Developing and d e m o n s t r a t i n g methods of a d d i t i o n o f s o l i d W 4 ( o r p r o p e r m i x t u r e s of UF4 plus UF3) s h o u l d be p o s s i b l e .

These w i l l b e in-

h e r e n t l y more complex (and r a d i o a c t i v e l y dirty), and s t a t i n g which s f t h e o p t i o n s would be p r e f e r r e d is n o t p r e s e n t l y p o s s i b l e . Mafntafnfng t h e d e s i r e d UF3/UF4 r a t i o . demonstrated t h a t i n s i t u p r o d u c t i o n of

Operation of t h e PlSREl

m 3 could

9s22

be accomplished r e a d i l y

and c o n v e n i e n t l y by p e r m i t t i n g t h e c i r c u l a t i n g f u e l to react in t h e pump bowl w i t h a rod of m e t a l l i c b e r y l P i m suspended i n a cage of Rastelloy-Ne

This t e c h n i q u e c o u l d be adapted f o r use i n a DMSR; b e r y % l i m r e d u c t i o n would b e d e s i r a b l e i f t h e f i s s i o n a b l e and f e r t i l e uranium a d d i t i o n s are t o be made as 7 ~ 1 ~ ~ 1 7 ~ ~

as. A d d i t i o n s would be made as d e p l e t e d m a t e r i a l o r a s material c o n t a i n i n g not more than 28% of 235U i n 23*U; t h e materials would be added as mixed f l u o r i d e s o r p o s s i b l y as W 4 p l u s LF3.

evelo loping and demonstrating a method of a d d i t i o n of a high-uraniumc o n t e n t l i q u i d w i t h c o n s i d e r a b l y less LiF than L I ~ U Fare ~ almost c e r t d d y goss Bbl e e

that event, adding BeF2 will be necessary t o preserve the L f F / BeFz r a t i o i n t h e f u e l a t approximately i t s initial value.

'In

Tt?e original charge of f u e l contains some 9 - 3 5 x IO4 moles of ura-

nium.

If all this were supplied as W 4 , about 1840 moles ( 1 6 , 5 4 kg) of

B e o would be required to reduce 5 % of it to ?IF3.”

While the fnitial

preparation procedure could be modified so t h a t some of the bTF3 would be present as the fuel was delivered, in s i t u production likely wouEd be

more convenient, The delivered f u e l c o u l d be made slightly deficient in BeF2 t o accommodate that generated by

m 3 production.

As indicated previously, the f i s s i s n process occurring with Wb, is During the 30-year reactor lifetime (22.5 f u l l -

s8gnificantl.y oxidizing. years asskuned)

pow€!%

topeso nearly 6.12

W i t h

x 104

70% of the

fiSSiQfnS

occukring in uranim i s o -

moles o f uranium w i l l have been fissioned.

the fissioning uranium fs 95% U F ~ ,as much as 5.8

x 104

I C ~

moles of U F ~might

be oxidized [at a generally u n i f o m rate of 7 moles ( - 2 - 1 kg) per fullpower day] during the reactor lifetime. 2.9

Its reduction would require some

I O b moles (261 kg) of metallic b e r y l P i m .

The B e F 2 produced in that

manner represents about 3.0% of t h a t p r e s e n t in the original f u e l charge. Some additional reduction of UF4 t o W 3 will be required if t h e fuel must b e treated to remove oxide ion. Accomplishing the reduction of UF4 to uF3 in situ w0ul.B certainly seem feasible By using metallic uranium in place of beryllium.

Should

the decision be made to add the fissionable and fertile uranium as UFks reduction performance by use of uranium would have the advantage o f not appreciably diluting the fuel. Removal of inadvertent oxide contamination, Treatment of complex molten f l u o r i d e s w i t h anhydrous HF-Ifz mixtures has been used

COIIUSIQE~~Y to

reduce the oxide concentration t o completely innocuous l e v e l s . 8 4 %85

real doubt exists that such treatment could be used if required for purification of DMSR f u e l mixtures.

However, there is little basis t o assess

the necessity of such purification.

Operation of the MSRE during a four-

year period with many shutdowns and several minor repair operations showed

no evidence of oxide contamination.

In early versions, equipment should

be included in which HF-PI2 mixtures and then H2 could be used to remove “ ~ nadditional 9 . 6 3 x 104 moles of uranium tetraflusride %sillbe added d u r i n g the reactor lifetime. Reduction of 5% of this will require 81%additional 2.41 x IO3 moles (21.7 kg) of beryllium.

y,...:.L

such contamination.

F o r a demonstration r e a c t o r , t h i s r e l a t i v e l y s i n p l e

equipment probably s h o u l d be sized to p e r n i t t r e a t m e n t of the f u e l on a

300-d c y c l e ; i f i t were pessimistically assmed t h a t 3 . 3 d would b e required t o

~ P Q C ~a ~b aSt c h ,

t h e equipment should be s i z e d t o a s s o m o d a t e

1%of t h e f u e l c h a r g e (-1 m3)0 Some f i s s i o n p r o d u c t s would be a f f e c t e d by t h i s t r e a t m e n t ; i o d i n e , i n p a r t i c u l a r , would be evolved and would have t o b e managed i n the o f f gas.

~ e l e n i mand t e l l u r i u m (if t h e y a r e s o l u b l e a s ~ e 2 - and ~

e i ~n the -

k i d a t i o n of pa49 t o pas+ could be

molten f u e l ) might a l s o evolve.

avoided by i n e l u s i o n of a few p e r c e n t of H2 w i t h t h e HF. However, o x i d a t i o n of a l a r g e fraction of t h e UF3 t o UF4 would res u l t unless t h e IIF-H2 m i x t u r e c o n t a i n e d so l a r g e a f r a c t i o n of M2 t h a t i t

would be r e l a t i v e l y i n e f f i c i e n t a t o x i d e removal.

Accordingly9 t o a l l o w

f o r a d d i t i o n a l ( b e r y l l i u m o r uranium) r e d u c t i o n of t h i s UF4 would be nece s s a r y t o m a i n t a i n the d e s i r e d ~ J F ~ / U rP a~t i o .

For example, i n t h e u n l i k e l y e v e n t t h a t the f u e l must be t r e a t e d f o r o x i d e removal each 1000 full-power clays, t h e i n v e n t o r y would r e q u i r e t r e a t m e n t 8.2 t i m e s d u r i n g t h e r e a c t o r l i f e t i m e .

The i n v e n t o r y of ura-

nium i s o t o p e s (see Table 9 ) increases r e g u % a r l y d u r i n g t h e reactor l i f e t i m e and may a v e r a g e 1.07

x 105

moles during the 3 0 y e a r s .

~f ( a s i s

n o t t r u e ) a l l t h e U F ~were o x i d i z e d each t i m e and i f 5% of t h e uranium i n v e n t o r y were to be reduced, some 2.2 x IO4 moles of B e Q would be required during the r e a c t o r l i f e t i m e .

nis, when added t o t h e 2 . 9 x 104

moles of b e r y l l i u m e s t i m a t e d p r e v i o u l s y t o be r e q u i r e d t o overcome t h e o x i d a t i v e e f f e c t of uranium f i s s i o n , would t o t a l some 5.1

BeF2 g e n e r a t e d o r n e a r 5.4%

x 104 moles of

of t h e BeF2 3n t h e o r i g i n a l feed.

This added

BeF2, though added a t a slowly i n c r e a s i n g rate d u r i n g r e a c t o r l i f e , is a good match f o r the 6 . 8 % of 7

~ needed i ~ to add the uranium as ~ i 3 ~ ~A 7 .

p e r f e c t match of LiF and BeFz a d d i t i o n s i s c e r t a i n l y n o t requ-ired; t h e maintenance processes b r i e f l y i n d i c a t e d above might p r o v i d e a s u f f i c i e n t l y good a d d i t i o n r a t e f o r L i F and BeF2. P o s s i b l e a d d i t i o n of thorium.

I f making a few a d d i t i o n s of t h o r i m

t o t h e r e a c t o r f u e l d u r i n g i t s l i f e t i m e i s n e c e s s a r y , t h e n adding i t as

a l i q u i d c o n t a i n i n g %iF

and TRF4 s h o u l d be p o s s i b l e .

b e a m e l t c o n t a i n i n g about 70 nole % L i F and 30 mole

A p o s s i b i l i t y would

X ThF4 m e l t i n g wear

600째C (see Fig.

11).

A l t e r n a t i v e l y , a procedure presumably c o u l d be de-

veloped f o r a d d i t i o n of s o l i d ThFbe P a r t i a l removal of n o b l e and seminoble metals.

The behavior of these

i n s o l u b l e f i s s i o n - p r o d u c t s p e c i e s , as i n d i c a t e d previously, is n o t underI f t h e y p r e c i p i t a t e a s a d h e r e n t deposits on t h e DMSR

stood i n d e t a i l .

h e a t exchanger, t h e y would c a u s e no p a r t i c u l a r l y d i f f i c u l t problems.

ever, should t h e y

f ~ l ^ t ao n l y

How-

l o o s e l y a d h e r e n t d e p o s i t s t h a t b r e a k away and

c i r c u l a t e w i t h the f u e l , t h e y would be r e s p o n s i b l e f o r a p p r e c i a b l e paras i t i c n e u t r o n captures.

If t h e s e s p e c i e s were t o d e p o s i t on t h e moderator

g r a p h i t e , t h e y would c o n s t i t u t e a n even worse n e u t r o ~ ~ is ci t u a t i o n . To the e x t e n t t h a t they c i r c u l a t e as p a r t i c u l a t e material i n t h e

fuel, i n s o i u b l e f i s s i o n - p r o d u c t species c o u l d probably be u s e f u ~ i yremoved by a small bypass f l o w through a r e l a t i v e l y s l m p l e H a s t e l l ~ y - ~ o o l f i l t e r system.

Presumably, s u c h a system would need t o have a r e a s o n a b l y

I s w p r e s s u r e d r o p and probably would need t o c o n s i s t of s e c t i o n s i n p a r a l -

le1

t h a t u n i t s whose c a p a c i t y was exhausted could be r e a s o n a b l y re-

placed o

3.3.3.4

Summaryp,c o n s t r a i n t s , and u n c e r t a i n t i e s

Very Ifke%y, a number of optkn-is f o r f u e l maintenance are a v a f l a b l e . Some of t h e s e have been demonstrated and o t h e r s couEd be made a v a i l a b l e if

there were good reasons why t h e y were needed. Several uncertainties also exist.

P r e s e n t % y , w e d o n o t h i o w whether

( 1 ) treatment t o remove i n a d v e r t e n t ~ o n t ~ ~ ~ n i nby a t oxide i ~ n w i l l be necess a r y $ ( 2 ) a d d i t i o n of uranium t o the DMSR f u e l will b e done by use of ' L ~ ~ U( 3F) ~the ~ o~x i d a t i v e e f f e c t of fission i s n e a r 1 oxidative equival e n t p e r mole of uranium f i s s i o n e d , o r ( 4 ) the removal of n o b l e and s e m f n o b l e metals from t h e DNSR f u e l i s n e c e s s a r y o r d e s i r a b l e . ShQu%d 'they prove d e s i r a b l e , a r e l a t i v e l y l a r g e number of: QptiOHlS

could b e made a v a i l a b l e .

A g r e a t amount of f u r t h e r o p t i m i z a t i o n of the

f u e l c y c l e f o r DNSR w i l l be r e q u i r e d b e f o r e we know which, if a n y , of t h e s e o p t i o n s are n e c e s s a r y

desirable.

3.4

R e a c t o r Materials

Although s p e c i a l , h i g h - q u a l i t y materials probably would Re used throughout i n t h e c o n s t r u c t i o n of a BMSR, most of them Could be o b t a i n e d

from cowmerical s o u r c e s t h a t r o u t i n e l y s u p p l y s u c h materials u s i n g curr e n t l y a v a i l a b l e technology.

Two notable e x c e p t i o n s t o t h i s g e n e r a l i z a -

t i o n are t h e S t r u c t u r a l a l l o y t h a t would have t o be used f o r components normally exposed t o molten s a l t and t h e g r a p h i t e f o r t h e reactor c o r e moderator and r e f l e c t o r .

Both of t h e s e materials would r e q u i r e s p e c i f i -

c a t i o n s p e c u l i a r t o t h e MSW system.

3.4.1

3.4.1.1

Structural allov

Requirements

The metallic s t r u c t u r a l m a t e r i a l used i n c o n s t r u c t i n g t h e primary c i r c u i t of a m o l t e n - s a l t 700째C.

r e a c t o r w i l l s p e r a t e a t t e m p e r a t u r e s up t o about

The i n s i d e of t h e c i r c u i t w i l l be exposed t o s a l t t h a t c o n t a i n s

fission p r o d u c t s and w i l l r e c e i v e a m a x i m u m thermal f l u e n c e of about 6

neutrons/m2 over t h e o p e r a t i n g ldfetime of about 30 y e a r s .

This

f l u e n c e w i l l cause some e m b r i t t l e m e n t because of helium formed by t r a n s m u t a t i o n b u t w i l l n o t cause s w e l l i n g such a s i s noted a t h i g h e r f a s t f l u -

ences.

The o u t s i d e of t h e primary c i r c u i t w i l l be exposed t o n i t r o g e n

t h a t c o n t a i n s s u f f i c i e n t ajir from i n l e a k a g e t o make i t o x i d i z i n g t o t h e metal.

Thus, the m e t a l must (1) have moderate o x i d a t i o n r e s i s t a n c e , ( 2 )

resist c o r r o s i o n by t h e s a l t , and ( 3 ) resist s e v e r e e m b r i t t l e m e n t by thermal n e u t r o n s . I n t h e secondary c i r c u i t , t h e metal w i l l be exposed t o t h e c o o l a n t

s a l t under much t h e same c o n d i t i o n s d e s c r i b e d for t h e primary c i r c u i t .

The main d i f f e r e n c e s w i l l be t h e l a c k of f i s s i o n p r o d u c t s and uranium i n t h e c o o l a n t s a l t and much lower n e u t r o n f l u e n c e s .

T h i s material must have

moderate o x i d a t i o n r e s i s t a n c e and m u s t resist c o r r o s i o n by a s a l t n o t eont a i n i n g f i s s i o n p r o d u c t s o r uranium. The primary and secondary c i r c u i t s i n v o l v e numerous s t r u c t u r a l shapes ranging from s e v e r a l c e n t i m e t e r s t h i c k t o t u b i n g having w a l l t h i c k n e s s e s

of o n l y a millimeter o r so.

These shapes must be f a b r i c a t e d and j o i n e d

( p r i m a r i l y by welding) i n t o an i n t e g r a l e n g i n e e r i n g s t r u c t u r e .

The s t r u c -

t u r e must b e designed and b u i l t by t e c h n i q u e s approved by t h e American S o c i e t y ~f Mechanical Engineers (ASME) Boiler and P r e s s u r e Vessel Code.

3.4.1.2 E a r l y materials s t u d i e s l e d t o t h e development of a nickel-base

Poy, HastePloy-N,

~ Q use P

w i t h f l u o r i d e salts.

%E-

A s shown i n Table 28, t h e

alloy contained 16% molybdenum f o r strengthening and ehromium s u f f i c i e n t t o impart moderate o x i d a t i o n r e s i s t a n c e i n a i r b u t n o t enough t o Bead t o h i g h c o r r o s i o n r a t e s in salt.

This alloy w a s the s o l e s t r u c t u r a l material

Table 28. Chemical csmpositfon of MastelBoy-N Content oâ&#x201A;Ź a l l o y

(wt X f

Element

Standard

Nickel Molybdenm Chr om 1m Iron Manganese Silicon Phosphorus Sulfur Boron Titanium Niobium

Modified

Base 1Ej-18 6-8

Base 1 1-1 3

5 1 1

0. I b 0.16-4.25 0.1 0.01 0.01 0.801

O.OB% 0. Q20 0.01.

-b

1-2

"single values a r e maximum amounts allowed. The a c t u a l conc e n t r a t i o n s of these elements i n an a l l o y c a n be much lower, bThese elements art? n o t f e l t t o b e v e r y important. Alloys are now being purchased w i t h t h e small c o n c e n t r a t i o n s speeified, b u t t h e s p e c i f i c a t i o n may b e changed i n t h e f u t u r e t o allow a higher concentration.

....., ..’ -. ..WY

used i n t h e MSRE amd c o n t r i b u t e d s i g n i f i c a n t l y t o the s u c c e s s sf the ex-

two problems were noted w i t h Hastelloy-N which needed

HQW~WX,

periment.

f u r t h e r a t t e n t i o n before more advanced r e a c t o r s could be builto

First,

Haste1loy-N w a s found t o be embrittled by helium produced d i r e c t l y from

traces of ’QB and i n d i r e c t l y from nickel by a two-step reactiom.

nis

t y p e of r a d i a t i o n embrittlement i s common t o most iron- and nickel-base The second problem a r o s e from the f i s s i o n - p r o d u c t

alloys.

tellurium d i f -

f u s i n g a short d i s t a n c e i n t o t h e metal along t h e grain Bs~undariesand em-

brittling t h e boundaries. Considerable s u c c e s s w a s encountered i n modifying t h e composition of

Bastelloy-N ts s b t a f n better r e s i s t a n c e t o embrittlement by i r r a d i a t i o n . The key f a c t o r was to modify the c a r b i d e p r e c i p i t a t e from t h e c o a r s e type found

itl

Standard Bastelloy-N t o a Very fine type.

‘Ehe p%eSeKXe Of 16%

wolybdemaurn and 0.5% s i l i c o n l e d t o t h e formation of a c o a r s e c a r b i d e t h a t Reduction s f t h e m~lybdenumc o n c e n t r a t i o n t o 12X and

had l i t t l e b e n e f i t . the

SiliCSrt

c o n t e n t t o 0.1% and a d d i t i o n of

r e a c t i v e c a r b i d e former such

a s t i t a n i u m o r niobium l e d to t h e formation of a f i n e c a r b i d e p r e c i p i t a t e and an a l l o y with good r e s i s t a n c e t o e m b r i t t l e m e n t by helium.

Consider-

a b l e progress was macle in t h e scale-up of an a l l o y c o n t a i n i n g 2% t i t a n i u m , b u t t h i s alloy does n o t have s u f f i c i e n t r e s i s t a n c e t o intergranular crack-

ing by t e l l u r i u m .

An a l l o y c o n t a i n i n g B t o 2% niobium was fasted t o b e

v e r y r e s i s t a n t t o c r a c k i n g by t e l l u r i u m and was produced i n small commer-

c i a %IIleltS.

Table 28.

The cOmpoSit%Qn O f the niobfWll-modif$ed

a l l o y iS

ShQwsl

This alloy m a j n t a i n s good d u c t i l i t y up t o t h e 46-ppm max3mm

helium c o n t e n t a n t i c i p a t e d i n t h e w a l l of a m o l t e n - s a l t

reactor vessel.

I n s t u d y i n g t h e t e l l u r i u m e m b r i t t l e m e n t problem, c o n s i d e r a b l e e f f o r t

w a s s p e n t i n s e e k i n g b e t t e r methods of exposing t e s t specimens t o t e l l u rium. about

1x1 t h e MSRE, t h e f l u x of t h e t e l l u r i u m atoms r e a c h i n g t h e metal was

atoms m”2

s-’,

high-performance b r e e d e r .

and t h i s v a l u e would be 10n4 atoms m-2

s-l f o r a

Even t h e v a l u e f o r a high-performance b r e e d e r

i s v e r y mall from t h e e x p e r i m e n t a l s t a n d p o i n t . would r e q u i r e t h a t a t o t a l of 7.6

x 10-6

For example, t h i s f l u x

g of t e l l u r i u m be t r a n s f e r r e d t o

a sample having a s u r f a c e area of IO cm2 i n 1000 h.

Electrochemical

probes were immersed d i r e c t l y i n salt melts k n ~ ~t m o contain t e l l u r i u m ,

84 and t h e r e w a s never any evidence sf a soEuble t e l l u r i d e s p e c i e s ,

However,

c o n s i d e r a b l e evidence e x i s t e d t h a t t e l l u r i u m "moved" through s a l t from one p o i n t to a n o t h e r i n a salt system.

The hypothesis w a s t h a t t h e t e l l u r i u m

actually moved as a low-pressure p u r e metall vapor and n o t as a r e a c t e d species.

'Ehe most r e p r e s e n t a t i v e experimental system dle~el~iped f o r ex-

posing metal specimens eo t e l % u r i m i n v o % ~ e d suspendling the s p e ~ i m e nin ~ a skirred

vessel of salt: w i t h granu%@S

Cr3Te4 and CPgT@6 l y i n g at t h e

A v e r y low p a r t i a l p r e s s u r e of t e P E u r i m w a s in equi-

bottom of the s a l t .

l i b r i m with the Ck3T@4 and C ~ g T e 6which ~ resulted in BaStePEOy-M speci-

mens w i t h crack s e v e r i t i e s s i m i l a r t o t h o s e noted i n samples from t h e MSRE,

NUEIBRH.LS

s a m p l e s were exposed t o s a l t that c o n t a i n e d t e l l u r i u m , and

t h e most important f i n d i n g w a s that modified Hastelloy-N c ~ n t a i ~ ~ i1nt go 2% niobiW had good reSistaXlC@ t o eTRb8-i&e%ementby h e Sekfe.5

Qf experiments

Was

tt?llUgf-UlR

(Fig. 16).

Tun t o inVeStfgatf2 t h e effects

oxidation state of tellurium-containing

the

salt on t h e temdency f o r c r a c k s

t o be f o ~ ~ ~me d ,s u p p o s i t i o n being examined was that the s a l t might be made reducing enough to t i e up t h e t e l l u r i u m i n some f n n o c u o ~metal ~ complex.

The s a l t was made more o x i d i z i n g by a d d i n g WiF2 and more reducing

by adding e l e m e n t a l b e r y b l i m .

The experiment had e l e c t r o c h e m i c a l probes

f o r determining t h e r a t i o of urarnim in t h e +4 s t a t e ( U P q ) t o that im t h e 4-3 s t a t e (UF3) a s a n i n d i c a t o r of t h e sxfdation s t a t e of t h e salt,

Ten-

s i l e s p e ~ i m e n sof s t a n d a r d ~ a s t e i l o y -were ~ suspended in the s a l t f o r

about 260 k at 700째C.

The o x i d a t i o n s t a t e of the salt w a s s t a b i l i z e d , and

t h e specimens were i n s e r t e d so t h a t each s e t of specimens w a s exposed t o

one c o n d i t i o n .

A f t e r exposurep t h e specimens were s t r a i n e d t o f a i l u r e and

were examined m e t a l l o g r a p h i c a l l y t o determine t h e extent of c r a c k i n g e

'%he

results of measurements a t s e v e r a l o x i d a t i o n s t a t e s are shown i n Fig. 17. ~t ~ 4 + / ~ 3 ratios + sf

QO o r less, very l i t t l e c r a c k i n g o c c u r r e d , and a t

r a t i o s above 80, t h e c r a c k i n g was e x t e n s i v e . encouragement. that

These o b s e r v a t i o n s o f f e r

r e a c t o r could be o p e r a t e d i n a chemical regime where

the tellurium would n o t be embrittling even t o s t a n d a r d Hastelloy-N.

least l e 6 % of t h e uranium would need t o be i n t h e +3 oxidation s t a t e (UF3js amd t h i s e o n d i t i ~ nseems q u i t e r e a s o n a b l e from chemical and practical c o n s i d e r a t i o n s .

€9

QRPIL-DWG

77-922 D

Nb CONTENT ( o / o )

Fig. 16. V a r i a t i o n s of s e v e r i t y of c r a c k i n g w i t h Nb c o n t e n t . Samples were exposed f o r i n d i c a t e d times to s a l t c o n t a i n i n g C r 3 T e 4 and C r g T e g a t 900°C.

..... i... ....., ... Y d

86 ORNL-DWG 77-4680A

CRACKING 600 PARA MET E W [FREQUENCY ~cm-') x AVG. DEPTH Eptml]

0 (8 26 46 7 0 '168 200 400 SA LT 0X ID AT i0 N POTE N T IA k [U (IPB I/' U ( I I 1

Cracking behavior s f H a s t e E 1 ~ g r - N exposed 260 h a t 900째C P i g , 17, t o MSBR f u e l s a l t c o n t a i n i n g Cr3Te4 and Cr5Teg.

Presently, t h e modified a l l o y composition shorn in Table 28 is favored.

Considerable progress had been made i n e s t a b l i s h i n g t e s t methods

f o r e v a l u a t i n g a rnaterialPs r e s i s t a n c e t o m b r i t t l m e n t by teEPurfum.

Modified Hastelloy-N c o n t a i n i n g from 1 t o 2% r t i g l b i ~ iwas ~ found t o o f f e r improved r e s i s t a n c e t o e m b r i t t l m e n t by t e l l u r i u m , beat t h e t e s t condi-

t i o n s were not s u f f i c i e n t l y long o r d i v e r s i f i e d to show that t h e a l l o y t o t a l l y r e s i s t e d mbrittLment,

h e irradiation experiment showed t h a t

t h e niobium-modified alloy o f f e r e d adequate r e s i s t a n c e t o i r r a d i a t i o n e m b r i t t l a e n t , b u t more d e t a i l e d tests are needed.

Several m a l l m e l t s

c o n t a i n i n g up t o 4 . 4 % niobium were found t o fabricate and weld w e l l , so p r o d u c t s containllng 1 t o 2% niobium likely can be produced w i t h a m i n i m u m

of scale-up d i f f i c u l t i e s .

3.4.1.3

Uncertainties

Although no basic scale-up grobl@as are a n t i c i p a t e d w i t h t h e niobiummodified alloy, s e v e r a l l a r g e h e a t s must be melted and processed into s t r u c t u r a l shapes t o show t h a t t h e a l l ~ ycan be p r o d ~ e e de o m e r ~ i a l l y ~

A f u r t h e r need exists for longer exposure of tkfs a l l o y t o i r r a d i a t i o n

and t o s a l t c o n t a i n i n g t e l l u r i u m t o show t h a t i t w i l l resist e m b r i t t l e ment by t h e s e two processes aver long p e r i o d s of t i m e .

i c a l property t e s t s must be r u n

Numerous mechan-

t h e new alloy to d e v e l o p the d a t a

needed f o r ASME B o i l e r and P r e s s u r e Vessel Code a p p r ~ v a lof t h e a l l o y and t o e s t a b l i s h adequate d e s i g n methods.

3.4.2 3.4.2.1

Moderator

hquiKementS

The graphite i n a s i n g l e - f l u i d molten-salt

reactor serves no s t r u c -

tural purpose o t h e r t h a n t o d e f i n e t h e flow p a t t e r n s of t h e s a l t and, sf c o u r s e , t o s u p p o r t i t s o m weight.

The requirements ~n the material are

d i c t a t e d n o s t s t r o n g l y by n u c l e a r c o n s i d e r a t i o n s , t h a t i s , s t a b i l i t y of t h e material a g a i n s t radiation-induced

d i s t ~ r t i o nand n o n p e n e t r a b i l i t y by

P r a c t i c a l limitationas of meeting these re-

t h e f u e l - b e a r i n g molten salt.

quirements impose c o n d i t i o n s on t h e c o r e d e s i g n -

specifically t h e neces-

s i t y to l i m i t t h e c r o s s - s e c t i o n a l area of t h e g r a p h i t e prisns.

The re-

quirements of p u r i t y and impermeability ts salt are e a s i l y m e t by several h f g h - q u a l i t y f i n e - g r a i n e d graphites, and t h e main problems a r i s e from the requirement of s t a b i l i t y a g a i n s t radiation-induced

3.4.2.2.

distortion.

S t a t u s sf development

The dimensional changes ~f graphite d u r i n g i r r a d i a t i o n have been

s t u d i e d f o r a number of years.

The dimensional changes largely depend

on t h e d e g r e e of c r y s t a l l i n e i s o t r o p y , b u t t h e volume changes f a l l i n t o

a rathear c o n s i s t e n t p a t t e r n . tion

QCCUPS

A s shorn i n Fig. 18, a p e r i o d of densffica-

f i r s t d u r i n g which t h e vslurae d e c r e a s e s , and a p e r i o d of s w e l l -

ing t h e n o c c u r s i n which t h e vo%me i n c r e a s e s .

The first p e r i o d i s of

concern only because of t h e dimensional changes t h a t t a k e p l a c e , and t h e second period i s sf concern because of t h e dimensional change and t h e

formation of c r a c k s .

The f o r m a t i o n of c r a c k s would e v e n t u a l l y a l l o w s a l t

t o penetrate the graphite.

Data shown i n Fig. 18 are f o r 7%5"C, and t h e

damage rate increases with increasing temperature.

Thus, the g r a p h i t e

section s i z e should be k e p t s m a l l enough t o prevent temperatures i n the g r a p h i t e f r m g r e a t l y exceeding t h o s e i n t h e s a l t . ... ... w ...y

-----? -IIA SM-I AT-d-! A

25 I

With the d i f f e r e n t ~ B j e c t i v e sof n s n p r o l i f e r a t i n g MSRs, the require-

ments for the g r a p h i t e have diminished from those of t h e high-performance

First, t h e peak neutron f lux in the c o r e can b e reduced t o

breeder.

Bevels such t h a t the g r a p h i t e w i l l last f o r t h e l i f e t i m e of t h e r e a c t o r ~ e e o n d ,b o t h t h e low power d e n s i t y ana the pow r a t e of xenon mass

plant,

t r a n s f e r to t h e g r a p h i t e tend t o l i m i t t h e xenon p o i ~ o n i n ge f f e c t i n tkPs reactor

t h a t sealing t h e graphite may n o t b e necessary.

me lessened

gas p e r m e a b i l i t y r e q u i r e m e n t s also mean that t h e g r a p h i t e can be i r r a d i a t e d t o h i g h e r fluences ( F i g s . 18 and 1 9 ) .

l i f e t i m e c r i t e r i o n adopted

f o r t h e b r e e d e r was a damage fluenee of about 3 x 1026 neutrons/&.

Fig. 19,

nits

20 30 FLUENCE [neutrons/crn* ( E > 5 0keVB]

'Volume changes f o r m o n o l i t h i c graphites i r r a d i a t e d a t 7 1 5 " 6 ,

w a s e s t i m a t e d to be t h e fluence a t which t h e g r a p h i t e s t r u c t u r e would cont a i n s u f f i c i e n t c r a c k s to be permeable t o xenone

Experience h a s shown

t h a t , even a t v o l m e changes of about lox, t h e g r a p h i t e i s n o t cracked brat i s u n i f s m l y dilated.

For nonproliferating d e v i c e s , xenon perme-

a b i l i t y will n o t be of as much eoncern, and t h e limit probably w i l l be e s t a b l i s h e d by the

trusion.

~ O I - ~ T E ~ ~ ~of Q I Kc r a c k s

s u f f i c i e n t l y l a r g e f o r s a l t in-

GLCC" K-364 g r a p h i t e l i k e l y c o u l d be used t o 3 x 1 0 ' ~

n e u t r o ~ s / m 2 , and improved graphites w i t h a l i m i t of 4 x 10'6

neutrons/m2

c o u l d be developed. The s p e e i f i c performance requirements f o r g r a p h i t e s u i t a b l e f o r the r e a c t o r d e s i g n p r e s e n t e d i n t h i s r e p o r t a r e a lifetime fluence c a p a b i l i t y of 2 . 7 x BOz6 neutrons/rn2 (E

50 keV) a t a peak t e m p e r a t u r e of ?'50Q@,

Most probably, e x i s t i n g cmmereial g r a p h i t e s will s a t i s f y t h % s meed. 3.4,2.3

Uncertainties

Although existing commercial graphites l i k e l y w i l l meet t h e needs of the presemt d e s i g n , g r a p h i t e samples having t h e same cross s e c t i o n as t h e r e f e r e n e e - d e s i g n moderator elements need t o be i r r a d i a t e d .

These

t e s t s need t o be kt%n t o t h e dâ&#x201A;Ź?stPUction O f t h e graphite t o detePnaiRe the

p o i n t a t which t h e g r a p h i t e a c t u a l l y heals. t h e p r e s e n t concept.

This w i l l d e f i n e f a i l u r e i n

P h y s i c a l p r o p e r t i e s , p a r t i c u l a r l y thermal esndue-

t i v i t y , need t o be measured as a f u n c t i o n of fluenee. A %onger-range e f f o r t t o develop improved g r a p h i t e f o r f u t u r e re-

a c t o r s Should be i'ilitiatecf.

E a r l y e f f o r t s Show prOlEise t h a t graphitf2.s

w i t h improved dimensional s t a b i l i t y can be developed,

3.5

Safety Considerations

The main f e a t u r e of t h e DMSR which sets i t a p a r t from s o l i d - f u e l re-

a c t o r t y p e s i s that t h e nuclear f u e l i s i n f l u i d form ( m o l t e n f l u o r i d e

salt) and i s c i r c u l a t e d throughout the primary c o o l a n t system, becoming c r i t i c a l only i n t h e graphite-moderated core.

P o s s i b l e problems and en-

g i n e e r e d s a f e t y features a s s o c i a t e d w i t h t h i s t y p e of r e a c t o r w i l l be

*Great

Lakes Carbon Company.

91 t ... .;$-

q u i t e d i f f e r e n t from those of t h e p r e s e n t EWR and liquid-metal

fast

A d e t a i l e d s a f e t y a n a l y s i s of the DMSK

b r e e d e r r e a c t o r QLMFBR) d e s i g n s .

must await t h e r e s u l t s of a r e s e a r c h and development C U D , > program; howe v e r , i d e n t i f y i n g p o s s i b l e g e n e r i c problem areas and some of t h e advant a g e s and d i s a d v a n t a g e s of t h i s concept i s a l r e a d y p o s s i b l e .

In t h e DMSW, the primary system f l u i d serves t h e d u a l r o l e of bein t h e m e d i m i n which h e a t i s g e n e r a t e d w i t h i n the r e a c t o r c o r e and t h e medium t h a t t r a n s f e r s h e a t from t h e core t o t h e primary h e a t exchangers. Thus, t h e e n t i r e primary system w i l l be s u b j e c t t o b o t h high t e m p e r a t u r e s (900째C a t t h e core e x i t ) and h i g h l e v e l s of r a d i a t i o n by a f l u i d c o n t a i n -

ing most of t h e daughter p r o d u c t s of the f i s s i o n process.

Because of t h e

low f u e l - s a l t vapor p r e s s u r e s however, t h e primary system d e s i g n p r e s s u r e W i l l

b@ IOW, -3s i n an LMPBRe

%XI

t@KUIS O f

level

ConfinelTlent, t h e e n t i r e

r e a c t o r primary s y s t e n i s analogous t o t h e f u e l c l a d d i reactor.

i n a s~%id-fuel

Although much l a r g e r , i t w i l l n o t b e s u b j e c t t o the r a p i d t h e r -

mal t r a n s i e n t s ( w i t h m e l t i n g ) a s s o c i a t e d w i t h LWR and LMFBR a c c i d e n t seenarios.

Two a d d i t i o n a l l e v e l s of confinement w i l l be provided in t h e

BMSW, i n accord w i t h present p ~ a e t i c e , Note t h a t t h e once-through DMSR

concept has safety advantages

Q V ~ Pt h e

break-even DMSR because a l a r g e

and complex p a r t of t h e primary containment

plant

- the

chemical r e p r o c e s s i n g

i s s u b s t a n t i a l l y reduced and because less r a d i o a c t i v e material i s

r o u t i n e l y removed from containment.

The problem of developing a r e a c t o r

primary system t h a t w i l l be r e l i a b l e , m a i n t a i n a b l e ( u n d e r remote condit i o n s ) , i n s p e c t a b l e , and s t r u c t u r a l l y sound over t h e p l a n t ' s SO-year l i f e -

t i m e w i l l probably be t h e key f a c t o r i n demonstrating u l t i m a t e s a f e t y and I i c e n s a b i lf t y

The breach of t h e r e a c t o r primary system b ~ ~ d a r y vr e, s u l t i n g i n a s p i l l of h i g h l y r a d i o a c t i v e s a l t i n t o t h e primary containment, w i l l proba b l y provide t h e design-basis accident,

The analogous event i n a s o l i d -

fuel r e a c t o r would be major c l a d d i n g f a i l u r e .

P o s s i b l e i n i t i a t o r s o% t h i s

accidemt i n c l u d e p i p e f a i l u r e , missiles, and p r e s s u r e o r temperature t r a n -

sients i n t h e primary s a l t system.

F a i l u r e of the boundary between t h e

p r h a r y and secondary s a l t i n t h e primary h e a t exchangers could be espec i a l l y damaging.

In t h e e v e n t of salt s p i l l , a p o s s i b l y redundant

sys-

t e m of d r a i n s would be a c t i v a t e d t o channel t h e salt t o t h e c o n t i n u o u s l y

92 cooled d r a i n tank.

The system primary containment, which i s d e f i n e d as

t h e s e t s f h e r m e t i c a l l y s e a l e d c o n c r e t e - s h i e l d e d equipment c e l l s , would probably n o t be t h r e a t e n e d by such a s p i l l , b u t c l e a n u p o p e r a t i o n s would be d 5 f f i c u l t e A unique s a f e t y feature of t h e DMSR i s that, under a c c i d e n t shutdown

c o n d i t i o n s , t h e f u e l material would be l e d t o t h e emergency core c o o l i n g system gECCS> ( r e p r e s e n t e d by d r a i n t a n k c o o l i n g ) r a t h e r than v i c e v e r s a . The r e a c t o r and containment must be designed

t h a t the decay-heated f u e l

s a l t r e a c h e s t h e d r a i n tamk under any c r e d i b l e a c c i d e n t c o n d i t i o n s .

any case, t h e decay h e a t i s a s s o c i a t e d w i t h a v e r y l a r g e mass of f u e l s a l t

so t h a t melt-through ( o r â&#x20AC;&#x153;China Syndromeâ&#x20AC;?) i s a p p a r e n t l y n o t a problem.

The s a f e t y philosophy f o r a c c i d e n t s i n v o l v i n g t h e r e a c t o r c o r e i s very d i f f e r e n t f o r f l u i d - f u e l e d t h a n f o r s o l i d - f u e l e d r e a c t ~ r sbecause t h e h e a t s o u r c e i s mainly i n t h e l i q u i d - f u e l

s a l t and n o t i n a s o l i d ,

which r e q u i r e s c s n t i n u o u s c o o l i n g t o avoid m e l t i n g .

An LMFBR, f o r exam-

p l e , has a l a r g e amount of s t o r e d energy (which must be removed under any a c c i d e n t c o n d i t i o n s ) i n t h e f u e l p i n s .

D P y ~ u t ,which means immediate

meltdown i n an LMFBR, would n o t be nearly a s s e v e r e i n t h e DMSR because t h e h e a t s o u r c e i s removed along d t h t h e c o o l i n g c a p a b i l i t y .

First-order

a n a l y s i s has shsm t h a t a flow blockage of a c e n t r a l c o o l a n t channel of t h e r e f e r e n c e DKSR which r e d u c e s t h e flow t o less t h a n -20% of nominal w i l l p r o b a b l y r e s u l t i n l o c a l v o i d i n g of t h a t channel.

This w a s n o t t r u e

o f t h e o l d MSBR design8 because t h e c h a n n e l s were more s t r o n g l y t h e m a l l y coupled.

Whether the s a f e t y i m p l i c a t i o n s of t h i s w i l l l e a d t o modifica-

t i o n s of t h e DMSR r e f e r e n c e d e s i g n must be shorn by f u t u r e s a f e t y analysis studies.

Under any o f f - n s m a l

c o n d i t i o n s , t h e f u e l s a l t will be chan-

n e l e d t o t h e d r a i n t a n k , which m u s t have r e l i a b l e s y s t e m s f o r decay h e a t removal.

No c r e d i b l e means e x i s t s f o r a c h i e v i n g r e c r f t i c a l i t y once t h e

fuel s a l t has l e f t t h e graphite-moaeratea c o r e s

There are no s i g n i f i c a n t d i f f e r e n c e s in t h e e n v i r o m e n t a l e f f e c t s of

r o u t i n e o p e r a t i o n s between an MSl? and r e a c t o r s p r e s e n t l y i n commercial operation.

No gaseous o r l i q u i d r a d i o a c t i v e e f f l u e n t d i s c h a r g e o c c u r s

during noma1 operation.

Minor amounts of such e f f l u e n t s way r e s u l t from

maintenance o p e r a t i o n s i n v o l v i n g opening t h e primary system. The MSR ( a l o n g w i t h t h e HTGR and t h e LMFBR) i s i n t h e c l a s s of rea c t o r s which o p e r a t e s a t h i g h t e m p e r a t u r e s and h i g h thermal e f f i c i e n c i e s about 4QX compared w i t h about 32% f o r LWRs.

For t h e same e l e c t r i c a l ea-

p a c i t y , t h e s e more e f f i c i e n t r e a c t o r s r e j e c t about 40% less b e a t t o t h e environment.

'%%lis can reduce i m p a c t s such a s consumptive use of water re-

Sources, atmospheric e f f e c t s , and e f f e c t s on a q u a t i c l i f e .

In t h e r e f e r e n c e DMSR c o n c e p t , n e i t h e r t h e n u c l e a r f u e l nor t h e f i s s i o n p r o d u c t s ( e x c e p t f o r t h e v s l a t i l e s , i n c l u d i n g xenon) are removed from t h e primary system d u r i n g t h e r e a c t o r l i f e t i m e . environmental problem of p r e s e n t day LWWs:

This e l i m i n a t e s a major

f r e q u e n t t r a n s p o r t a t i o n of

h i g h l y r a d i o a c t i v e s p e n t f u e l from t h e r e a c t o r s i t e t o t h e reprocessing! storage f a c i l i t y .

Most r a d i o a c t i v e m a t e r i a l remains w i t h i n t h e BMSW p r i -

mary containment f o r t h e 30-year r e a c t o r l i f e t i m e but must be d e a l t w i t h

a t end-of-life.

Uranium, E i t h i m , and p o s s i b l y o t h e r v a l u a b l e elements

w i l l probably b e recovered for r e u s e , b u t t h e remainder, which c o n t a i n s

t h e a c t i n i d e s americium and curium ( n o t found i n s i g n i f i c a n t amounts i n s p e n t EWW f u e l ) , w i l l have t o be d i s p o s e d o f .

Decommissioning t h e p l a n t

way b e more d i f f i c u l t t h a n f o r a n EWR because t h e entire primary c i r c u i t w i l l be i n t e n s e l y r a d i o a c t i v e . A l a r g e amount of t r i t i u m is g e n e r a t e d i n MSRs as a r e s u l t of n e u t r o n

r e a c t i o n s w i t h the l i t h i u m i n t h e f u e l s a l t .

Tritium is

to diffuse

through metal walls such as heat-exchanger t u b e s , t h u s p r o v i d i n g a potent i a l r o u t e f o r t r a n s p o r t of gaseous trittum through t h e secondary c o o l a n t

l o o p t o t h e steam g e n e r a t o r s .

Recent experiments have shown t h a t tritium

i s o x i d i z e d i n t h e secondary c o o l a n t (sodium f l u o r o b o r a t e ) , which b l o c k s f u r t h e r t r a n s p o r t of tritium.

The release of t r i t i u m from MSRs t o the en-

vironment i s e s t i m a t e d t o be no g r e a t e r t h a n from LWRs and i s w e l l w i t h i n

NRC g u i d e l i n e s . A power economy i n which t h e MSR p l a y s an important r o l e would re-

q u i r e l a r g e q u a n t i t i e s of l i t h i u m , BeryPlium, f l u o r i n e ( f o r t h e f u e l - s a l t m i x t u r e ) , n i c k e l (which comprises 78% of t h e Hastelloy-N), ( n a o d e ~ ~ telements). o~

and g r a p h i t e

The environmental. e f f e c t s of o b t a i n i n g , using,, and

d i s p o s i n g of t h e s e materials would c e r t a i n l y have t o be e v a l u a t e d .

A major f e a t u r e of t h e DMSR i s t h e r e l a t i v e unavailabil2ty of special n u c l e a r material ( S M ) t h a t might be d i v e r t e d and converted into s t r a t e g i c s p e c i a l n u c l e a r material (S%NM) f o r use i n t h e p r o d u c t i o n of nuclear explosive devices.

Because a l l t h e fuel would be in a bonogeneous fluid,

there would never be any s u b u n i t s ( e 8 g o s f u e l elements) t h a t would be part i c u l a r l y e n r i c h e d i n a g i v e n " d e s i r a b l e " material o r d e p l e t e d w i t h respect t o s p e c i f i c contaminants.

IR a d d i t i o n , because t h e i n i t i a l f u e l

c h a r g e as w e l l a s a l l makeup f u e l would be denatured 235U and because "spentr8f u e l would n o t be

XXTIIO'T~~

.from the primary c o ~ ~ t a i m e n except t dur-

i n g decommissioning a t the end of r e a c t o r l i f e , t h e a c c e s s i b i l i t y sf even

the mixed fueh would be s e v e r e l y r e s t r i c t e d .

P o s t u l a t i n g ways of o b t a i n -

i n g S S M from any m i x t u r e c o n t a i n i n g fissile n u c l i d e s i s always p o s s i b l e ,

b u t , i n t h e case of t h e DMSR, t h e s e appear t o involve s p e c i a l d i f f i c u l t i e s

as w e l l

l o w prsauctivity,

3.7.1

P o t e n t i a l s o u r c e s of SSMM

M t e r t h e f i r s t f e w y e a r s of power o p e r a t i o n , " t h e p r i n c i p a l f i s s i l e n u c l i d e i n a DMSR would be 233U with

substantial amount of 235U.

HOW-

ever, b o t h n u c l i d e s would remain f u l l y d e n a t u r e d d u r i n g t h e e n t i r e operation.

Thus s af ter d i v e r s i o n alad Separation from o t h e r

CkaelD?Cd.

speC%@s

(many of which would be h i g h l y r a d i o a c t i v e ) , the f i s s i l e uranium would

s t i l l have t o be s u b j e c t e d t o an i s o t o p e enrichment p r o c e s s t o produce

SSNM.

o t h e r i s o t o p i c contaminants i n the uranium, notably 2 3 2 ~ swould

tend t o make tklis a d i f f i c u l t approach.

Tne n e x t m ~ s tabundant f i s s i l e material i n DMSR f u e l s a l t would be p l u t o n i m , w$th a m a x i m u m t ~ t a l - p l a n t fnventory ( a t end of p l a n t l i f e ) of

334 kg of 239Pu -k 241Pu.

However, this material would also c o n t a i n 182 kg

of 242h and 1.39 kg of 2 4 0 ~ u , which w u i d tend to detract from its value

*I n i t i a l l y ,

t h e dominant f i s s i l e nuclei would be 23'~ d e n a t u r e d with 2 3 8 ~ ,b u t because this mixture presumably would be a n item of i n t e r n a t i o n a l comerce, the DMSR would not r e p r e s e n t a p a r t i c u l a r l y a t t r a c t i v e s o u r c e of supply.

95 ... .. ............. = .w .

a s SSm,

A m o r e a t t r a c t i v e i s o t o p i c m i x t u r e would e x i s t e a r l y in t h e

p l a n t l i f e t i m e (e.g., tory

be nueh smaller

WQUM

24(4Pu

a f t e r one year of o p e r a t i o n ) , b u t t h e t o t a l inven-

- only

86 kg o i 23%u

24bpu w i t h 13 kg of

2%u,

h o t h e r p o t e n t i a l source of SSNM i n a DMSR would be 2 3 3 ~ a . T ~ F Snuc l i d e would have its m a x i m u m i n v e n t o r y of -63 kg e a r l y i n t h e r e a c t o r l i f e and slowly decllbne to about 41 kg at the end of l i f e ,

In principle,

this n u c l i d e , if it c o u l d be c l e a g a ~ yand r a p i d l y s e p a r a t e d from t h e res%: of t h e f u e l salt, c o u l d p r o v i d e an e q u i v a l e n t amount of h i g h - p u r i t y 23% through s i m p l e r a d i o a c t i v e decay.

3.%,2

A c c e s s i b i l i t y sf SSMM

A major c o n s i d e r a t i o n r e g a r d i n g the a c c e s s i b i l i t y o f v a r i o u s foms of p o t e n t i a l SSm i n a DMSR i s t h a t a l l t h e materials a r e i n t i m a t e l y mixed w i t h -350 Mg of h i g h l y radioactive f u e l s a l t w i t h no known method for

simple p h y s i c a l s e p a r a t i o n .

n u s , d i v e r s i o n of only a modest amount ( a

f e w kilograms of SSNM without plans f o r i s o t o p i c e n r i e b e n t would r e q u i r e

the removal of a number of tons of f u e l salt from t h e reactor system. need f o r such l a r g e (and o t h e r w i s e u n j u s t i f i a b l e ) s a l t removals,

The

which

w i t h o u t rep~lacenentwould s h u t d ~ m the r e a c t o r , couple$ w i t h t h e need for

an e l a b o r a t e chemical t r e a t m e n t f a c i l i t y t o i s o l a t e the produet

appears

to make t h i s approach r e l a t i v e l y i m p r a c t i c a l . In p r i n c f p a l , pure 233U could be diverted v i a the 233Pa route by

modifying t h e i n - p l a n t h y d r o f l u o r i n a t o r t o permit i t a use as a f l u o r i -

This would requlre two f l u o r i n a t i o n s of e a c h batch of salt, with

nator.

one o c c u r r i n g immediately after removal of t h e s a l t from t h e reactor t o s t r i p o u t the denatured uranium and a second about two months l a t e r t o r e c o v e r t h e 233U produced by 233Pa decay.

However, i f the system were

o r i g i n a l l y d e s i g n e d t o handle batches si s a l t no larger t h a n -1 m3, t h e

*Presumably, ~

t h i s approach would be rased o n l y t o d i v e r t plutonium because uranium d i v e r s i o n w o u ~ dr e q u i r e i s o t o p i c e n r i c b e n t and 23 +a d i v e r s i o n would e n c o u n t e r s e r i o u s timing problems, as well a s r e q u i r i n g t h e h a n d l i n g of more salt.

233U p r o d u c t i o n c a p a b i l i t y would be l e s s t h a n 3 k g f y e a r , which seems i m p r a c t i c a l l y low. AI%hough t h e removal of f i s s i l e material from a DMSR may b e awkward,

i f i t could be ascomplfshed w i t h o u t removing l a r g e q u a n t i t i e s of s a l t , t h e n t h e removal could be e a s i l y ~ ~ n c e s l eby d a d d i t i o n s of d e n a t u r e d 235U to t h e f u e l s a l t .

The change i n t o t a l uranium c o n c e n t r a t i o n would not

become s i g n i f i c a n t u n t i l a f t e r t h e exchange of a few t e n s of kilograms of fissile fuel Although a much more d e t a i l e d , q u a n t i t a t i v e a n a l y s i s t h a t c o n s i d e r e d t h e r e l a t i v e v a l u e s of v a r i o u s f o m s of SSMM would be r e q u i r e d t o p e r m i t

a comprehensive

~ S S ~ S S T I E of R~

t h e p r o l i f e r a t i o n s e n s i t i v i t y of t h e once-

through DNSR, t h i s g e n e r a l t r e a t m e n t s u g g e s t s t h a t t h i s c o n c e p t may cornp a r e f a v o r a b l y w i t h o t h e r a l t e r n a t i v e s i n terns of r e s i s t a n c e t o prolife r a t i o n of n u c l e a r e x p l o s i v e s .

...

.... .3 .... v

ALTERNATIVE DMSR CONCEPTS

Of t h e s e v e r a l MSR c o n c e p t s t h a t have been c o n s i d e r e d , t h e DMSR des c r i b e d i n t h e preceding s e e t i o n w a s judged t o be t h e one most f i r m l y based on c u r r e n t l y a v a i l a b l e technology.

However, i t i s n o t t h e o n l y

p r o l i f e r a t i o n - r e s i s t a n t MSR concept t h a t could be c o n s i d e r e d .

However,

because a h i g h l e v e l of p r o l i f e r a t i o n r e s i s t a n c e i n a n MSR a p p a r e n t l y req u i r e s d e n a t u r e d f u e l , which imposes some d e s i g n r e s t r i c t i o n s , t h e major d i f f e r e n c e s among t h e a l t e r n a t e c o n c e p t s i n v o l v e t h e f u e l c y c l e .

4.1

F u e l Cycle Choices

P o s s i b l y t h e most f a v o r a b l e f u e l c y c l e f o r any DMSR, a t l e a s t from t h e p o i n t of r e s o u r c e u t i l i z a t i o n , would be one w i t h break-even b r e e d i n g performance.

C a l s u l a t i o n s f o r a DMSR c o r e without n e u t r o n â&#x201A;Źlux f l a t t e n -

i n g t o extend t h e life expectancy of t h e g r a p h i t e moderator showed9 t h a t break-even breeding w a s m a r g i n a l l y p o s s i b l e w i t h f u l l - s c a l e f i s s i o n p r o d u c t t r e a t m e n t of t h e f u e l u s i n g a redustive-extraction/mgtaB-tPansfer process'"

s i m i l a r t o t h a t proposed f o r t h e MSBR.

Even i f break-even

performance were n o t a t t a i n e d , t h e i n i t i a l f u e l change could be "used" f o r s e v e r a l r e a c t o r p l a n t l i f e t i m e s by f e e d i n g moderate amounts of f i s -

sile f u e l . Tne n e x t s t e p downward i n performance might be a concept i n v o l v i n g t r e a t m e n t of t h e f u e l f o r p a r t i a l f i s s i o n - p r o d u c t removal by chemical o p e r a t i o n s s i g n i f i c a n t l y d i f f e r e n t from t h e r e f e r e n c e process.

This ap-

proach probably would l e a d t o s t i l l lower c o n v e r s i o n r a t i o s , b u t i t might permit i n t e r n a l r e c y c l e of t h e f u e l through a few g e n e r a t i o n s of r e a c t o r s and, t h e r e f o r e , o f f e r b e t t e r r e s o u r c e u t i l i z a t i o n t h a n t h e once-through f u e l cycle. Some improvement i n f u e l u t i l i z a t i o n over c u r r e n t - t e c h n o l o g y LWRs could be achieved even w i t h o u t o n - s i t e chemical t r e a t m e n t f o r f i s s i o n product removal.

P e r i o d i c replacement of t h e f u e l c a r r i e r s a l t ( a f t e r re-

covery and r e t u r n of only t h e uranium) w i t h material t h a t i s f r e e of %is-

sion products and higher actinides WQUMimprove t h e utilization of fiss i l e f u e l , though i t would i n c r e a s e t h e consumption of o t h e r f u e l - s a l t constituents e ..... ..:.w ....,

A l l t h e MSR o p t i o n s from t h e b r e e d e r through t h e s i m p l e s t c ~ r t v e r t e f

WQUMt a k e advantage of t h e f a c t t h a t t h e noble-gas

f i s s i o n products

Q i n c l u d i n g t h e v e r y i m p o r t a n t w u c ~ e a rpoison b 4%e) are v e r y s p a r i n g l y s o l u b l e i n t h e molten f l u o r i d e f u e l .

I ~ ? ~ P I St, h e y would all use simple

s t r i p p i n g with gaseous helium t o remove krypton and xenon from the p r i I n a d d i t i o n , t h e y would a l l t a k e advantage of the f a c t t h a t

mary system.

t h e noble-metal

and seminoble-metal f i s s i o n p r o d u c t s do not f o m s t a b l e

f l u o r i d e s i n t h e f u e l and would p r e c i p i t a t e as e l e m e n t a l species, p r i -

m a r i l y on metal s u r f a c e s o u t s i d e t h e r e a c t o r c o r e .

4 , E. 1 Breakeven b r e e d i n g The presence of 2 3 8 U i n a DMSR, combined w i t h t h e e f f e c t s of f l u x

f l a t t e n f n g , s u f f i c i e n t l y reduces t h e n u c l e a r p e ~ f o ~ ~ ~s on tch ae t a net b r e e d f n g r a t i o s u b s t a n t i a l l y greater t h a n 1.0 p r o b a b l y c o u l d not b e

achieved

even w i t h f u l l - s c a l e

fission-product

breeding g a i n p r e s ~ m a i b l ywould be u n d e s i r a b l e

processing ita a

(A positive

pPoliferation-resfstant

system because P t would r e q u i r e t h e p e r i o d i c 'qexport.n'of excess f i s s i l e

rnaterfal_.)

However, t h e s t u d i e s t h a t have been c a r r i e d o u t i n d i c a t e t h a t

bseak-even b r e e d i n g i s w i t h i n the u n c e r t a i n t y l i m i t s o f t h e n e u t r o n i c caP-

cralations f o r a f l u x - f l a t t e n e d DMSR core w i t h a 30-year moderator l i f e expectancy. tinized

Extended o p e r a t i o n a t break-even would r e q u i t e a c a r e f u l l y op-

COPE?

d e s i g n as well a s c o n t i n u o u s f u e l - s a l t processing on a rela-

t i v e l y s h o r t t i m e c y c l e (-20

d ) t o remove f i s s i o n p r o d u c t s and r e t a i n ( ~ r

r e t u r d a l l f i s s f l e anel h i g h e r - a c t i n i d e n ~ c l i d e s . The r e f e r e n c e f u e l p r o c e s s i n g concept proposed f o r t h e MSBR c o u l d not b e d i r e c t l y a p p l i e d t o a DMSR f o r s e v e r a l reasonse

I s o l a t i o n of 23%

would lead to 2,

W Q ~ I .n ~ot

be acceptable i n a DMSR because i t s decay

s u p p l y o f diversiom-sensitive, h i g h - p u r i t y

233ue

I s o l a t i o n of p r o t a c t i n i u m would be accompanied b y removal and loss of p l u t o n i m f ~ o mthe a p e r a t i n g s y s t e m .

This would not o n l y d e g r a d e sys-

tern performance brit a l s o provide a source of p l u t o n i u m t h a t would have t o be safeguarded a n d / o r d i s p o s e d o f .

99 ... ..."W ..........

The r e f e r e n c e system w i t h o u t p r o t a c t i n i u m i s o l a t i o n would have no means f o r removing f i s s i o n - p r o d u c t

zirconium, which would then reach

u n d e s i r a b l y high chemical c o n c e n t r a t i o n s . However, t h e r e f e r e n c e p r o c e s s could b e modified t o meet t h e requirements

A modified p r o c e s s ( d e s c r i b e d i n Ref. 9 1 , i n addi-

of t h e BMSR concept.

t i o n t o p r o v i d i n g the r e q u i r e d f i s s i o n - p r o d u c t removal, t ~ o u l do f f e r other advantages

The t o t a l pButoniew i n v e n t o r y would b e limited because t h e plutonium

would eventually b e c o n s u e d at its p r o d u c t i o n r a t e in t h e reactor.

The r e a c t o r would s e r v e as its o m " i n c i n e r a t o r " f o r t r a n s p l u t o n 8 m a c t i n i d e s , which would be c o ~ ~ t i n u o u ~ recycled ly i n the fuel.

N e i t h e r t h e p r o t a c t i n i u m nor t h e plutonium would e v e r be i s o l a t e d f r m

a l l o t h e r hfghly r a d i o a c t i v e species. T h i s modified p r o c e s s i n g concept would use all t h e b a s i c u n i t o p e r a t i o n s proposed f o r the MSBR system lira e s s e n t i a l l y t h e same sequence.

However,

a d d 8 t i o n a l , though s i m i l a r , p r o c e s s s t e p s would be r e q u i r e d t o remove zirconium on a r e a s o n a b l e t i m e s c a l e , and t h e s e are included i n t h e conc e p t u a l flow s h e e t .

Some removal sf neptunium a l s o might be d e s i r a b l e t o

avoid t h e long-term poisoning e f f e c t s of a37Np and 238Pu; t h i s probably could be included without adding s i g n i f i c a n t l y t o t h e complexity of the processing f a c i l i t y . With f u l l - s c a l e f i s s i o n - p r o d u c t

removal and break-even b r e e d i n g

f u e l in a DMSR c o u l d be used i n d e f i n i t e l y .

the

That i s , at t h e e n d - s f - l i f e

one r e a c t o r p l a n t , t h e fuel s a l t could be t r a n s f e r r e d t o a new p l a n t and used without any s i g n i f i c a n t i n t e r m e d i a t e t r e a t m e n t .

During t h e l i f e of any g i v e n p l a n t , adding thorium as t h e p r i n c i p a l f e r t i l e material and 2 H U

t o m a i n t a i n compliance w i t h d e n a t u r i n g requirements would be n e c e s s a r y , b u t no f i s s i l e a d d i t i o n s would be r e q u i r e d . fuel-carrier salt (Lip

+ BeF2

Other r o u t i n e removals s f

4- ThFb) and a d d i t i o n s o f ReFz and ThFt+"

would be r e q u i r e d t o maintain t h e d e s i r e d chemical composition o f the The removed c a r r i e r s a l t c o u l d be disposed of ( a f t e r c o n v e r s i o n

salt.

t o a s u i t a b l e form) o r c h e m i c a l l y processed f o r r e c y c l e i n t o o t h e r

MSRs.

Lithium f l u o r i d e would be formed c o n t i n u o u s l y in t h e s a l t from t h e lithium used i n t h e redustivg-extPaction/meta%-tpansfep s t e p s .

100

Converter o p e r a t i o n w i t h f u e l p r o c e s s i n g

4,1.2

Because t h e r e s u l t s of t h e c u r r e n t l y completed n e u t r o n i c calcuba-

t i o n s will n o t s u p p o r t any f i n a l c o n c l u s i o n s about the b r e e d i n g p o t e n t i a l of f u l l y optimized DNSR c o r e s , c o n s i d e r a t i o n must be g i v e n to the consequences o f c o ~ ~ e r s i or an t i o s lower t h a n 1,OO. formed f o r t h e two-zone

The e v a l u a t i o n s were pes-

f l u x - f l a t t e n e d c o r e d e s c r i b e d f o r the 30-year f u e l

c y c l e w i t h t h e f u e l p r o c e s s i n g concept f o r t h e break-even b r e e d e r added. I f t h i s system were o p e r a t e d w i t h no c o n s t r a i n t on t h e enrichment sf t h e

uranium i n t h e r e a c t o r and no 2 3 8 ~addition, i t would g r a d u a l l y d e v e l o p

into a n MSBR as t h e 23*U was consmed.

The system would t h e n be f u l l y

s e l f - s u s t a i n i n g on thorium w i t h a b r e e d i n g r a t i o of about 1.03 b u t w i t h a v e r y h i g h enrichment of fissile uranium.

Breeding r a t f o s as h i g h as 1.11

could be a t t a i n e d by changing t h e thorium c o n c e n t r a t i o n a n d / o r t h e size of t h e i n n e r core zone.

w i t h t h e a d d i t i o n of enough 2 " ~ t o keep the in-

p l a n t uranium d e n a t u r e d a t a l l times, t h i s p a r t i c u l a r r e a c t o r system would u l t i m a t e l y r e q u i r e am a d d i t i o n a l 2 % i n n u c l e a r r e a c t i v i t y to be i n d e f i n i t e l y operable.

This r e a c t i v i t y d e f i c i t , i f r e a l , c o u l d be s u p p l i e d i n

a number of ways. A m o d e r a t e f e e d of 2 3 % a t 20% enrichment would extend t h e f u e l s y c ~ e

t o about 308 y e a r s .

A t t h a t t i m e , t h e 238U l o a d i n g would become exces-

s i v e , and the r e a c t o r could no l o n g e r be wade c r i t i c a l .

While even 300

y e a r s may b e much l o n g e r t h a n any r e a s o n a b l e planning h o r i z o n , t h i s res u l t i n d i c a t e s t h a t a f u l l y d e n a t u r e d MSR could have a very long, i f n o t unlimited, f u e l lifetime.

Iâ&#x201A;Ź t h e enrichment of t h e f e e d material were

allowed t o r i s e t o 33% 2 3 5 U 9 r e a c t o r o p e r a t i o n could be s u s t a i n e d i n d e f i n i t e l y WfthOblt fuel d i s c a r d . Because t h e b u i l d u p of 2 3 % i s t h e l i m i t i n g phenomenon i n t h e f u e l c y c l e of any nonbreeding DMSW, any p r o c e s s t h a t would have t h e e f f e c t of removing 2 3 8 1 ~would improve t h e c h a r a c t e r i s t i c s sf t h e cycle. fate% feed

With t h e

enrichment s e t a t 20% 2 3 5 U 9 t h e b u f l d u p of 238U could be l i m i t e d

I n d e f i n i t e l y o p e r a b l e " i s a r b i t r a r i l y d e f i n e d h e r e as m a i n t a i n i n g kefg 1. 1.0 f o r 600 y e a r s o r bong el^. I n a l l extended f u e l c y c l e s , t h e f u e l i s presumed t o be t r a n s f e r r e d w i t h o u t l o s s f r m one r e a c t o r p l a n t t o a n o t h e r as r e q u i r e d by hardware l i f e t i m e c o n s i d e r a t i o n s ,

.,.. w .."

by removing some uranium from t h e f u e l s a l t and r e p l a c i n g i t wieh f r e s h I f onlgr 1% of t h e uranium i n v e n t o r y were removed each year and con-

feed.

s i g n e d t o waste o r t o o f f - s i t e r e c o v e r y , t h e i n - p l a n t i s o t o p i c composition would r e a c h e q u i l i b r i u m w i t h i n 300 y e a r s s and t h e f u e l c y c l e c o u l d b e continued indefinitely.

faTa even more a t t r a c t i v e c h o i c e would be t o remove

some of t h e uranium, s t r i p o u t p a r t of t h e 238U, and r e t u r n t h e remainder t o the reactor plantr

To examine t h i s case, w e assumed t h a t 2% of t h e re-

a c t o r i n v e n t o r y would be t r e a t e d each year and t h a t the r e t u r n i n g uranium would c o n t a i n one-half e v e r was g r e a t e r . tion.)

t h e o r i g i n a l 23*U o r enough f o r d e n a t u r i n g

which-

(Only 238U was e x t r a c t e d i n t h i s p r e l i m i n a r y c a l c u l a -

The c a l c u l a t i o n showed t h a t t h i s approach a l s o would a l l o w i n d e f i -

n i t e o p e r a t i o n and would r e q u i r e l e s s f e e d material ( s e e t h e following d i s c u s s i o n ) than t h e o t h e r o p t i o n s . 4.1.3

P a r t i a l f i s s i o n - p r o d u c t removal

Although t h e r e f e r e n c e f i s s i o n - p r o d u c t

could p r ~ c e s s f ~concept g

s t r o n g l y a f f e c t t h e v e r y long-term v i a b i l i t y of DMSRs, t h e f i s s i o n - p r o d u c t p r o c e s s would r e q u i r e s u b s t a n t i a l t i m e and e f f o r t f o r commercial development, and, even t h e n , i t might n o t be a market s u c c e s s .

Consequently,

c o n s i d e r i n g a l t e r n a t i v e p r o c e s s e s might b e u s e f u l A v a r i e t y o f a l t e r n a t i v e s e p a r a t i o n s procedures have been examined

o v e r t h e y e a r s i n t h e ORNL MSR program f o r p o s s i b l e a p p l i c a t i o n i n f u e l reprocessing operations.

P o s s i b l e r e c o v e r y of p r o t a e t i n t m , uranium, and

o t h e r a c t i n i d e s by s e l e c t i v e p r e c i p i t a t i o n of o x i d e s has been examined, though most methods have p r e f e r r e d removal of uranium i s o t o p e s by f l u o r i n a t i o n t o v o l a t i l e UF6.

Attempts t o remove t h e l a n t h a n i d e s ( t h e most

important p a r a s i t i c a b s o r b e r s of n e u t r o n s ) have included p r o c e s s e s based o n i o n exchange, p r e c i p i t a t i o n of i n t e r m e t a l l i s compounds, and even v o l a t i l i z a t i o n a t low p r e s s u r e of t h e o t h e r melt c o n s t i t u e n t s " t o l e a v e tihe v e r y n o n v o l a t i l e l a n t h a n i d e t r i f l u o r i d e s behind.

A l l such p r o c e s s e s re-

q u i r e s o l i d s h a n d l i n g , and many a l s o have o t h e r d i s a d v a n t a g e s .

*Such

None w a s

a s e p a r a t i o n w i g h t be f e a s i b l e , a f t e r f l u o r i n a t i o n of t h e uranium, f o r a. f u e l c o n s i s t i n g o n l y of LiF, BeF2, and W 4 , b u t i n c l u s i o n o f c o n s i d e r a b l e TkF4 Cas i n a DMSR f u e l ) d e f e a t s s u c h a p r o c e s s .

102 developed f a r enough t o l e a d t o a n i n t e g r a t e d process.

Further, a f t e r

process, Which, dPSCQVery O f t h e s@ductiVe-eXtractisn/metal-~ra~S~er

though cmplbex, involved h a n d l i n g o n l y l i q u i d s and gases, s t u d i e s of a l l o t h e r s e p a r a t i o n s were l a r g e l y abandoned.

An ion-exchange process â&#x201A;Ź o r s e l e c t i v e removal of l a n t h a n i d e ions f r ~ m the molten f u e l has long seemed a t t r a c t i v e in principle,B9 b u t no attrac-

tive i o n exchanger â&#x201A;Ź o r t h e s e materials h a s been demonstrated.

An obvious

d f f f i ~ u l t yi s posed by t h e a g g r e s s i v e tendency of t h e molten LiF-BeF2ThF4-WF4 system t o react w d t h most materials t h a t a r e l i k e l y t o b e u s e f u l ,

C e r t a i n r e f r a c t o r y l a n t h a n i d e compounds ( s u c h a s c a r b i d e s , n i t r i d e s , or s u l f i d e s ) could c o n c e i v a b l y b e u s e f u l and s u f f i c i e n t l y s t a b l e .

The o n l y

c a n d i d a t e materials t o d a t e have been materials such a s CeF3 and LaF3.

By v i r t u e of t h e f o m a t i ~ s f~ n~e a r l y i d e a l s o l i d soluti~nsamong t h e rare e a r t h t r i f E u o r i d e s , these compounds are c a p a b l e of removing o t h e r ( h i g h e r cross-section)

l a n t h a n i d e s from t h e m o l t e n f l u o r i d e s .

The ~ l e u t r ocross ~~

sections of cerium and Pantharem are n o t n e g l i g i b l e ; because such an exchange p r o c e s s s a t u r a t e s t h e t r e a t e d f u e l with CeF3 o r LaF3, t h e r e s u l t i n g f u e l s o l u t i o n still has s u b s t a n t i a l p a r a s i t i c n e u t r o n a b s o r b e r s .19

CeF3 ( o r LaF3) exchanger a l s o would presumably remove trivalent a c t f n i d e s ( i n c l u d i n g plutonium) from t h e molten f u e l , fQP

ThPs would be u n a c c e p t a b l e

EbMSR.

No o v e r a l l chemical p r o c e s s based on such s e p a ~ ~ t f has o ~ sbeen de-

scribed e

Obviously, much development would be n e c e s s a r y b e f o r e such a

p r o c e s s could b e demonstrated,

hkaso, s e v e r a l s o l i d s - h a n d l i n g o p e r a t i o n s

a p p a r e n t l y would be r e q u i r e d and no p r o c e s s based

QII

these operations

could b e c a p a b l e of p r o c e s s i n g a DMSR on a s h o r t t i m e cycle.

However,

g i v e n the p r e s e n t s t a t e of knowledge, t h e f o l l o w i n g p r o c e s s can b e visua l i z e d t o o p e r a t e m r e l a t i v e l y l a r g e (1- t o 2-hn3) b a t c h e s of DMSR f u e l , possibly a f t e r c o d i n g f o r 5 o r 6 d e

The f o l l o w i n g s t e p s would be neces-

sary *

.w .... .9 ...

probably s t i l l be m o s t l y t r i â&#x201A;Ź % u ~ r i d e ~This , o x i d i z e d system w i l l b e corrosive, b u t i t should b e manageable i n equipment of n i c k e l

o r nickel-elad Hastelloy. S t e p 2.

P r e c i p i t a t e t h e i n s o l u b l e o x i d e s u s h g water vapor d i l u t e d i n helium.

'RE o x i d e s UO2% Pa205, Pu02, Ce02, probably Np02, and

p o s s i b l y Am02 and @m02 should be obtairped,

With the e x c e p t i o n

of %r02and Pa205, t h e s e w i l l be l a r g e l y i n s o l i d S U P U ~ F Q ~ The .

oxide s o l i d s o l u t i o n i s l i k e l y t o c o n t a i n 15 t o 20% of Tho2; this

W Q U P ~correspond t o a few (less t h a n 5 ) p e r c e n t of t h e T%F4 p r e s ent i n the fluoride.

Recover t h e o x i d e s by d e c a n t a t i o n and f i l -

tration. S t e p 3.

H y d r o f l u o r i n a t e t h e o x i d e s from s t e p 2 i n t o t h e p u r i f i e d LIPBeP2-ThF4 m e l t from s t e p 7 and reduce t h e m e l t w i t h H2 and t h e n

with l i t h i u m , t h o r i m , o r beryllium t o r e c o n s t i t u t e f u e l with t h e d e s i r e d U F ~ / U Fr a~t i o . Step 4.

H y d r ~ f l u o ~ i n a tteh e l i q u i d from s t e p 2 t o remove e x c e s s o x i d e ion.

k i d i z e t o g e t samarium and ( i f p o s s i b l e ) europium t o SmF3

and EuF3. S t e p 5.

Treat t h e m e l t from s t e p 4 w i t h a n excess of CeF3.

This might

be done i n a column o r i n a two- o r three-batch c o u n t e r c u r r e n t

operation.

This

T ~ ~ Q V a ~major S

f r a c t i o n of t h e rare e a r t h s b u t

does e s s e n t i a l l y nothing f o r cesium, rubidium, s t r o n t i u m , and barium.

( I f neptunium, americium, and curium a r e a p p r e c i a b l y

h a r d e r t o o x i d i z e than plutonium, t h e y s h o u l d remain i n t h e s a l t i n s t e p 2 and should be removed on t h e CeF3 i n s t e p 5.) Step 6,

The LfF-BeF2-ThF4 m e l t from s t e p 5 c o n t a i n s s d y a f r a c t i o n of

t h e r a r e e a r t h p o i s o n s b u t , of c o u r s e , i s s a t u r a t e d w i t h CeF3. oxidize the Step 7.

ee*

to ~ e 4 + .

~ r e c i p i t a t et h e ce4+ a s

c ~ o ~ some .

Ce02, b u t t h e q u a n t i t y should be s m a l l .

by d e c a n t a t i o n and f i l t r a t i o n .

w i l l accompany t h e Separate the p r e c i p i t a t e

Feed t h e molten LiP-BeP2-ThF4

the f i s s i l e material r e c o v e r y o p e r a t i o n in s t e p 3. S t e p 8.

D i s s o l v e t h e s o l i d &F3

(contaminated w i t h rare e a r t h s ) from s t e p

5 in some s u i t a b l e s a l t ( p r e f e r a b l y n o t SLiF-BeP2) and o x i d i z e t h e Ce3+ t o @e4'@

, ....., -c%w

104

Step 9.

P r e c i p i t a t e t h e cerium a s Ce02 and recover t h e p r e c i p i t a t e by d e c a n t a t i o n and f i l t r a t i o n .

Discard a p o r t i o n of t h e molten

s a l t , which c o n t a i n s rare e a r t h f i s s i o n p r o d u c t s , t o waste s t o r age. Return t h e remainder with t h e n e c e s s a r y makeup t o s t e p 8. S t e p 10.

Combine t h e Ce82 from s t e p 9 w i t h t h a t from s t e p 7 and treat t h e s e sollids w i t h %%F and H 2 t o o b t a i n CeF3 ( p l u s some ThF4). U s e t h i s as t h e major p a r t of t h e r e a g e n t f o r s t e p 5.

This process would have a number of d i s a d v a n t a g e s when compared w i t h t h e reductive-extraction/metal-transfer

process.

Zirconium, cesium, ru-

bidium, s t r o n t i u m , and barium would not be removed, though none of t h e s e

i s a major problem.

Neptunium p r o b a b l y would n o t be removed, though am-

e r i c i u m and curium may be.

Iodine

W Q U ~be~

QXidatiOn Or SUbSe3qUent hydrOflUOrfnaeiQnS.

removed e i t h e r during. t h e f u e l SeleIliWll and

s w i n g t h a t they a r r i v e a t t h e p r o c e s s i n g p l a n t -rnsnight

tellUriUIT3

- a%-

be v o l a t i l i z e d as

elements o r as f l u o r i d e s d u r i n g t h e f u e l o x i d a t i o n s t e p (and t h e y might cause a

C O ~ ~ O Sproblem ~ Q ~

f o r t h e process).

Heat g e n e r a t i o n by t h e f u e l ,

even a f t e r a few d a y s c o o l i n g t i m e , would p r e s e n t problems, and t h e conp l e x process would be d i f f i c u l t ( p o s s i b l y i m p o s s i b l e ) t o e n g i n e e r .

b e s t , s e v e r a l days would be r e q u i r e d t o g e t a b a t c h of DMSR f u e l s o l v e n t

through t h e p r o c e s s , though t h e f i s s i l e materials might be r e t u r n e d t o t h e r e a c t o r w i t h a 2-el holdup.

An a p p r e c i a b l e i n v e n t o r y of fuel material ( b u t

perhaps not more t h a n 5% of r e a c t o r i n v e n t o r y ) would be c o o l i n g and i n t h e p r o c e s s i n g area.

4 * I e 4 S a l t replacement Even w i t h no chemical removal of f i s s i o n p r o d u c t s , t h e n e u t r o n poisoning e f f e c t Fn a BMSR does n o t begin t o approach s a t u r a t i o n u n t i l a f t e r

about 1 5 y e a r s of power o p e r a t i o n a t a 75% c a p a c i t y f a c t o r . fission-product

Thus, i f t h e

i n v e n t o r y could be h e l d a t o r below that corresponding t o

a 15-year l e v e l , a s i g n i f i c a n t r e d u c t i o n i n f u e l i n g requirements could be realized.

The simplest way ts l i m i t the f i s s i o n - p r o d u c t c o n c e n t r a t i o n i n

t h e s a l t i s t~ d i s c a r d a p o r t i o n of t h e s a l t on a r o u t i n e s c h e d u l e and r e p l a c e i t wfth c l e a n s a l t .

with no r e f i n e m e n t , s a l t d i s c a r d would re-

q u i r e replacewent sf t h e f i s s i l e material as well a s t h e fertile cowponerat

and the ~ d v e n t( o r c a r r i e r ) s a l t and, t h e r e f o ~ e ,would a c t u a l l y r e q u i r e

a l a r g e r uranium s u p p l y t h a n t h e 30-year once-through f u e l cycle proposed f o r t h e r e f e r e n c e IBmR concept.

However, uranium i s e a s i l y and e f f e c t i v e l y

s e p a r a t e d f r s m t h e rest o f t h e f u e l m i x t u r e , so t h e d e n a t u r e d uranium c o u l d b e removed and r e c y c l e d

%e: t h e r e a c t o r site w i t h a minimum of ef-

Depending on t h e rate 06 s a l t replacement, t h i s approach would

fort.

s i g n i f i c a n t l y reduce t h e requirement f o r f i s s i l e uranium below t h a t fsr t h e simple once-through c y c l e 4.2

F u e l C y d e Performance

O f t h e a l t e r n a t e f u e l cycles c o n s i d e r e d i n t h i s s e c t i ~ n ,t h e break-

even b r e e d e r , if i t were s u c c e s s f d , would p r o v i d e f o r t h e b e s t utilizat i o n of f i s s i l e f u e l resources ~ ~ 3 % ) . ~f t h a t system were s t a r t e d up

o n 20% e ~ r i ~ 2h 3e %~ ~~ i~ t would probably r e q u i r e 700 to 1000 ~g of natu-

ral U308 t o p r o v i d e t h e i n i t i a l f u e l l o a d i n g f o r each 1 CX of e l e c t r i c generating capability.

[The s e p a r a t i v e work t o e n r i c h t h i s f u e l t o 20%

235U would be less t h a n 1 m i l l i o n s e p a r a t i v e work u n i t s (SWU),]

However,

once p r o v i d e d , t h i s f u e l would c o n t i n u e t o p ~ ~ d l u ceel e c t r i c i t y i n an a r b i t r a r i l y long s u c c e s s i o n of power stations ( o r as long a s f e r t i l e material

was a v a i l a b l e ) .

Thus, t h e e f f e c t i v e resource requirement could be made

a r b i t r a r i l y small by averaging I t Over

large number

plants.

Even i f

t h e i n i t i a l f u e l charge were used in o n l y one pliant, t h e r e s o u r c e requirement would be o n l y 110 t o 20% of t h a t f o r an LWR w i t h similar e l e c t r i c gen-

erating c a p a b i l i t y . The c o n v e r t e r o p t i o n s w i t h f u e l processing p r o v i d e o t h e r estimates of t h e potential performance of DMSRs with f i s s i ~ n - p r o d u c t c l e a n u p (Table 2 9 ) . The o p t i o n s , which w e r e d e s c r i b e d e a r l i e r , m y b e s u m a r i z e d as f~klows: Option

Fuel cycle

I n i t i a l l o a d i s 20% 235U; makeup f u e l i s 20% 2 3 5 U

I n i t i a l load i s 20% 2 3 5 U ;

makeup f u e l i s 33% 7 3 5 U

I n i t i a l l o a d i s 20% 2 3 s U ; makeup f u e l i s 20% 2 3 5 U

annual d i s c a r d o f 1%of uranium i n v e n t o r y ;

I n i t i a l l o a d i s 20% 2 3 5 U ; a n n u a l r e e n r i c h m e n t of 2% of uranium i n v e n t o r y t o d e n a t u r i n l i m i t o r t o one-lialf o f p r i o r 7 3 8 ~c o n t e n t ; makeup f u e l i s 20% 2 5 5 C

106 Table 29, Performance d a t a f o r long-term f u e l c y c l e options f o r DMSRS" with f u l l - s c a l e f i s s i o n - p r o d u c t removal Optionb

0.90 0.74

0.94

0.94 0.93

0.94 0.89 0.89

Conversion ratio a f t e r

20 y e a r s 300 years 608 y e a r s

0.92 0.92

860 860

860 890

860

1000

428 460

5 80

500

600

1000 e

440 478

580 60 0

50 8 60 0

Uranium reenrichment, m / y e a r

0.60

Uranium d i s c a r a , ~ g i y e a r

0.24

2.9 3.1

2.7 3.0

2e8

0*93

Requirement f o r i n i t i a l core asaaing U398S

S e p a r a t i v e work, Mg SWU

860

988 980

Average requirement %or f u e l makeup p e r 30-year c y c l e

k$$,@3

During y e a r s 0--300 During y e a r s 301-600 S e p a r a t i v e work, Mg SW During years 0-380 During years 301-600

Fissile i ~ ~ e n t o ar yt equiPlbrium, Ng

ur a R ium A l l fissile nuclides

1.2 3.2

3.1

% o r 1 G W ~a t 75% c a p a c i t y f a c t o r .

b ~ e et e x t f o r c h a r a c t e r i z a t i o n of o p t i o n s . Mot o p e r a b l e beyond 300 y e a r s . d A t 300 y e a r s .

The t a b u l a t e d r e s u l t s show that a l l f o u r of t h e s e o p t i o n s would m a i n t a i n r e l a t i v e l y h i g h c o n v e r s i o n r a t i o s f o r v e r y long times.

The U308 r e s o u r c e

r e q u i r e m e n t s f o r t h e i n i t i a l c o r e l o a d i n g s a r e a l l s i m i l a r , and all a r e s l i g h t l y higher than t h a t f o r the once-through f u e l cycle (because sf t h e

volume of f u e l i n t h e p r o c e s s i n g system).

BO 7 TPle f u e l makeup r e q u i r e m e n t s a r e expressed i n m e t r i c t o n s of UaOg

for 30 y e a r s of o p e r a t i o n i n a 1-GWe p l a n t a t 75% c a p a c i t y f a c t o r and are a v e r a g e s f o r t e n 30-year c y c l e s .

The e f f e c t of t h i s a v e r a g h g i s most

pronounced f o r o p t i o n A; t h e f u e l makeup requirement i s o n l y a f r a c t i o n

of t h e a v e r a g e f o r t h e f i r s t one o r two r e a c t o r l i f e t i m e s and i s somewhat

g r e a t e r than t h e a v e r a g e f o r t h e l a s t c y c l e s .

Thus, while t h i s o p t i o n

would r e q u i r e more u r a m i m t h a n t h e o t h e r s i n t h e v e r y long tern, i t s per-

f o m a n c e f o r t h e f i r s t few r e a c t o r l i f e t i m e s would be q u i t e a t t r a c t i v e . Even f o r t h e long-term,

this r e s o u r c e requirement would be w e l l below

t h a t of c u r r e n t - g e n e r a t i o n LWRs.

Option B i l l u s t r a t e s t h e long-term sav-

ing i n uranium r e s o u r c e s t h a t could be achieved i f h i g h e r enrichments could be t o l e r a t e d f o r t h e r e l a t i v e l y s m a l l amsunts of makeup f u e l ,

Be-

cause the r e s o u r c e s a v i n g s are p r i n c i p a l l y long tlem and t h e r e q u i r e d uranium enrichment exceeds c u r r e n t l y p e r c e i v e d d e n a t u r i n g l i m i t s a p p e a r s t o be

OR^

of t h e less g r o m h s i ~ ~ogp t i o n s ,

this

The two remaining 0p-

t i o n s , C and D, b o t h show f a v o r a b l e r e s o u r c e u t i l i z a t i o n p r o p e r t i e s f o r long t i m e s with o n l y minor penalties f o r d i s c a r d e d uranium ( o p t i o n C > o r uranium s u b j e c t e d t o reenrichment ( o p t i s m B ) ,

O f t h e s e , o p t i o n P) c l e a r l y

would be p r e f e r a b l e if r e e n r i c h e n t were an a c c e p t a b l e procedure

Tke preceding f o u r c o n v e r t e r o p t i o n s a n d / o r t h e break-even breeder would r e q u i r e the a v a i l a b i l i t y of a cornpPex and expensive f u e l c l e a n u p f a c i l i t y w i t h h the primary containment of each r e a c t o r i n s t a l l a t i o n .

The technology f o r an i n t e g r a t e d proeessfng f a c i l i t y has n o t been f u l l y developed, and p a s t work c l e a r l y i n d i c a t e s t h a t a s u b s t a n t i a l development e f f o r t would be r e q u i r e d t o produce a commercially f u n c t i o n a l system. Even t h e n , t h e c a p i t a l , o p e r a t i n g , and maintenance c o s t s of such a system p o s s i b l y would have a s i g n i f f c a n t a d v e r s e impact on t h e overall economic

p e r f o r n a m e of the a s s o c i a t e d DMSR.

Other f a c t o r s t o be c o n s i d e r e d f o r

t h e s e o p t i o n s i n c l u d e d t h e w i l l i n g n e s s of t h e r e a c t o r o p e r a t o r t o assme

*Conceivably,

a s i n g l e cleanup f a c i l i t y could serve several r e a c t o r s a t a cornon s i t e , but such an arrangement would c o m p l i e a t e t h e 0 p e r a t i Q n and would add p r o b l e m ~of i n v e n t o r y a c c o u n t a b i l i t y among t h e v a r i o u s units.

r e s p o n s i b i l i t y f o r a chemical p r o c e s s i n g f a c i l i t y , t h e s o c i o p o l i t i c a l acc e p t a b i l i t y of c o l o c a t i n g such a f a c i l i t y w i t h e a c h DMSR, and t h e Picensi n g q u e s t i o n s t h a t way arise from s u c h an arrangement.

The ~ t h e rend of t h e rartge of p o s s i b l e f u e l cycle performances f o r DMSRs i s r e p r e s e n t e d by t h e 38-year cycle d e s c r f b e d e a r l i e r i n t h i s r e p o r t Although t h i s system, w i t h a l i f e t i m e r e q u i r e -

a s t h e r e f e r e n c e concept.

ment of 1860 Mg (2000 s h o r t t o n s ) of U308, would be t h e l a r g e s t consumer of natural

U K ~ R ~ U I Tand I

s e p a r a t i v e work among t h e DMSR o p t i ~ n sc o n s i d e r e d ,

i t s t i l l wuuld r e q u i r e s u b s t a n t i a l l y less of t h e s e commodities than t h e

once-through f u e l c y c l e i n l i g h t - w a t e r reactors.

In t h e a b s e n c e of fa-

c i l i t i e s f o r r e c y c l i n g t h e won-SIW constituents of t h e f u e l s a l t , t h i s approach would u s e less of s u c h materials t h a n any of t h e o t h e r a l t e r n a -

tives.

However, d e s p i t e t h e 38-year f u e l c y c l e , t h i s concept would n o t

e l i m i n a t e all o n - s i t e chemical t r e a t m e n t u f t h e f u e l s a l t .

t o maintain the desired

u3+/u4+

Tlne a c t i v i t i e s

r a t i s i n t h e f u e l ana t h e t r e a t m e n t s t o

l i m i t t h e l e v e l of oxide c o n t a m i n a t i o n i n t h e s a l t would s t i l l be needed. Thus, even t h e "sFmpPest" DMSW would r e q u i r e

SQIW

equipment f o r and some

t e c h n i c a l competence i n chemical p r o c e s s i n g , even though n e i t h e r would directly

iRVQlVe

the

ill

t h e system.

The i n t e r m e d i a t e concepts t h a t make u s e of a s h o r t e r s a l t d i s c a r d

c y c l e merely s u b s t i t u t e consumption of o t h e r f l u o r i d e salts f o r p a s t o f t h e f i s s i l e uranium consumption i n t h e r e f e r e n c e 30-year

cycle,

Because

t h e s e o t h e r f l u o r i d e s ( e s p e c i a l b y 'Lip> may a l s o be r e l a t i v e l y e x p e n s i v e t h i s s u b s t i t u t i o n might not always b e c o s t e f f e c t i v e .

I n a d d i t i o n , any

system t h a t used s a l t d i s c a r d would have t o r e c o v e r uranium from t h e " w a ~ t e ' s~a l t t o p r e v e n t e x c e s s i v e uranium consumption.

This would add

y e t a n o t h e r chemical p r o c e s s i n g o p e r a t i o n t o t h e r e a c t o r p l a n t .

The a l t e r n a t i v e s t h a t r e l y on s p e c i a l t r e a t m e n t schemes to remove

flssion p r o d u c t s from t h e f u e l salt may have a t t r a c t i v e fuel. u t i l i z a t i o n c h a r a c t e r i s t i c s , b u t t h e y have n o t been analyzed i n s u f f i c i e n t d e t a i l eo

p e r n i t an accurate c h a r a c t e r i z a t i o n .

I n addition, c o n s i d e r a b l e r e s e a r c h

and development would be r e q u i r e d b e f o r e such p r o c e s s e s cou$d be shown t o be t e c h n i c a l l y f e a s i b l e .

Consequently, l i t t l e i n c e n t i v e is a p p a r e n t a t

t h i s t i m e t o propose new and d i f f e r e n t chemical p ~ ~ e s s cio n~c e~p g ts BMSRs

â&#x201A;Ź 0 ~

CQPIPIERCIWLPZATION CONSIDERATIONS

While t h e t e c h n o l o g i c a l f e a s i b i l i t y , t h e o v e r a l l t e c h n i c a l perfomance, and t h e p r o l i f e r a t i o n r e s i s t a n c e of t h e DMSR are important charac-

t e r i s t i c s t o be c o n s i d e r e d i n a s s e s s i n g I t s v a l u e as an a%ternat%ve nuc l e a r c o n c e p t , a n o v e r r i d i n g c o n ~ i d e ~ a t i oins l i k e l y t o be t h e c o m e r c i a l t z a t i o n p o t e n t i a l of t h e system.

This g e n e r a l a t t r i b u t e i n c l u d e s a

number of c o n s i d e r a t i o n s , such a s :

t h e probable total. c o s t of developing a commercia1Py r e a d y system;

t h e t i m e r e q u i r e d f o r such development, which s t r o n g l y a f f e c t s t h e impact a system can have on power weeds;

t h e probable n e t economic performance of commercial u n i t s , which det e r m i n e s the a t t r a c t i v e n e s s of t h e concept t o i t s p o t e n t i a l u s e r s , t h a t i s , t h e e l e c t r i c power u t i l i t i e s ;

t h e ease of licensability of t h e commercial p l a n t s , which i s a ref l e c t i o n of t h e ~ o n c e p t ~sso c i o p o l i t i c a l a t t r a c t i v e n e s s , a s well as i t s t e c h n i c a l performance.

Some r e l e v e n t i n f o r m a t i o n about t h e DMSR with r e s p e c t t o each of t h e s e p o i n t s is p r e s e n t e d i n t h e following d i s c u s s i o n .

5.1

Research and Development

S i n c e MSR r e s e a r c h and development has been under way f o r some 30

years, t h e b a s i c technology i s w e l l understood.

However, much of i t has

n o t been developed t o t h e s t a g e and scale t h a t would be r e q u i r e d f o r t h e c o n s t r u c t i o n of l a r g e r e a c t o r systems,

Thus, B s i g n i f i c a n t R&D e f f o r t

would be an important p a r t of any program t o commercialize MSRs,

I n ad-

d i t i o n , u n t f l r e c e n t l y , development w a s c o n c e n t r a t e d on r e a c t o r concepts w i t h a good breeding g a i n and a low f i s s i l e i n v e n t o r y

that the result-

i n g thermal b r e e d e r r e a c t o r system would have a r e a s o n a b l y s h o r t d ~ k ~ b l i ~ l g

t i m e and could be considered a v i a b l e a l t e r n a t i v e ( o r complement) t o f a s t b r e e d e r systems.

The technology needs of t h e modified r e a c t o r concept

t h a t h a s been developed i n response t o t h e r e c e n t emphasis on p r o l i f e r a -

t i o n r e s i s t a n c e d i f f e r from t h o s e of t h e nominal b r e e d e r concept.

110

5.1.1

current status

MSW development h a s been c a r r i e d through t h e d e s i g n and o p e r a t i o n t e s t r e a c t o r , t h e MSRE, which w a s an 8-Wt reac-

of a proof-of-principle

t o r that o p e r a t e d a t ORNL from 1965 t o 1969. t h e b a s i c r e l i a b i l i t y of a molten-salt

T h i s r e a c t o r demonstrated

system, s t a b i l i t y of t h e f u e l

salt, c o m p a t i b i l i t y s f f l u o r i d e s a l t s w i t h U s t e l l o y - N and graphite, rel i a b i l i t y of m o l t e n - s a l t pumps and h e a t exchangers, and maintenance of a r a d l o a c t i v e flufd-fueled

system by remote methods.

The reactor w a s c r i t -

ical over 17,000 h , c i r c u l a t e d f u e l s a l t f o r n e a r l y 22,000 h , and genera t e d over 100,000 plwh of thermal e n e ~ g y . The MSRE had achieved a l l t h e

objectives of t h e r e a c t o r t e s t program when i t w a s r e t i r e d i n 1969. After t h e s u c c e s s f u l o p e r a t f o n of t h e MSRE, t h e r e a c t o r concept appeared r e a d y f o r commercial development,

In p r e p a r a t i o n f o r f u r t h e r de-

ve%opment, t h r e e major reports w e r e prepared:

of an MSBR i n 1971 (Ref.

1972 (Ref.

loa),

$I9

a c o n c e p t u a l design study

a review of the s t a t u s o f development i n

and a program p l a n f o r development i n 1974 ( R e f .

21).

For r e a s o n s o t h e r t h a n t e c h n o l o g i c a l , t h e government d e c i d e d not t o fund f u r t h e r development of MSRs.

T%e program was c a n c e l l e d i n 1 9 7 3 , r e s t a r t e d

i n 1974, and finally terminated i n $976,

The development of a p r o l i % e r a t f s n - r e s i s t a n t DKSR would r e q u i r e basic a l l y t h e same t e c h n o l o g i c a l development program as was proposed f o r t h e

MSBR, b u t t h e emphasis would be on r e l i a b i l i t y , ease of c ~ m ~ ~ e ~ ~ i a l i ~ a t i ~ n , P i c e n s i n g , and p r o l i f e r a t h n r e s i s t a n c e r a t h e r t h a n on high breeding performance.

K i t h t h e s e o b j e c t i v e s in mind, t h e 1972 status-of-devebopment

r e p o r t h a s been updated, and t h e program p l a n for development h a s been

m o c ~ i l ~ i efco~r t h e DMSR."~

( m i l e t h e main o u t l i n e of BMSR development

requirements will be p r e s e n t e d i~ t h i s report, t h e reader I s r e f e r r e d t o

Ref.

102 f o r g r e a t e r d e t a i l . ) 5.1.2

Technology base f o r r e f e r e n c e DMSR

The base technology f o r MSWs i s w e l l e s t a b l i s h e d and h a s been l a r g e l y "'proven i n princâ&#x201A;Źp%e" by t h e o p e r a t i o n of t h e 24SWE.

While no major un-

~ e s o l v e dt e c h n i c a l i s s u e s e x i s t a t t h e p r e s e n t t i m e , a large R&B e f f o r t would be r e q u i r e d t o b r i n g nnol.ten-salt

technology t o commercialization.

11%

At t h e c l o s e of MSRE o p e r a t i o n , two major t e c h n i c a l issues appeared

unresolved.

The first was t h e c o n t r o l of tritium, which i s produced in

f a i r l y l a r g e q u a n t i t i e s i n a molten-salt f u s e through metal w a l l s .

system and which i s known t o d i f -

Subsequent e n g i n e e r i n g - s c a l e t e s t s have demon-

s t r a t e d t h a t tritiUll is o x i d i z e d in

S O d ~ W i lf l U O r o b O P a t e ,

t h e pPQpos@d S e C -

ondary s a l t f o r t h e DMSR, and a p p e a r s t o be handled r e a d i l y .

However, this

p r o c e s s i s n o t y e t w e l l understood, and t h e e f f e c t s of m a i n t a i n i n g an adeq u a t e c ~ n c e n t r a t i ~ofn t h e o x i d a n t on t h e long-term c o m p a t i b i l i t y of t h e

s a l t w i t h t h e s t r u c t u r a l a l l o y are unknown.

The second i s s u e involved t h e

c o m p a t i b i l i t y of Hastelloy-N w i t h f u e l s a l t .

Operation of t h e MSRE showed

t h a t t h e g e n e r a l c o r r o s i o n of Hastelloy-M and g r a p h i t e i n an o p e r a t i n g MSR

was n e a r z e r o , a s expected.

However, metal s u r f a c e s t h a t had been exposed

t o f u e l salt c o n t a i n i n g f i s s i o n products were unexpectedly found t o e x h i b i t grain-boundary

a t t a c k , which was s u b s e q u e n t l y shown t o be caused by reac-

t i o n w i t h t h e f i s s i o n product, t e l l u r i m .

F u r t h e r work has shorn t h a t t e l -

lurium a t t a c k c a n be c o n t r o l l e d by e i t h e r a m o d i f i c a t i o n s f t h e Hastelloy-N

a l l o y o r by e o ~ l t r o Pof t h e o x i d a t i o n p o t e n t i a l of t h e f u e l s a l t . The major areas of r e s e a r c h r e q u i r e d f o r c o m e r c i a b i z a t i o n of MSWs

would i n v o l v e improvement of t h e materials sf c o n s t r u c t i o n (Hastelloy-PS and g r a p h i t e ) , the d e s i g n of i n - l i n e i n s t r u m e n t a t i o n f o r high-temperature u s e , and t h e development of f u e l p r o c e s s i n g ( a t least f o r t h e end of reac-

t o r l i f e and p o s s i b l y a l s o f o r u s e on-line).

The major areas of develop-

ment i n v o l v e t h e scale-up of reactor components ( e o g c 9pumps) and the de-

sign and development of components t h a t were not p r e s e n t i n t h e MSW (e.g.,

steam g e n e r a t o r s and mechanical v a l v e s ) .

I n a d d i t i o n , we a ~ t k i -

p a t e t h a t t h e d e s i g n of some components such as t h e f u e l d r a i n system and t h e r e a e t o r c e l l w i t h i t s i n s u l a t i o n , h e a t i n g , and c o o l i n g requirements would be e x t e n s i v e l y modifed t o m e e t c u r r e n t l y u n s p e c i f i e d l i c e n s i n g requirements.

Another l a r g e area of development would be t h e c o n t r o l sf

t h e t e m p e r a t u r e s and flows i n t h e primary and secondary s a l t systems and i n t h e steam system t o avoid s a l t f r e e z i n g and e x c e s s i v e thermal stress. A l t e r n a t i v e l y , some c o ~ ~ ~ p o n e might nts be designed t o accommodate such freezing.

S t i l l a n o t h e r area of development would be advanced remote

maintenance t e c h n i q u e s , i n c l u d i n g t h e replacement o f components u s i n g remote p i p e c u t t i n g and welding.

112 The concept of t h e DMSR has emphas-ized p r o l i f e r a t i o n r e s i s t a n c e , and f u r t h e r d e s i g n e f f o r t s would be expected t o a d h e r e t o p r o l i f e r a t i o n resistance criteria.

However, no major areas have been i d e n t i f i e d i n

which t h e R&D requirements f o r a DNSR would be s u b s t a n t i a l l y d i f f e r e n t

from t h o s e f o r o t h e r v e r s i o n s of t h e MSR concept.

E s s e n t i a l l y t h e same

R&D program would be pursued as w a s planned f o r t h e MSBR.

of a low-power-density

The se1eetFon

c o r e f o r t h e DMSR h a s r e l i e v e d t h e r e q u i r e m e n t s f u r

c o r e g r a p h i t e ( e s p e c t a l l y f o r gas p e r m e a b i l i t y ) and has s i m p l i f i e d v e s s e l

The s e l e c t i o n of

d e s i g n (because g r a p h f t e replacement i s n o t r e q u i r e d ) .

a r e f e r e n c e DMSR w i t h o u t on-line f u e l p r o c e s s i n g h a s removed t h e develop-

ment of o n - l i n e p r o c e s s i n g from the expected c r i t i c a l p a t h f o r r e a c t o r development.

P r o c e s s i n g development should proceed, however, t o meet two

clQsely related objectives:

(I) development of on-line r e p r o c e s s i n g t o

o b t a i n the improved f u e l u t i l i z a t i o n of t h e break-even b r e e d e r DMSR opt i o n a s soon a s p o s s i b l e and ( 2 ) development ~f a process ( p r o b a b l y using

t h e same b a s i c technology) f o r e v e n t u a l c e n t r a l p r o c e s s i n g of f u e l from once-through DMSRs, p o s s f b l y i n s e c u r e f u e l s e r v i c e c e n t e r s .

5.1.3

Base program s c h e d u l e and c o s t s

h R&D b a s e program has been p r e s e n t e d i n some d e t a i l in t h e program plan.â&#x20AC;?â&#x20AC;&#x2DC;

me p r o j e c t e d c o s t s c h e d u l e ( i n 1 9 7 8 d o l l a r s ) f o r each major

development a c t i v i t y a n n u a l l y from 1980 t o 1994 and a s a t o t a l f o r 1995 through 2011 is g i v e n i n Table 30.

The complete b a s e program is pro-

j e c t e d t o c o s t about $700 m i l l i o n o v e r about 30 y e a r s . Some of t h e c o s t s are t a r g e t e d f o r elether t h e Molten-Salt t o r (MSTR) o r t h e d e m o n s t r a t i o n DXSR ( a s d i s c u s s e d

ita

T e s t Reac-

t h e f o l l o w i n g sec-

t i o n ) , w h i l e o t h e r c o s t s a p p l y g e n e r a l l y t~ t h e MSR develupment program. However, t h e s e c o s t s d o n o t i n c l u d e d e s i g n and c o n s t r u c t i o n c o s t s f o r t h e reactor plants The s c h e d u l e of f u e l p r o c e s s i n g technology development w a s set up f o r t h e c o n c u r r e n t development of on-line p r o c e s s i n g .

This s c h e d u l e c o u l d be

s t r e t c h e d i f t h e once-through c y c l e were chosen f o r the f i r s t DMSRs.

How-

e v e r , t h e development of p r o c e s s i n g technology i s an important g o a l of t h e program i n any e v e n t .

Table

30.

Projected

research

and

develnpnent

(Thousands

costs

for

1918 dollars)

12YS

61” 1,175 2,170

RZ” 2,070 2,;s”

95” :,xlo 2,455

93” 7”” 2, SF””

765 4”” 1,“““(1

6”” 35” 3,230

4”” 25” 3,670

75 22””

1,060 i,Y””

12,750 b 3,c25

0 3,590

51” 1,755

0 :,h12

xl” 1,560

ion i,i3*

955 30”

1,170 30”

1,502

507 6””

:69 5””

15” 6”Q

lib 65”

137

45”

1°C

15”

:co

13”

1””

13,968

2R,ZWb

1”” 7462

l4,Rl”

95,s:ot

L7,29hh

18,8?6

?C,ZSI

XSR base

development

program

114

5.2

Reactor B u i l d Schedule

5.2.1

Reactor sequence

I n a d d i t i o n t o t h e program of base techn~loggr~ u t k i n e dp r e v i o u s b y ,

a series of t h r e e developmental r e a c t o r s c u b m i n a t i ~ gw i t h a s t a n d a r d i z e d cmmercial p l a n t a r e proposed f o r c o n s t r u c t i o n . p l a n is g i v e n in c o n s i d e r a b l e d e t a i l in R e f .

The proposed development

102.

The development sequence would s t a r t with a p r e l i m i n a r y c o n c e p t u a l

d e s i g n for a k800-MWe DMSR (which would a c t u a l l y be t h e second r*eactor

i n t h e s e r i e s ) t o f u r t h e r d e f i n e t h e d e v e l o p m e ~ tproblems.

This would

be followed by c o n s i d e r a b l e c o m p ~ n e n tdevelopment ( i n t h e base technoPogy

program), a f t e r which the MSTR would be designed.

The MSTR i s proposed

eo be i n t h e 108- to 250-MWe s i z e r a n g e , which would employ components i n the one-fifth t~ full-scale range (based on the EO0Q-MMe conceptual

design).

Most of the MSTR eomponents would be tested i n an MSTR non-

r a d i o a c t i v e mockup b e f o r e the MSTR w a s a c t u a l l y assembled. During c o n s t r u c t i o n of t h e MSTR, component d e v e l o p m e ~ tand d e s i g n

of t h e p r o t o t y p e DMSR would proceed, w i t h d e t a i l e d d e s i g n and c o n s t r u c t i o n c o i n c i d e n t w i t h o p e r a t i o n of t h e MSTW.

This would a l l o w c l o s e feed-

back from WSTR e x p e r i e n c e i n t o t h e d e s i g n and c o n s t r u c t i o n o f the prototype.

F i n a l l y , d e t a i l e d d e s i g n of t h e f i r s t standard DMSR would proceed

d u r i n g cons$rUctiOll of t h e p r o t o t y p e

cOIlstrUCtion Of t h a t s"eaCtop Could

begin s h o r t l y a f t e r t h e p r o t o t y p e s t a r t e d o p e r a t i o n .

5.2.2

Schedule and c o s t s

A p o t e n t i a l r e a c t o r b u i l d s c h e d u l e i s g i v e n i n Fig. 28.

This i s t h e

same development s c h e d u l e as w a s proposed f o r t h e break-even b r e e d e r DMSR option.9

~n t h e l a t t e r case, t h e assumption was t h a t development of t h e

o n - l i n e r e p r o c e s s i n g system proceeded i n p a r a l l e l w i t h development of t h e reactors.

T h e r e f o r e , no c r e d i t can be taken f o r o m i t t i n g p r o c e s s d e s i g ~

i n t h e s c h e d u l e f o r t h e once-through DMSR,

However, removal of t h e pro-

cess development f r m the expected c r i t i c a l p a t h f o r r e a c t o r development removes a major s o u r c e of u n c e r t a i n t y and p o t e n t i a l d e l a y from t h e development schedule,

115

116

The c o n s t r u c t i o n cost of t h e MSTR way be e s t i m a t e d by u p d a t i n g t h e c o s t estimate f o r t h e MSTR p r e p a r e d i n 1 3 9 5 f o r t h e MSBR program.

Using

a c o n s t r u c t i o n materials and l a b o r i n c r e a s e of P2%/year g i v e s a m u l t i p l i e r of 1.4

and a c o s t f o r t h e MSTR i n 1978 of about $680 m i l l i o n .

A d e t a i l e d estimate of the c o s t of a l0QO-MWe DMSR based on a mature

technology is given i n

fol%owing s e e t i o n ,

From t h i s , we have e s t i m a t e d

t h e c o s t of a f i r s t s t a n d a r d i z e d DMSR by a p p l y i n g a f a c t o r of 1.5 t o a l low f o r i n c r e a s e d f i r s t - o f - a - k i n d

c o s t s and t h e c o s t of a l e a d

CQIIIIIW~C~~

p r o t o t y p e by a p p l y i n g a n o t h e r f a c t o r o f 1.5 t o a l l o w f o r i n c r e a s e d prototype costs,

Using t h i s p r o c e d u r e e t h e c o s t o f a POOO-Mk p r o t o t y p e DMSR

i s e s t i m a t e d to be $1470 miPliom and t h e f i r s t s t a n d a r d i z e d DMSR $980 million The p r o t o t y p e DMSR need n o t be as l a r g e a s 1800 MWe; t h e c o s t could be reduced some, f o r example, by b u i l d i n g a 5OO-We p r o t o t y p e w i t h two steam-generator Poops rather t h a n f o u r .

E s t i m a t i n g t h e p r o b a b l e c o s t of e x p e r i m e n t a l and p r o t o t y p e r e a c t o r s

fn advance of d e s i g n i s exseedingPy d i f f i c u l t .

The c o s t e s t f m a t e s pre-

s e n t e d were made by a s t a f f t h a t has had e x p e r i e n c e i n t h e d e s i g n and ope r a t i o n of e x p e r i m e n t a l r e a c t o r s , p a r t i c u l a r l y t h e MSRE.

glke

MSRE w a s

c o n s t r u c t e d and o p e r a t e d w i t h i n budget, which is a n i n d i c a t i o n t h a t t h e technology i s r e a s o n a b l y well understood and t h a t c o s t estimates f o r fut u r e r e a c t o r s are probably r e a l i s t i c , i f

R Q ~ absolutely

accurate.

Con-

verselye t h e proposed r e a c t o r s are a l a r g e step up i n s c a l e from t h e MSRE and would be s u b j e c t t o t h e v a g a r i e s of t h e l i c e n s i n g p r o c e s s f o r a new

reactor t y p e ,

These f a c t o r s i n t r o d u c e u n c e r t a i n t i e s i n t o t h e c o s t e s t i -

mates whish are beyond e v a l u a t i o n a t t h e p r e s e n t t i m e .

5.3

Economic Performance of Commercial DMSR

The p r o j e c t e d c o s t of power from a c ~ m ~ r ~DMSR i a l c a n be e s t i m a t e d

o n l y a p p r o x h a t e l y because t h e DMSR i s in t h e c o n c e p t u a l s t a g e and n o t even a d e t a i l e d c o n c e p t u a l d e s f g n has been prepared.

However, by t a k i n g

advantage of the s i m i l a r i t y o f t h e BMSR t o t h e MSBR ( f o r which a d e t a f l e d c o n c e p t u a l d e s i g n and c o s t estimate have been p r e p a r e d ) and by c a r e f u l l y comparing t h e DMSR w i t h EWW and c o a l - f i r e d power p l a n t s (which would have

... ........ .. . w.:. ..>

aa 7 some components i n common w i t h o r similar t o t h o s e of a DMSR), a reasona b l e c o s t estimate can be made. A c o s t estimate was prepared f o r t h e MSBR i n 1998 and a p p e a r s i n

Ref. 8.

These c o s t s were t a k e n ( f o r most a c c o u n t s ) a s t h e b a s i s f o r t h e

DMSR estimate using t h e f o l l o w i n g method.

TIE c o s t s were a d j u g t e a t o t a k e i n t o account t h e d i f f e r e n c e s P H ~

s i z e o r o t h e r r e q u i r e m e n t s f o r t h e DMSR.

For example, t h e r e a c t o r ves-

sel c o s t was i n c r e a s e d t o t a k e i n t o account t h e l a r g e r s i z e of t h e BMSR vessel0

The 1978 c o s t s were i n c r e a s e d by a m u l t i p l i e r based on t h e i n -

c r e a s e i n c o n s t r u c t i o n materials and l a b o r c o s t s from 1978 t o 1978. m u l t i p l i e r w a s c a l c u l a t e d t o be i n t h e range 2 . 3 t o 2.5; t i v e , t h e multiplier 2.5 was used.

'Fhe

t o be c ~ n s e % ~ a -

This r e p r e s e n t s a n annual r a t e of in-

c r e a s e of about 12X,

In a d d i t i o n , t h e c o s t s of a pressurized-water r e a c t o r (PWR), a b a i l i n g - w a t e r r e a c t o r (BUR),

and a c o a l - f i r e d

mated usimg t h e CONCEPT V code. "3 accounts were e s t i m a t e d based

CONCEPT e s t i m a t e s .

p l a n t i n 1978 were e s t i -

mere a p p r o p r i a t e , some DMSR c o s t the analogous account i n one of the

For example, t h e t u r b i n e - g e n e r a t o r

c o s t w a s based on

the c o a l - f i r e d p l a n t estimate because t h e same t y p e of s u p e r c r i t i c a l steam t u r b i n e would be used.

En r e p o r t i n g t h e r e s u l t s , the c o s t estimates f o r t h e PWK and ( i n the appendix) t h e c o a l - f i r e d

p l a n t are g i v e n f o r comparison.

(The c o s t s

of the BWR w e r e n o t s u b s t a n t i a l l y d i f f e r e n t from t h e PWR.)

The r e s u l t s of t h e estimates f o r c a p i t a l , monfuel o p e r a t i o n and maintenance, and decommissioning c o s t s are p r e s e n t e d i n t h e n e x t t h r e e sections.

The fuel c y c l e c o s t s were c a l c u l a t e d i n d e p e n d e n t l y and a r e

presented i n a f o u r t h section.

5.5.1

Capital costs

The b a s e s f o r t h e c a p i t a l c o s t estimates a r e g i v e n i n Table 31. Some minor a d j u s t m e n t s i n t h e code of a c c o u n t s used i n t h e 1970 MSBR c o s t

estimate were r e q u i r e d t o c~nfom-mt o t h e p r e s e n t code of accounts.

118 Table 31.

Bases f o r c a p i t a l o r investment c o s t estimates Excluded c o s t s

Bases Plant s i t e , Middletom (New England area)

eode of a c c o u n t s , NUS-531 and NUWG-8241, -0242, -0243 ( d i r e c t and i n d i r e c t a c c o u n t s ) Cost d a t e , 1978.6

Nuclear l i a b i l i t y i n s u r a n c e

Regulation codes a n d s t a n d a r d s , 1976

I n t e r e s t during c o n s t r u c t i o n

Evaporative c001511g

E s c a l a t i o n during c o n s t r u c t i o n

Commercial p l a n t s i z e (optimum), loo0 me

Contingency allowance

Cash flow, 1978 t o eomercial o p e r a t i o n i n 1988

OWtIerSi C Q s t s

C a p i t a l c o s t s i n 1978 d o bl ar s 1kWe

General and a d m i n i s t r a t i v e

%lTcPUdiI'lg expenses f o r t a x e s and p r o p e r t y i n s u r a n c e , spare parts, s t a f f training

site s e l e c t i ~ n ,and o t h e r owner-related expenses

S a l t and f u e l i n v e n t o r y , incIuding chemical p r o c e s s i n g system

c a p i t a l c o s t s estimated f o r t h e DHSR and t h e PWR by major ( t w o -

digit) accounts are given i n Table 32.

A f u r t h e r breakdown ( f o r three-

d i g i t a@co%mts), which a i s o i n c l u d e s the estimate f o r the coaa-fired p l a n t , i s g i v e n i n Appendix A (Table A.1).

A cumulative c a s h f l a w sched-

u l e f o r t h e DMSR (adapted t o the CONCEPT V c a s h â&#x201A;ŹHow s c h e d u l e ) i s a l s o g i v e n i n Appendix A (Table ik.2).

The c a p i t a l c o s t s of t h e f u e l t r e a t m e n t

f a c i l i t i e s are n o t i n c l u d e d h e r e b u t a r e taken i n t o account i n the f u e l

cycle c o s t (see Sect. 5 . 3 . 4 ) .

However, space and equipment f o r handling

t h e c o o l a n t s a l t a r e included i n the r e a c t o r p l a n t estimates. The DXSR i s e s t i m a t e d t o c o s t $653 m i l l i o n , o r about .$65O/kWe i n 1978 dollars.

This compares w i t h about $600/kWe f o r a PWR p l a n t and $38O/kWe

f o r a coal p l a n t without flue-gas c l e a n i n g .

Table 3 % .

C a p i t a l c o s t estimate O f commercial 1-GWe DKsa and PWR p l a n t s

(Expressed i n m i l l i o n s of 1998.0 d o l l a r s ) Ae eo un t No *

Item

DKSR

JPWR

Direct C o s t s 20

21 22 23

24 25 26

Land and Land r i g h t s S t r u c t u r e s and Lmpr ovement s Reactor p l a n t equipment Turbine p l a n t equipment E l e c t r i c p l a n t equipment Miscellaneous p l a n t equipment Naira condenser h e a t r e j e c t i o n system

1124 180 180 54 317

E4 ~

T o t a l d i r e c t costs

491

2 11%

139 113 44 1% 22

444

Indirect costs 91 92

COnstructiQIl S e H V i C e s Home o f f i c e e n g i n e e r i n g and s e r v i c e F i e l d o f f i c e e n g i n e e r i n g and s e r v i c e Total indirect costs

Total p l a n t c a p i t a l c o s t

95 53 34

53 30

162

153

65%

597

A d i s c u s s i o n OF some sf t h e i m p o r t a n t assmptiofts ana results f o r

t h e major a c c o u n t s f o l l o w s e

Account 21.

S t r u c t u r e s and improvements.

The primary and major

s t r u c t u r e i n t h e DMSR p l a n t i s t h e r e a c t o r containment b u i l d i n g .

While

l a y o u t s of i n t e r n a l areas would d i f f e r , comparable c o s t s w i t h t h e PWR

are expected.

An allowance h a s been e n t e r e d f o r p l a n t l i f e t i m e s t o r a g e

of r a d i o a c t i v e l y contaminated i t e m s i n c l u d i n g p r o v i s i o n s t o f a c i l i t a t e decommissioning o p e r a t i o n s .

The o t h e r s t r u c t u r e s p a r a l l e l t h e PWR s t r u c t u r e s i n c o s t . b i n e rooms are c o n s i d e r e d comparable.

The tur-

A s u p e r c r i t i c a l steam t u r b i n e f o r

t h e DMSR i s c o n s i d e r a b l y smaller than a PWR t u r b i n e , but s p a c e w a s allowed f o r extra p i p i n g and equipment t h a t may be r e q u i r e d t o adapt t h e super-

c r i t i c a l system t o a m o l t e n - s a l t steam g e n e r a t o r .

Space was a l s o allowed

f o r h a n d l i n g and s t o r i n g c o o l a n t s a l t fog. noma1 o p e r a t i o n s and f o r

120 s t o r i n g the blowdown mterialb t h a t would r e s u l t froan a

lnajQK

Steam leak

i n t h e s t e a m generator. Account 2 2 , heat-transfer

R e a c t o r p l a n t equipment,

The r e a c t o r and a s s o c i a t e d

s y s t e m c o s t s have been updated from t h e 1978 estimate u s i n g

a m u l t i p l i e r ~f 2.5.

About 10% of t h i s t o t a l ( a c c o u n t s 2 2 1 and 2 2 2 ) has

been added f o r engineered s a f e t y f e a t u r e s (which were n o t p r e v i o a a s ~ ycons i d e r e d ) and f o r l a r g e r s a l t v o l u n e s ,

%his amount a l s o c o v e r s e x t e r n a l

h e a t d i s s i p a t i o n equipment â&#x201A;Źâ&#x20AC;&#x2DC;or engineered s a f e t y f e a t u r e s .

Radioactive

waste h a n d l f ~ gi n the DMSR was e s t i m a t e d t a c o s t a b o u t t h e same a s radioa c t i v e waste p r o c e s s i n g f o r t h e PWR.

Fuel handling and s t o r a g e and maintenance equipment were updated from t h e MSBR estimate u s i n g t h e 2.5 m u l t i p l i e r .

An allowance was made f o r t h e

c o n t r o l f e a t u r e s n e c e s s a r y f o r making t h e p l a n t o p e r a t e

011

t h e salt-steam

cycle e Account 23.

T u r b i n e p l a n t equipment.

T h i s account p a r a l l e l s t h e

c o a l p l a n t c a s e , whish u s e s s u p e r c r i t i c a l steam-cycle equipment.

The

feed-heating account was i n c r e a s e d 50% t o allow f o r o p e r a t i n g d e s i g n feat u r e s p e s u l t a r to t h e Salt-%QQpa p p l i c a t i 0 R .

Account 2 4 .

E l e c t r i c plant equipment.

Except f o r p r o v i s i o n of about

25 Mwe of e l e c t r i c h e a t i n g a s s o c i a t e d w i t h t h e s a l t l o o p s , t h i s account i s s i m i l a r t o t h e PWR case.

Account 25.

Miscellaneous plant equipment.

The a u x i l i a r y steam sup-

p l y c o s t f o r t h e PWW h a s been i n c r e a s e d t o a d j u s t t h e PWR c o s t t o t h e DMSR basis Account 26.

Main condenser h e a t r e j e c t i o n system.

Design f o r t h e

coal p l a n t i s comparable t o t h e DMSW; t h e r e f o r e , t h e same c o s t s have been assumed e Accounts 91, 9 2 , and $ 3 ,

Indirect costs.

The DMSR c o s t s are based

on PWR c o s t s a d j u s t e d upward f o r t h o s e a c c o u n t s i n which h i g h e r l a b o r requirements are a n t i c i p a t e d .

5,3.2

ESo~fuel oloeration and maintenance c o s t

Estimates of n o n f u e l o p e r a t i o n and maintenance (06M) c o s t s of 2.82 wills/kh% a r e based on t h e s i n g l e - u n i t base-load

plant.

procedure

121 .. .. .....p..

,.,.,.,.,

i s based on t h e ONCOST code. I O 4

Annual expenses a r e d e r i v e d f o r s t a f f ,

maintenance m a t e r i a l s , s u p p l i e s and expenses, n u c l e a r liability i n s u r ante, G

o p e r a t i n g f e e s , and g e n e r a l a d m l n i s t r a t i v e a c t i v i t i e s .

Operation

and maintenance c o s t s a r e p r e s e n t e d f n 1938 d o l l a r s and are d i v i d e d i n t o f i x e d (demand r e l a t e d ) and v a r i a b l e ( e n e r g y - r e l a t e d )

components.

S t a f f r e q u i r e m e n t s a r e g i v e n i n Table 3% f o r a one-unit plant.

An-

n u a l c o s t s have been d e r i v e d from t h e O&M c o s t code w i t h m o d i f i c a t i o n s t o a d j u s t t h e maintenance-labor r a t i o t o 7 0 : 3 0 .

Estimates were t h a t a DKSR

might r e q u i r e major p l a n t work a t ten-year i n t e r v a l s ( o v e r and above PWR r e q u i r e m e n t s ) , f o r which maintenance l a b o r was i n c r e a s e d -50%.

summary o f annual 6 & M c o s t s i s g i v e n i n Table 3 4 , 5.3.3

Decommissioning and d i s p o s a l c o s t

Costs f o r decontamination and decommissioning of t h e f a c i l i t i e s would be i n c u r r e d a t t h e end of p l a n t l i f e .

A nuellear waste working g r o u p cow-

p r i s e d of DOE, Nuclear Regulatory Commission (NRC),

and Environmental Pro=-

t e c t i o n Agency (EPA) o f f i c i a l s i s working t o i d e n t i f y l e g i s l a t i v e needs

on h a n d l i n g n u c l e a r wastes,

P r e f e r e n c e i s wow g i v e n f o r dependence on

"engineered and n a t u r a l b a r r i e r s " f o r controP of o n - s i t e material a f t e r decommissioning w i t h f a l l - b a c k dependence on i n s t i t u t i o n a l c o n t r o l s f o r a " f i n i t e time.'o

The g r o u p a l s o o p t s f o r d i s m a n t l i n g a decommissioned

s i t e a f t e r a s h o r t decay p e r i o d , r a t h e r t h a n e i t h e r of two o t h e r s p t i s n s , w h i c h are entombing and mothballing n u c l e a r facilities.

The c o s t of d i s m a n t l i n g a DMSR i s expected t o be g r e a t e r t h a n f o r a n LWR because t h e a c t i v i t y l e v e l of components i n the primary c i r c u i t

i s higher.

A number of estimates of t h e decommissioning c o s t of LWRs

have been p r e p a r e d ; a s a b a s i s f o r our e s t i m a t e s , we have s e l e c t e d a r e p r e s e n t a t i v e r e c e n t (1978) estimate by t h e Tennessee V a l l e y A u t h o r i t y (TVA) f a r t h e i r Yellow Creek p l a n t e a r l y s i t e review.

The e s t i m a t e d de-

commi~sioning c o s t f o r t h i s p l a n t was $78 m i l l i o n f o r a BWR.

I f we as-

sume t h a t t h e c o s t f o r a DMSW would be about 10% g r e a t e r , t h e n t h e e s t i mated decommissioning c o s t f o r a DMSR would be about $86 m i l l i o n .

"This i s excluded d u r i n g c o n s t r u c t i o n p e r i o d when no f u e l i s on

site.

122 Table 33.

Staff requirement f o r one 1-GWe DNSR power p l a n t

Employee t y p e

Number

Plant manager's office Manager Assistant Quality assurance Environmental c o n t r ~ l P u b l i c relations Training Safety A d ~ a i n i s t r a t i o nand services Health s e r v i c e s Security

Subt 0 tal

1 i 3 1 1 1 1

13 1 66 89

Operatisms

2 33 Subtotal

Maintenance Supervision Crafts Peak maintenance, a n n u a l i z e d Subtotal

16 96

120

T e c h n i c a l and e n g i n e e r i n g Reactor Radiochemical I n s t r u m e n t a t i o n and control Technical support s t a f f Subtotal

1 2

2 17

Total

2 66

Less s e c u r i t y

280

Less s e c u r i t y and peak mintemance

184

.v .....< ..... .. ..." -3

123

Table 3 4 .

Summary of annual nonfuel Q&M c o s t s f o r base-load steam-electric power p l a n t s i n 1978. oa

Plant type

DMSR w i t h e v a p o r a t i v e c o o l i n g towers

Number of u n i t s per s t a t i o n

Thermal input: per unit, 'MWt

2270

P l a n t n e t heat r a t e

7755

Plant net efficiency, 2

44 00

Power o u t p u t , net,

PO00

Annual n e t g e n e r a t i o n , million lewh

6570

Pla n t fac to P

8.75

Annual costs, thousands

dollars

S t a f f , 266 persons a t $23,412

6228

Maintenance material Fixed Variable S u p p l i e s and expenses F i x e d , plant Variable, plant I n s u r a n c e and fees Comercial l i a b i l i t y bnsuranee Government l i a b i l i t y i n s u r a n c e R e t r o s p e c t i v e premium P n s p e ~ t i of e~e~s and expenses A d m i n i s t r a t i v e and g e n e r a l Total f i x e d costs T o t a l v a r i a b l e costs T o t a l annual OdM c o s t s

6555 6555 6

3317 3800

317 488 2 84 18

6 100 2367 18,500

317 18,875

Unit c o s t s , m i l l s / k % ( e ) Fixed u n i t O&M c o s t s V a r i a b l e u n i t O&M c o s t s T o t a l u n i t O&M c o s t s

2*75 0.07 2.82

aExcludes the s a l t i n v e n t o r y l o s s e s ; assumes nuclear i n s u r a n c e a t LWR sates.

124

l a r g e u n c e r t a i n t y i s p r e s e n t i n this estimate, of c o u r s e , because of t h e l i m i t e d e x p e r i e n c e i n decommissioning.

However, t h e c o s t sf decommission-

ing does n o t appear t o b e a l a r g e f r a c t i o n of t h e c o s t of t h e c o n s t r u c -

tion, and t h e p r e s e n t worth of e x p e n d i t u r e s t o be made i n t h e f u t u r e i s small.

'Fhe p r e s e n t worth of t h e estimated c o s t of decommissioning a DMSR

40 y e a r s a f t e r s t a r t - u p , d i s c o u n t e d t o t h e s t a r t - u p d a t e a t a 4.5% d i s count f a c t o r , would be about $15 m i l l i o n .

5.3*4

Fuel cvcle costs

A% t h i s s t a g e of development and o p t i m i z a t i o n of a ~ n ~ e - t h r o u g DMSR, h s e v e r a l assumptions are n e c e s s a r y i f a f u e l c y c l e c o s t i s t o be e s t i m a t e d . One assumption 1s t h a t t h e i n i t i a l f u e l charge w i l l c o n s i s t of 7 4

mole 2 7LiF, 16.5 mole % BeF2, 8.23 mole 2 ThF4, and 1.27 mole X UF, p l u s UP3.

If, as seems r e a s o n a b l e , an afkswanee is made for an additional 2%

o f molten f u e l i n t h e d r a i n t a n k , t h e i n i t i a l fuel s o l v e n t w i l l r e q u i r e

149 m e t r i c t o n s of ThF,, BO2"

E13 m e t r i c t o n s of L i F , and 45.6 metric t o n s of

I n a d d i t i o n , 23.5 metric t o n s of UP, e n r i c h e d t o 20% i n 23% i s re-

q u i r e d ; t h i s i s e q u i v a l e n t t o 804 m e t r i c t o n s of U308 and would r e q u i r e

8.05

x 105

swu f o r

i t s enrichment.

Tke p u r i f i e d f u e l d e l i v e r e d t o t h e re-

a c t o r s t o r a g e t a n k would consist of 33.1 metric t o n s (7.30

~ b 05 )

material T o t a l c o s t of the i n i t i a l f u e l i s based on U308 a t $35/lb, s e p a r a t i v e

work a t $SO/SWU, BeF2 a t $ 1 5 / l b , thorium at $15/kb Tho2$ and SLiF a v a i l a b l e a t $ 3 J g of contained

ki,

AII additional $ 2 / l b of t e t r a f l u o r i d e h a s

been a l l o t t e d f o r c o n v e r s i o n of Tho2 to n F 4 and f o r CQrWerSiQn of UPg to

UF$.

The fuel, wade by mixing t h e powdered i n g r e d i e n t s , must be p u r i f i e d

in the molten s t a t e b e f o r e use (as d e s c r i b e d in p r e v i o u s s e c t i o n s ) ,

Given

the component f l u o r i d e s a t t h e p r i c e s above, t h e assunaption i s t h a t t h i s p u r i f i c a t i o n can be performed f o r $ $ / l b of f i n i s h e d product. With t h e s e assumptions, t h e t o t a l c o s t of t h e i n i t i a l DMSR f u e l charge i s near $225 m i l l i o n (see Table 35).

* his

If w e assme t h a t the annual

i s t h e o f f i c i a l p r i c e f o r s m a l l q u a n t i t i e s of ~9,99+ a It i s almost c e r t a i n l y t o o h i g h ( p r o b a b l y f i v e - f o l d ) l i t h i u m were a c t u a l l y used i n such q u a n t i t i e s .

use i n PWRS.

~ fio r if

125 ...... ...,* x..:.w

Table 3 5 .

Cost of i n i t i a l f u e l charge f o r once-through DMSR ( 3 3 1 metric t o n s f u e l )

cost

Fuel

(dollars)

Fuel s o l v e n t Mater tals 'LiF (30.46 metric t o n s l i t h i u m a t $3/g) ThP4 ( 1 2 7 . 6 8 metric t o n s Tho2 a t $ â&#x201A;Ź 5 / P b ) BeP2 (45.60 metric t o n s a t $ 1 5 / l b )

Uranium (883.8 metric t o n s U308 a t $ 3 5 / l b ) S e p a r a t i v e work ( 8 . 0 5 x 105

swu

a t $80)

Conversion and p u r i f i c a t i o n msrium (148.96 metric t o n s ThP4 a t $ 2 / l b ) uranium (23.50 metric t o n s U F ~a t $ 2 / % b ) F u e l m i x t u r e (331.2 metric t o n s a t $ 6 / l b ) T o t a l c o s t of i n i t i a l f u e l Annual c h a r g e ( 1 2 % )

u s e c h a r g e is 122, t h i s i n i t i a l f u e l c o n t r i b u t e s $26.9 m f l l i o n / y e a r t o the f u e l cycle cost. A once-through DMSR must add uranium a t more o s l e s s r e g u l a r i n t e r -

v a l s over its o p e r a t i n g l i f e t i m e .

Though o t h e r modes of a d d i t i o n a r e pos-

sible, f o r t h i s assessment w e assumed t h a t t h e uranium a d d i t i o n s will b e

made as a l i q u i d 'LiF-UFk a b o u t 540OC) e

m i x t u r e c o n t a i n i n g 3 0 mole % UF4 ( m e l t i n g a t

Such a d d i t i o n s would a p p r e c i a b l y i n c r e a s e t h e 'LFF

c e n t r a t i o n of t h e f u e l .

con-

Adjustment of t h e UP3/UF4 r a t i o i s assumed t o be

done by i n s i t u r e d u c t i o n of UF4 w i t h m e t a l l i c b e r y l l i u m .

The f u e l stream

i s assumed t o be t r e a t e d (once i n each 1008 full-power d a y s ) w i t h an anhydrous HF-Hz m i x t u r e t o remove i n a d v e r t e n t o x i d e c o n t a m i n a t i o n ; t h e res u l t i n g o x i d a t i o n of UF3 i s managed by a d d i t i o n a l . r e d u c t i o n w i t h b e r y l l i m . With t h i s mode of o p e r a t i o n , BeF2 e q u i v a l e n t ts n e a r l y 6% of t h a t i n t h e 4

o r i g i n a l f u e l c h a r g e would be added over t h e r e a c t o r l i f e t i m e .

Such ad-

d i t i o n s of ?LiF and BeF2 d i l u t e t h e f u e l and a p p r e c i a b l y i n c r e a s e i t s volume s o t h a t a n i n c r e a s i n g ( t h o u g h r e l a t i v e l y s m a l l ) f r a c t i o n of t h e

__.

...... . i..

f u e l would remain in t h e reastor d r a i n tank.

me d i s s o l v e d p a r a s i t i c

neutron a b s o r b e r s , of c ~ u r s e ,would a l s o be d i l u t e d .

A t t h i s s t a g e of

DMSR deVelOpment, no d e t a i l e d Q p t i m i Z a t i O n f Q P such f u e l d i l u t i o n h a s been made.

For t h i s assessment, we assumed t h a t , as a consequence of

t h i s d i l u t i o n e f f e c t by f u e l maintenance, the uranium

addition^ shown in

Table I S p l u s uranium (at. 20% e n ~ i c h w e n t )and thorium e q u i v a l e n t t o 3% of t h e i n i t i a l i n v e n t o r y would be r e q u i r e d over t h e 38-year o p e r a t i n g

life o f t h e r e a c t o r . of 7LiF-ThP4

Thorium i s assumed t o be added as a molten m i x t u r e

contair~ing 28 mole % ThFq ( m e l t i n g p o i n t s f 570째C).

Table 36 shows t h e average annual c o s t of t h e s e a d d i t i o n s and f u e l maintenance.

C o s t s of ThF4, UPh, L i p , and s e p a r a t i v e work a r e t h o s e de-

scribed previously.

Metallic b e r y l l i u m i s assumed t o c o s t $T5/lb.

of preparing t h e 'LiF-ThFb

cost

and 7 L i F - U F ~ m i x t u r e s was assumed t o be $20/lb

( p l u s t h e c o s t of t h e s o l i d raw m a t e r i a l s ) .

T a b l e 36. Average annual c o s t 5f f u e l a d d i t i o n s and maintenance cost

(dollars)

E2 7 There are no d e t a i l e d estimates of t h e c a p i t a l o r o p e r a t i n g c o s t s of t h e equipment f o r W-H2 t r e a t m e n t t o remove (I2-from t h e s m a l l b a t c h e s of

fuel.

For t h i s a s s e s s m e n t , we assumed t h a t t h e c a p i t a l c o s t i s $15 x 106

and t h a t i t s o p e r a t i o n c o s t s $500,00O/year. A s a consequence of t h e assumptions and t h e estimates d e s c r i b e d , t h e c o s t of producing 6.57 x 10â&#x20AC;&#x2122; k m j y e a r ( o p e r a t i o n a t 75% p l a n t f a c t o r ) ap-

p a r e n t l y a v e r a g e s $34,650,000, and t h e r e s u l t i n g f u e l c y c l e c o s t i s about

5.3 mills/krn. Note t h a t , i f t h e p r i c e of 7 L i were lowered by f i v e - f o l d

[ t o SO.4O/g

( $ 2 7 2 / l b ) l , t h e r e s u l t i n g f u e l c y c l e c o s t f o r t h e once-through DMSR would

fall t o s l l i g h t b y below 4 mills/kkJh. 5.3.5

N e t power cost

Because t h e r e t u r n on t h e p l a n t c a p i t a l investment woslPd be a subs t a n t i a l f a c t o r i n t h e n e t c o s t of power from a DMSR and because a number

of terms t h a t would be important i n a commercia% p l a n t were omitted i n dev e l o p i n g t h e capital c o s t estimate, p r o j e c t i n g a p o t e n t i a l n e t c o s t f o r DMSR power i s n o t a p p r o p r i a t e .

S u b s t a n t i a l l y more d e s i g n and development

would be r e q u i r e d t o s u p p o r t a r e a s o n a b l y r e l i a b l e estimate.

However, the

p r e v i o u s d i s c u s s i o n s s u g g e s t t h a t t h e c o s t of power from a BMSR

not

be g r e a t l y d i f f e r e n t t h a n that from o t h e r nuclear systems.

5.4

Licensing

Mthough two e x p e r i m e n t a l MSRs have been b u i l t and o p e r a t e d i n the United S t a t e s under government ownership, none h a s e v e r been s u b j e c t e d t o

formal l i c e n s i n g o r even d e t a i l e d review by t h e NRC.

A s a consequence,

t h e q u e s t i o n of l i e e n s a b i l i t y o f MSRs remains open; t h e NRC h a s not y e t i d e n t i f i e d t h e major l i c e n s i n g i s s u e s and the concept has n o t been eons i d e r e d by v a r i o u s p u b l i c i n t e r e s t o r g a n i z a t i o n s that a r e o f t e n involved i n n u c l e a r p l a n t l i c e n s i n g procedures.

F u r t h e r , t h e l i c e n s i n g experience

of s o l i d - f u e l e d r e a c t o r s can be used as o n l y a g e n e r a l g u i d e because of s i g n i f i c a n t fundamental d i f f e r e n c e s between t h o s e systems and MSB.

Pre-

sumably, MSRs would be r e q u i r e d t o comply w i t h t h e intents, r a t h e r t h a n

128

t h e l e t t e r , of NRC r e q u i r e m e n t s , p a r t i c u l a r l y where methods of compliance

are concept-specific. Any s p e c i a l i s s u e s t h a t might a r i s e from p u b l i c c o n s i d e r a t i o n of an

MSR l i c e n s e probably would be c l o s e l y a s s o c i a t e d w i t h t h o s e f e a t u r e s of

t h e r e a c t o r concept t h a t a f f e c % i t s s a f e t y and environmental a t t r i b u t e s . A number of t h e s e f e a t u r e s and a t t r i b u t e s have been i d e n t i f i e d i n e a r l i e r S ~ C ~ ~ O R SOwe ,

major d i f f e r e n c e between more c o n v e n t i o n a l r e a c t o r s and

MSRs i s i n t h e confinement of r a d i o a c t i v e f u e l and f i s s i o n products.

b a r r i e r s t o fission-product

The

release i n LWRs are ( I ) t h e f u e l element clad-

dingp ( 2 ) t h e r e a c t o r c o o l a n t p r e s s u r e b o u n d a ~ y(RCPB) ( i o e e 9t h e primaryPoop v e s s e l s , components, and p i p i n g ) , and ( 3 ) t h e r e a c t o r c ~ ~ t a i r n e n t , T h i s arrangement r e l i e s h e a v i l y on t h e ECCS t o prevent c l a d d i n g f a i l u r e i n t h e e v e n t of c o o l a n t l o s s by f a i l u r e of t h e RCPB.

Iqithsut adequate

E N S performance, a f a i l u r e of t h e WCPR conceivably @ o d d l e a v e t h e f i s s i o n p r o d u e t s w i t h o n l y one l e v e l of confinement i n t a c t .

A d i f f e r e n t s i t u a t i o n would p r e v a i l in an MSR because t h e f i s s i o n -

p ~ ~ d u cconfinement t b a r r i e r s are d i f f e r e n t .

The r e l e v a n t b a r r i e r s i n an

MSR are (1) t h e RCPB, ( 2 ) t h e s e a l e d r e a c t o r c e l l s o r p r i m a r y c s n t a i m e n t ,

and ( 3 ) t h e r e a c t o r containment b u i l d i n g o r secondary containment.

Be-

cause t h e f u e l i s a c i r c u l a t i n g l i q u i d t h a t i s also t h e primary ~ o ~ l a n t , t h e r e i s no t h i n f u e l clad t h a t c o u l d f a i l q u i c k l y on EQSS of c o ~ l f n go r

in a r e a c t o r power/temperature t r a n s i e n t .

~ h r a s , an e n t i r e class of po-

t e n t i a l a c c i d e n t s could be e l i m i n a t e d from t h e l i c e n s i n g c o n s i d e r a t i o n . F a i l u r e of t h e RCPB i n an MSR would cause no s h ~ r t - t e r m t h r e a t t o e i t h e r

of t h e remaining t w o b a r r i e r s t o f i s s i o n - p r o d u c t

release.

'Flze u l t i m a t e

requfrements f o r l o n g e r - t e r n p r o t e c t i o n of t h e f i s s i o n - p r o d u c t b a r r i e r s cannot be d e f i n e d without e x t e n s i v e system d e s i g n and s a f e t y a n a l y s i s , b u t p r e l i m i n a r y c o n s i d e r a t i o n s suggest 'chat t h e requirements may n o t be extensi%re, Although r a d i o a c t i v e materials would have t h r e e l e v e l s of confinement d u r i n g n o r m 1 o p e r a t i o n , a d i f f e r e n t c o n d i t i o n could e x i s t d u r i n g maintenance o p e r a t i o n s t h a t r e q u i r e d opening of t h e primary containment,

*F a i l u r e

of the WCPR is one sf t h e mechanisms f o r initiating a Psssof-coolant a c c i d e n t (LOCA>.

129

p a r t i c u l a r l y i f s u c h a c t i v i t i e s were undertaken a f t e r an RCPB f a i l u r e . However, i n a shutdown s i t u a t i o n , s u b s t a n t i a l confinement can be a c h i e v e d through a c c e s s l i m i t a t i o n and c o n t r o l l e d v e n t i l a t i o n because, a s shown by MSRE e x p e r i e n c e , f i s s i o n p r o d u c t s are n o t r e a d i l y r e l e a s e d and d i s -

persed from s t a g n a n t s a l t .

Thus, whether f i s s i o n - p r o d u c t confinement

would be a n e t favorable o r u n f a v o r a b l e f a c t o r f o r a BMSR i n a l i c e n s i n g proceeding i s n o t c l e a r a t t h i s t i m e . A t t h e end of r e a c t o r l i f e , a DMSR without f u e l p r o c e s s i n g would

c o n t a i n t h e e n t i r e f i s s i o ~ - p r ~ d u ci nt v e n t o r y a s s o c i a t e d w i t h t h e 30-year o p e r a t i n g h i s t o r y of t h e p l a n t .

Some of t h e v o l a t i l e n u c l i d e s , e s p e c i a l l y

8 5 K r and 3H, would have been accumulated i n s t o r a g e c o n t a i n e r s o u t s i d e t h e primary c i r c u i t , and t h e n o b l e metals would have p l a t e d o u t on s u r f a c e s in t h e primary c i r c u i t .

The i n v e n t o r i e s of t h e s e n u c l i d e s , which would n o t

be s t r o n g l y a f f e c t e d by n u c l e a r burnup, would be about t h e same a s t h o s e produced i n a s o l i d - f u e l duty factor.

r e a c t o r wPth the same thermal power l e v e l and

However, because t h e BMSR would g e n e r a t e o n l y about two-

t h i r d s a s much thermal power as a n EWR f o r t h e same e l e c t r i c a l o u t p u t , i t would produce a ~ o ~ ~ e s p o n d i n g smaller ly i n v e n t o r y of fission p r o d u c t s . Most of t h e o t h e r f i s s i o n p r o d u c t s and all the transuranium nucltcles

wou%dl remain w i t h t h e f u e l s a l t i n a DMSR.

The i n v e n t o r i e s of t h e s e mate-

r i a l s would be f u r t h e r reduced by n u c l e a r burnup r e s u l t i n g from exposure of t h e n u c l i d e s t o t h e neutron f l u x i n t h e r e a c t o r core.

would be p a r t i c u l a r l y i m p o r t a n t f o r t h e h i g h - c r o s s - s e c t i o n

a s t h e major plutonium i s o t o p e s .

This e f f e c t

nucPides such

Consequently, t h e n e t p r o d u c t i o n of plu-

tonium would be much smaller f o r a DMSR t h a n f o r a comparable s o l i d - f u e l r e a c t o r , b u t t h e p r o d u c t i o n of h i g h e r a c t i n i d e s would be much g r e a t e r bec a u s e of t h e long e f f e c t i v e f u e l exposure t i m e . Although a DMSW W Q U produce I~ a much smaller t o t a l i n v e n t o r y of some

i m p o r t a n t n u c l i d e s o v e r i t s l i f e t i m e t h a n an LWR, t h e a c t u a l i n - p l a n t

in-

v e n t o r y c o u l d be s u b s t a n t i a l l y h i g h e r f o r t h e DMSR because t h e r e would be

no p e r i o d i c removal d u r i n g r e f u e l i n g o p e r a t i o n s .

(There would a l s o be no

major shipments of h i g h l y r a d i o a c t i v e s p e n t f u e l from t h e p l a n t d u r i n g i t s l i f e t i m e arid no o u t - o f - r e a c t o r

t h e f i n a l shutdown.)

s t o r a g e of suck materials u n t i l a f t e r

Thus, i f a major release of in-p%ant r a d i o n u c l i d e s

could o c c u r , t h e consequence might be more s e r i o u s i n a DMSR than i n an

LWR.

However, c o n s i d e r i n g t h e mechanisms and p r o b a b i l i t f e s f o r release

e v e n t s , along w i t h t h e

C C X I S ~ ~ U ~ Rwould C ~ ~ ,

be n e c e s s a r y in a s s e s s i n g any

e f f e c t on system licensability. Before amy MSW i s l i c e n s e d , we probably w i b k need t o d e f i n e a com-

plete n e w spectrum of p o t e n t i a l t r a n s i e n t s and a c c i d e n t s and t h e i r a p p l i c a b l e L n i t i a t i n g e v e n t s t h a t are t o be t r e a t e d i n s a f e t y a n a l y s i s r e p o r t s . Some of t h e more i m p o r t a n t s a f e t y - s i g n i f i c a n t e v e n t s f o r a n MSR were went i o n e d e a r l i e r , b u t even r o u t i n e o p e r a t i o n a l e v e n t s may have a d i f f e r e n t o r d e r of importance f o r t h i s r e a c t o r concept.

FOP example, moderate re-

a c t o r power d i s t u r b a n c e s would n o t be v e r y important because one o â&#x201A;Ź t h e p r i n c i p a l consequences, f u e l c l a d d i n g f a i l u r e , i s a nonevent i n a n MSR. Conversely, a small l e a k of r e a c t o r c o o l a n t would be an important event because of t h e h i g h Bevel of r a d i o a c t i v i t y i n t h e MSR

C Q Q ~ ~ I K ~ .

The above examples of s i g n i f i c a n t d i f f e r e n c e s between MSRs and o t h e r

l i c e n s e d reactors i l l u s t r a t e why a s u b s t a n t i a l d e s i g n and analysis e f f o r t would be r e q u i r e d

f i r s t t o e s t a b l i s h l i c e n s i n g c r i t e r i a f o r MSRs

i n g e n e r a l and a DMSR i n p a r t i c u l a r and second

e v a l u a t e MSR l i c e n e -

a b i l i t y i n r e l a t i o n t o t h a t of o t h e r r e a c t o r t y p e s .

This r e q u i r e m e n t ,

w i t h no a p r i o r i a s s u r a n c e t h a t an MSR could be l i c e n s e d , makes i t un-

l i k e l y t h a t p r i v a t e o r g a n i z a t i o n s i n t h e United S t a t e s would u n d e r t a k e

t h e development and commercfaPSzation of KSRs.

I n s t e a d , i f such develop-

ment were pursued, g o v e ~ m e n tfunding probably would be r e q u i r e d , a t

l e a s t u n t i l t h e l i c e n s i n g issues c o u l d be r e s o l v e d and near-commercial u n i t s could he c o n s t r u c t e d .

131

~.........* .-,p

... ......*:.

SUMMARY AND CONCLUSIONS

me technology of MSRs w a s under development w i t h

gsverment

U.S.

funding from 1947 t o 1976 w i t h a nominal one-year i n t e r r u p t i o n from 11993 t o 1974.

Although no s i g n i f i c a n t e f f o r t t o commercialize MSRs w a s in-

volved i n t h i s work, a v e r y p r e l i m i n a r y c o n c e p t u a l d e s i g n w a s g e n e r a t e d for

10QO-We MSBR, and some a l t e r n a t e f u e l c y c l e s were examined.

The

c u r r e n t s t u d y of denatured PIS& w a s supported by t h e program (NASAP) t o i d e n t i f y , c h a r a c t e r i z e , and assess p s o 8 i g e r a t i o n - r e s i s t a n t

alternatives

t o c u r r e n t l y p r o j e c t e d n u c l e a r power systems.

I n p r i n c i p a l , MSRs could be o p e r a t e d w i t h a number of f u e l cycles r a n g i n g from plutonium-fueled p r o c ~ u c t i o nof d e n a t u r e d 2 "u, breeding with

t o break-even

f u e l , t o high-performance c o n v e r s i o n of r h o r i m t o

2 3 3 ~w i t h d e n a t u r e d 2 3 5 ~makeup f u e l .

me l a s t of these c y c l e s c u r r e n t l y

a p p e a r s t o be t h e most a t t r a c t i v e and i s t h e one chosen f o r c h a r a c t e r i z a t i o n i n t h i s study.

The f u e l c y c l e would i n v o l v e an i n i t i a l l o a d i n g of

denatured 2 3 5 ~ ;o p e r a t i o n f o r ~ C y I e a r s ( a t 75% c a p a c i t y f a c t o r ) w i t h 2 3 5 ~ makeup, mo f u e l d i s c h a r g e , and no chemical t r e a t m e n t f o r f i s s i o n - p r o d u c t removal; and end-of-life

s t o r a g e B d i s p o s a 1 of t h e spent f u e l .

me r e s o u r c e

u t i l i z a t i o n of t h i s c y c l e c o u l d be s i g n i f i c a n t l y enhanced by e n d - o f - l i f e r e c o v e r y of t h e denatured uranium i n t h e f u e l s a l t v i a f l u o r i n a t i o n .

6.1

Reference-Concept DMSR

The d i f f e r e n c e s between a DMSR and t h e c o n c e p t u a l d e s i g n MSBR inv o l v e p r i m a r i l y t h e r e a c t o r c o r e and t h e f u e l cycle. t h e primary c i r c u i t (e.g.,

Thus, t h e r e s t of

pumps and h e a t exchangers) and t h e b a l a n c e

of t h e p l a n t would be v e r y s i m i l a r f o r b o t h c o n c e p t s , and t h e d e s c r i p t i o n s developed f o r t h e MSBR a r e presumed t o be a p p l i c a b l e t o t h e DMSB. Minor v a r i a t i o ~ st h a t wight be a s s o c i a t e d w i t h d e s i g n o p t i m i z a t i o n are n o t considered.

The r e a c t o r v e s s e l f o r t h e DMSR, about 18 m i n d i a m e t e r and 10 m h i g h , would be s u b s t a n t i a l l y l a r g e r t h a n t h a t f o r t h e high-performance breeder.

This would permit the Pow power d e n s i t y r e q u i r e d t o a l l o w a

38-year l i f e expectancy f o r t h e r e a c t o r g r a p h i t e and would a l s o reduce

. .

,* .. ..! . ,A

3-32 neutronic l o s s e s t o 233Pa.

Other e f f e c t s of t h e %ow power d e n s i t y in-

c l u d e reduced poisoning e f f e c t s from in-core

f i s s i o n p r o d u c t s and a n in-

c r e a s e d f i ss il e i n v e n t o r y a The r e a c t o r c o r e would ~ ~ n s i of s t a c e n t r a l r e g i o n c o n t a i n i n g 20

v o l Z f u e l s a l t and a l a r g e r s u r r o u n d i n g zone c o n t a i n i n g 13 v s l % s a l t . Neutron m o d e r a t i o n would be provided by v e r t i c a l c y l i n d r i c a l unclad graphi t e G o l o g s w i t h f u e l s a l t flowing upward thrseaih c e n t r a l p a s s a g e s and 911

between t h e moderator elements.

me c o r e would be'surrounded

f i r s t by

s a l t plenums and @ ~ p a n s i sop~a c e s and t h e n by a g r a p h i t e r e f l e c t o r and

the r e a c t o r v e s s e l . With t h i s c o r e d e s i g n , a 1-GWe p l a n t would r e q u i r e a n i n i t i a l f i s -

s i l e l o a d i n g of 3450 kg 235U a t 28% enrichment ( e x t r a c t a b l e from a b o u t 870 s h o r t t o n s sf U,O,>.

Over 30 y e a r s a t 7 5 % c a p a c i t y f a c t o r , t h e f u e l

makeup r e q u i r e m e n t would be about 4478 kg of 20% e n r i c h e d 235U (from 1125 s h o r t t o n s of ~ ~ 0 ,f o) r a l i f e t i m e U308 demand of 2800 s h o r t tons,

Row-

ever, a t t h e end of plant l i f e , t h e f u e l s a l t would c o n t a i n d e n a t u r e d f i s s i l e uranium (233U and 235U) e q u i v a l e n t t o at l e a s t $00 s h o r t t o n s of

n a t u r a l U308.

If t h i s material c o u l d be r e c o v e r e d ( e o g o pby f l u o r i n a t i o n )

and reenriched, i t would s u b s t a n t i a l l y r e d u c e t h e n e t f u e l r e q u i r e m e n t of

t h e BMSR. P r e l i m i n a r y c a l c u l a t i o n s of t h e k i n e t i c and dyrtamic c h a r a c t e r i s t i c s of the BbfSB system i n d i c a t e t h a t i t would e x h i b i t h i g h l e v e l s of controll a b i l i t y and s a f e t y .

'Fhe system would a l s o possess i n h e r e n t dynamic sta-

b i l i t y and w o u M r e q u i r e o n l y modest amounts o f r e a c t i v i t y c o n t r o l c a p $ iI i

t ye A f i r s t - r o u n d a n a l y s i s of t h e t h e r m a l - h y d r a u l i c C h a r a c t e r i s t i c s of

t h e DMSR c o r e c o n c e p t u a l d e s i g n i n d i c a t e d t h a t t h e c y l i n d r i c a l moderator elements would be a d e q u a t e l y cooled by t h e flowing f u e l s a l t and t h a t reasonable s a l t t e m p e r a t u r e d i s t r i b u t i o n s c o u l d be achieved w i t h some o r i f i c i n g of t h e f u e l f l o w passages.

While some u n c e r t a i n t i e s a b o u t t h e

d e t a i l e d flow behavior i n t h e salt-graphite

system remain which would

have to be resolved by developmental t e s t i n g , t h e r e s u l t s would n o t be expected t o a f f e c t t h e fundamental f e a s i b i l i t y o f t h e concept.

Tke primary f u e l s a l t would be a molten m i x t u r e of E i F and BeF2 eont a i n i n g ThFb9 d e n a t u r e d UF4, and some PuF3.

L i t h i m highly enriched i n

133

.. .;.:w ........

i s o t o p e 0 9 9 . 9 9 % ) would be r e q u i r e d , and t h e mixture would gradu-

the

a l l y b u i l d up a s i g n i f i c a n t i n v e n t o r y of f f s s i o ~ - p ~ o d uand ~ t higheractinide fluorides. b

This m i x t u r e would have adequate n e u t r o n i c

physical $

t h e r m a l - h y d r a u l i c , and chemical c h a r a c t e r i s t i c s t o f u n c t i o n f o r 30 y e a r s a s a f u e l and primary r e a c t o r c o o l a n t .

Routine maintenance of t h e s a l t would be r e q u i r e d t o keep some of t h e uranium i n t h e p a r t l y reduced U 34s t a t e f o r t h e p r e f e r r e d chemical behavior. Although s e v e r e contamination of t h e s a l t w i t h oxide i o n c o u l d l e a d

t o p r e c i p i t a t i o n of plutonium and uranium o x i d e s , t h e s o l u b i l i t y of these o x i d e s i s h i g h enough t h a t a n i n c r e a s e i n oxide i o n c o n c e n t r a t i o n probably could be d e t e c t e d and stopped b e f o r e suck p r e c i p i t a t i o n o c c u r r e d .

In ad-

d i t l i o n , c l e a n u p of t h e s a l t on a r o u t i n e b a s i s t o m a i n t a i n t h e r e q u i r e d

Bow o x i d e c o n c e n t r a t i o n would be r e l a t i v e l y e a s y * '%he f u e l s a l t is a l s o h i g h l y c o m p a t i b l e , b o t h c h e m i c a l l y and p h y s i c a l l y , w i t h the proposed s t r u c t u r a l a l l o y , Hastelloy-N,

and w i t h t h e prsposed unclad g r a p h i t e mod-

eratore The r a d i a t i o n r e s i s t a n c e of t h e f u e l salt i s w e l l e s t a b l i s h e d , and

no r a d i a t i o n decomposition would be expected except a t v e r y low temperat u r e s (below -I0OaC>.

The noble-gas f i s s i o n p r o d u c t s , xenon and k r y p t o n ,

are o n l y s p a r i n g l y s o l u b l e i n f u e l s a l t and would be removed csntinususEy d u r i n g r e a c t o r o p e r a t i o n by a helium s p a r g i n g system.

P o r t i o n s of some

o t h e r v o l a t i l e f i s s i o n p r o d u c t s might a l s o be removed by t h i s system. Another c l a s s of f i s s i o n p r o d u c t s , t h e n o b l e and seminoble metals, would be expected t o exist in t h e m e t a l l i c s t a t e and t o plate o u t m o s t l y on

metal s u r f a c e s i n t h e primary c i r c u i t .

Keeping t e l l u r i u m , which can be

harmful t o Hastelloy-N when d e p o s i t e d on i t s s u r f a c e , i n s o l u t i o n i n t h e

s a l t may be p o s s i b l e by a p p r o p r i a t e c o n t r o l of t h e r e d u c t i o n / o x i d a t i o n p o t e n t i a l of t h e s a l t .

Most of t h e fission p r o d u c t s would remain i n solu-

t i o n i n the fuel salt.

It a p p e a r s ( b u t must be demonstrated) that a f u l l

30-year i n v e n t o r y of t h e s e materials could be t o l e r a t e d w i t h o u t exceeding solubility l i m i t s . Because r o u t i n e a d d i t i o n s of uranium would be r e q u i r e d to m a i n t a i n c r l t i c a l i t y i n t h e r e a c t o r , a d d i t i o n s of l i t h i u m and b e r y l l i u m would a l s o be r e q u i r e d t o m a i n t a i n t h e d e s i r e d chemical composition.

Some of t h e s e

a d d i t i o n s , c o n c e i v a b l y , could be used t o h e l p c o n t r o l t h e o x i d a t i o n s t a t e

of t h e s a l t , which would have t o be a d j u s t e d r ~ u t i ~ l e tl oy compensate f o r

the o x i d i z i n g e f f e c t of t h e f i s s i o n process.

Also, t h e t o t a l salt inven-

t o r y p o s s i b l y would have t o be l i m i t e d through o ~ c a s i o n a lwithdrawals sf

some salt. The DMSR, i n common with o t h e r systems t h a t would rase molten f l u s -

r i d e salts, would r e q u i r e a s p e c i a l primary s t r ~ c t u r a lalloy and, pos-

sib~y,

ana - ~ ' e f i e c t o ~ me . a l l o y that

graphite for the

w a s o r i g i n a l l y d e ~ e l ~ p ef d o r molten-salt

s e r v i c e , Hastelloy-N, was found

t o be e x c e s s i v e l y embrittled by n e u t r o n i r r a d i a t i o n and t o e x p e r i e n c e s h a l l o w i n t e r g r a n u l a r a t t a c k by f i s s i o n - p r o d u c t tellrarim.

Subsequently,

minor composition m o d i f i c a t i o n s were made which a p p e a r t o p r o v i d e adeq u a t e r e s i s t a n c e t o b o t h r a d i a t i o n embrittlemewt and tellurium a t t a c k . While e x t e n s i v e t e s t t n g and development would s t i l l be r e q u i r e d t o f u l l y

q u a l i f y t h e modified Hastelloy-N as a r e a c t o r s t r u c t u r a l material, t h e f u n d a m e n t a l technical issue of a n adequate m a t e r i a l a p p e a r s t o be re-

solved o

The r e q u i r e m e n t s imposed on t h e g r a p h i t e i n a DMSW a r e much less severe t h a n t h o s e t h a t would a p p l y t o a high-performance b r e e d e r r e a c t o r . The low f l u x levels i n t h e core would l e a d t o damage fluences of l e s s t h a n 3

lo25 neutrons/m2

iFl

30 y e a r s ,

c o u l d l a s t f o r the l i f e of t h e p l a n t .

Some CUPPent tC!ChnO%Ogy glPaphiteS

I n a d d i t i o n , t h e l o w power d e n s i t y

may e l i m i n a t e t h e need t o seal t h e g r a p h i t e s u r f a c e s t o limit

t r u s i o n and poisoning.

X ~ ~ Q % i nI-

This would s u b s t a n t i a l l y r e d u c e the technology

development e f f o r t a % s O e h t @ dWith the tlmaFiufactuPe Of BXSB graphite. The g e n e r i c s a f e t y f e a t u r e s of a DMSR would d i f f e r s i g n i f i c a n t l y

from t h o s e of o t h e r r e a c t o r t y p e s p r i m a r i l y because of t h e f l u i d n a t u r e

of t h e f u e l ana t h e c i r c u l a t i n g inventory of f i s s i o n p r o d u c t s .

kcarase

the f u e l i n a DMSR would be u n c l a d , t h e t h r e e l e v e l s of f i s s i o n - p r o d u c t confinement f o r t h i s system would be t h e RCPB and two s e p a r a t e l e v e l s of containment.

The primary containment would be a s e t of sealed and in-

ertea equipment cells that would be i n a c c e s s i b l e t o p e r s o n n e l a f t e r the o n s e t of p l a n t o p e r a t i o n ,

These cells would provide t h e p r i n c i p a l con-

finement of r a d i o a c t i v i t y i n a c c i d e n t s i n v o l v i n g f a i l u r e of t h e RCPB.

They could a l s o p r o v i d e a u x i l i a r y c o o l i n g of s p i l l e d f u e l s a l t i f t h a t

s a l t f a i l e a t o flow t o t h e cooled d r a i n tank.

TBSS

of cooling a c c i d e n t s

_... ..&/ ;....; .

.. ....

w i t h r e a c t o r scram may b e r e l a t i v e l y mild i n DNSRs because of t h e l a r g e h e a t c a p a c i t y and low vapor p r e s s u r e of t h e f u e l salt which i n h e r e n t l y r e t a i n s most of t h e f i s s i o n - p r o d u c t decay-heat g e n e r a t o r s .

However, l o s s

of c o o l i n g because of blocked c o r e f u e l passages a t f u l l power could Lead t o some l o c a l s a l t b o i l i n g .

A f u l l s a f e t y a n a l y s i s of t h e DMSR has n o t

been p e r f ~ ~ ~because ~ e d i t would r e q u i r e a much more comprehensive d e s i g n than i s c u r r e n t l y a v a i l a b l e . P r e l i m i n a r y c o n s i d e r a t i o n of t h e environmental e f f e c t s of DMSRs sugg e s t s t h a t such e f f e c t s would g e n e r a l l y be m i l d e r t h a n %or c u r r e n t l y opera t i n g n u c l e a r systems.

There would be l i t t l e o r

r o u t i n e gaseous and

l i q u i d r a d i o a c t i v e e f f l u e n t s , less waste h e a t r e j e c t i o n , no shipment of r a d i o a e t i v e s p e n t f u e l d u r i n g t h e normal p l a n t l i f e , r e l a t i v e l y l i t t l e s o l i d r a d i o a c t i v e waste, and l e s s impact from uranium mining.

t r a s t t o t h e s e more f a v o r a b l e f e a t u r e s , t h e DMSR a t end-of-life

In conwou%d

i n v o l v e a more complex decommissioning program and a l a r g e r s o l i d waste d i s p o s a l task.

I n a d d i t i o n , d u r i n g o p e r a t i o n , t h e r e t e n t i o n of tritium

and the r e l a t i v e l y l a r g e r i n v e n t o r y of r a d i o n u e l i d e s may r e q u i r e e x t r a e f f o r t s t o avoid p o s s i b l y u n f a v o r a b l e e f f e c t s .

In g e n e r a l , t h e a n t i p r o l i f e r a t i o n f e a t u r e s of t h e once-through DMSR appear t o be r e l a t i v e l y f a v o r a b l e .

The e n t i r e f i s s i l e u r a n i m f n v e n t o r y

would be f u l l y d e n a t u r e d , and t h e r e would be no c o n v e n i e n t means of i s o l a t i n g 233Pa f o r decay t o

2 3 3 ~ . me f i s s i l e plutonium inven-

t o r y would be s m a l l , of POOP q u a l i t y , and d i f f i c u l t t o extract from t h e l a r g e mass of h i g h l y r a d i o a c t i v e f u e l s a l t .

I n a d d i t i o n , no shipments

of s p e n t f u e l from t h e p l a n t would occur except a t t h e end-of-life.

6.2

A l t e r n a t e BMSR Concepts

Although a DMSR o p e r a t i n g on a 30-year,

once-through f u e l eycfe ap-

p e a r s t o have a number of a t t r a c t i v e f e a t u r e s , t h e b a s i c concept c o u l d b e adapted t o a number of a l t e r n a t i v e f u e l c y c l e s .

If f u l l - s c a l e ,

on-

l i n e p r o c e s s i n g of t h e f u e l s a l t t o remove f i s s i o n p r o d u c t s were adopted, some l i k e l i h o o d e x i s t s t h a t break-even b r e e d i n g performance c o u l d be achieved.

Mowever, even w l t h o u t break-even breedimg, t h e f u e l c h a r g e

could be r e c y c l e d through s e v e r a l g e n e r a t i o n s sf reactor^ t o g r e a t l y

reduce the a v e r a g e demand f o r mined uranium.

Other performance improve-

ments ( s h o r t of break-even b r e e d i n g ) c o u l d be achieved by combining t h e o n - l i n e f u e l p r o c e s s i n g w i t h p e r i o d i c removal o r reenrichment of p a r t of

I n a l l t h e s e o p t i o n s , t h e n e t consumption

t h e a c t i v e uranium i n v e n t o r y .

of n a t u r a l uranium would become a minor f a c t o r i n t h e a p p l i c a t i o n of BMSHPs.

Would

Some c o n s i d e r a t i o n w a s g i v e n to f u e l p r o c e s s i n g concepts t h a t reBIOVe

Only p a r t of t h e

SOlklbPe

f i s s i o n products.

such prQC@SseS

appear to o f f e r f e w ( i f any) advantages over e i t h e r t h e unprocessed o r t h e fuhky

plrOCeSsed

apprQaCheS,

6,3

Commercialization C o n s i d e r a t i o n s

S i n c e t h e MSW concept w a s under study and deVelQpmWIt f o r n e a r l y a0

y e a r s , most of t h e r e l e v a n t areas of the r e q u i r e d technology have r e c e i v e d a t least some a t t e n t i o n .

After t h e successful o p e r a t i o n of the EISRE, a

Pimjitecf amount of d e s i g n e f f o r t w a s expended on a commercial-size MSBB;

t h a t e f f o r t was d i s c o n t i n u e d i n 1993,

The technology development work

proceeded i n p a r a l l e l w i t h t h e d e s i g n s t u d i e s up t o t h a t t i m e .

A small

development e f f o r t ( w i t h o u t d e s i g n s u p p o r t ) w a s resumed in I 9 7 4 and canc e l l e d a g a i n i n l976,

This

WQI-~,

d e s p i t e i t s l i m i t e d scope, provided a n

e n g i n e e r i n g - s c a l e d e m o n s t r a t i o n of t r i t i u m management i n t h e secondary

s a l t and s i g n i f i c a n t p r o g r e s s toward the d e f i n i t i o n of a n a c c e p t a b l e s t r u c t u r a l a l l o y f o r molten-salt

service.

Work w a s under way toward dem-

o n s t r a t i o n of some of t h e chemical p r o c e s s i n g o p e r a t i o n s when t h e program

w a s ended A s i d e from t h e t e c h n i c a l p r o g r e s s , t h e l a s t development a c t i v i t y produced a comprehensive p l a n

~ S S the

f u r t h e r development of MSRs, which

s e r v e d as t h e b a s i s f o r the proposed DMSR development plan and schedule. T h i s p l a n s u g g e s t s t h a t t h e c o m m e r c i a l i z a t i o n of DMSRs c o u l d proceed v i a three reactor projects:

( a > a moderate-sized (100- t o 2BO-PIWe)

molten-

salt t e s t r e a c t o r t h a t could be a u t h o r f z e d i n 1985 and become o p e r a t i o n a l i n 1995, ( 2 ) an i n t e r m e d i a t e - s i z e d commercial p r o t o t y p e p l a n t a u t h o r i z e d i n 1995 and o p e r a t i n g i n 2005, and ( 3 ) a f i r s t standard-design DMSR to o p e r a t e i n 2011.

A p r e l i m i n a r y estimate for t h e c o s t o f this programrp

137

i n c l u d l n g $900 m i l l i o n f o r t h e c o n c u r r e n t base development work, i s

$3950 m i l l i o n ( i n 1978 d o l l a r s ) . A p r e l i m i n a r y estimate of t h e c o n s t r u c t i o n c o s t f o r a " s t a n d a r d " DMSR

( n e g l e c t i n g c o n t i n g e n c i e s , e s c a l a t f o n , and i n t e r e s t d u r i n g c o n s t r u c t i o n ) y i e l d e d about $650/kWe i n 1998 d o l l a r s .

This compares w i t h about $QOO/kWe

f o r a PWR aRd $38Q/kWe f o r a coal p l a n t ( w i t h o u t f l u e - g a s c l e a n i n g ) e s t i -

mated on t h e same b a s i s .

The BMSR c a p i t a l estimate d i d n o t i n c l u d e t h e

c o s t of o n - s i t e s a l t treatment f a c i l i t i e s o r t h e c o s t s of s a l t and f u e l i n v e n t o r i e s ; t h e s e q u a n t i t i e s a r e all included i n t h e f u e l c y c l e c o s t s . The e s t i m a t e d n o n f u e l B&M c o s t s were 2.82 mills/kWk, and fuel c y d e c o s t s Were

5.3 m i i l s / k m .

m e c o s t of decommissioning a DMSR was

be about POX h i g h e r t h a n t h a t f o r a comparably s i z e d LWRe

Tke liee~singof MSRs h a s n o t been s e r i o u s l y a d d r e s s e d because no

p r o p o s a l t o b u i l d a r e a c t o r beyond t h e MSRE w a s e v e r supported and none of t h e c o n c e p t u a l d e s i g n s t u d i e s proceeded t o t h a t l e v e l ,

Wowever, a

number of new l i c e n s i n g i s s u e s c l e a r l y would have t o be addressed.

Be-

c a u s e t h e t h r e e l e v e % s of f i s s i o n - p r o d u c t confinement i n a DYSR would d i f f e r from t h o s e in a s o i i d - f u e i e d

S Y S ~ ~ T ,demonstrating ,

compliance w i t h

t h e r i s k o b j e c t i v e s r a t h e r t h a n s p e c f f i c hardware d e s i g n s i n e s t a b l i s h e d l i c e n s i n g c r i t e r i a presumab%y would b e ~ e c e s s a r y . P r e l i m i n a r y s t u d i e s s u g g e s t t h a t t h e r i s k s a s s o c i a t e d w i t h t h e operatiom of MSRs may b e lower t h a n t h o s e f o r LWRs, w h i l e risks d u r i n g maintenance and i n s p e c t i o n of t h e r e a c t o r system may be h i g h e r .

'Fke p r e l i m i n a r y s t u d i e s of DMSRs d e s c r i b e d p r e v i o u s l y i n d i c a t e t h a t t h e s e r e a c t o r s could have a t t r a c t i v e performance and r e s o u r c e u t i l i z a t i o n f e a t u r e s while p r o v i d i n g s u b s t a n t i a l r e s i s t a n c e t o t h e f u r t h e r p r o l i f e r a -

tion of n u c l e a r e x p l o s i v e s .

I n a d d i t i o n , t h e environmental and s a f e t y

f e a t u r e s of DMSRs g e n e r a l l y a p p e a r t o b e a t l e a s t a s f a v o r a b l e a s t h o s e of o t h e r n u c l e a r power systems

are a t t r a c t i v e .

mile

and t h e system economic c h a r a c t e r i s t i c s

a s u b s t a n t i a l RDbB e f f o r t would be r e q u i r e d to

commercialize DMSRs, t h e r e are no major unresolved i s s u e s i n t h e needed technology.

Thus, a commercial DMSR w i t h o u t o n - l i n e f u e l p r ~ c e s s i ~ g

138 probably could be developed i n about 30 y e a r s ; w i t h a d d i t i o n a l s u p p o r t

%or IID&D, t h e technology f o r on-Bine f u e l p r o c e s s i n g could be developed on about t h e same t i m e schedule. Although t h e DMSR c h a r a c t e r i z a t i o n s p r e s e n t e d i n t h i s r e p o r t a r e approximate, t h e y p r o v i d e as much d e t a i l as i s j u s t i f i e d by t h e v e r y prel i m i n a r y s t a t u s of t h e system c o n c e p t u a l d e s i g n .

-Any e f f o r t t o substan-

t i a l l y improve t h e q u a l i t y and d e t a i l of t h e c h a r a c t e r i z a t i o n s would have to be accompanied by a s i g n i f i c a n t system d e s i g n e f f o r t o r i e n t e d toward

a s p e c i f i c DMSW power p l a n t .

C o s t s and times r e q u i r e d f o r such s t u d i e s

would be s e v e r a l t i m e s a s l a r g e as t h o s e f o r t h e p r e l i m i n a r y work and probably c o u l d be j u s t i f i e d o n l y i f a n a t i o n a l d e c i s i o n were made t o ree s t a b l i s h a f e d e r a l l y funded MSR program.

MSR program of s u b s t a n t i a l size presumably would i n c l u d e a n RD&D

Any

e f f o r t of some s i z e t o s u p p o r t e f f e c t i v e p u r s u i t of program g o a l s .

This

work, i n t u r n , would be complemented by t h e d e s i g n studies which would h e l p t o d e f i n e RD&D t a s k s and f o c u s t h e e n t i r e e f f o r t .

The combination

would a l l o w t h e a t t a i n m e n t of o b j e c t i v e s on t h e s h o r t e s t p r a c t i c a l t i m e schedule

From t h e p r e l i m i n a r y s t u d i e s r e p o r t e d , a once-through DMSR without o n - l i n e f u e l p r o c e s s i n g a p p a r e n t l y would be the most r e a s o n a b l e choice f o r development i f an RD6D program were e s t a b l i s h e d ,

Nowever, p a r a l l e l

development of t h e technology f o r ~ o n t i n ~ o uf us e l p r o c e s s i n g would add o n l y m o d e r a t e l y t o t h e t o t a l progrm c o s t and could p r o v i d e t h e o p t i o n of a more r e s o u r c e - e f f i c i e n t

(and p o s s i b l y a c h e a p e r ) f u e l cycle.

...-7......:

139

The authors want t o acknowledge t h e extensive and able s u p p o r t of

the Union Carbide C o r p o r a t i o n Computer S c i e n c e s Division, e s p e c i a l l y Greene, 0.

J e

R e

Knight, Ne

B i g g s , and E. M e Petrie.

M e

W e

H ~ P ~ B P P I PW ,e

M e

Westfall,

W e

Ford 111%

The nuclear c a l c u l a t i o n s depended h e a v i l y

on t h e i r a d v i c e and help. The estimate of c a p i t a l , o p e r a t i n g , and d e c o m i s s i a n i n g c o s t s f o r t h e DMSR were prepared and r e p o r t e d by PI. L. Myers.

REFERENCES 1.

Le E. MeNeese and M. W. Bosenthal, "MSBR: A Review of the S t a t u s flews 1 % ( 1 2 ) , 52-58 (September 19741. and F u t ~ r e , 'flucZ, ~

Re W e R o b e r t s , Energy Research and Development A d m i n i s t r a t i o n , l e t ter t o H. P Q S ~ I M , Oak Ridge N a t i o n a l Laboratory, March 1, 1976.

FI. F. Barnan e t ale, Molten-Salt Reactor Conceg-bs h t h Redwed Potential fop? f i o ~ i f e r a t i o nof Special NucZear MateP-.Zals, I n s t i t u t e f o r Energy Analysis, OWAU/IEA(M) 77-13 ( F e b r u a r y 1977).

Gerald R. F o r d , "Statement by the P r e s i d e n t on Nuclear Policy,'" Bff i c e of t h e White House Press S e c r e t a r y , October 28, 1976.

Department of Energy, Nonprolifepat&m AZterrea-biue Systems Assemf i o g r m plan, A s s i s t a n t S e c r e t a r y f o r Energy Technology Nuclear Energy P k ~ g r a ~ O ~ f~f si c, e of Fuel Cycle E v a l u a t i o n ( A p r i l 1978).

Jimmy Carter, "Statement on Nuclear Power F d i c y 9 ' * press release d a t e d A p r i l 7, 1977, B e p t , s t a t e BUZZ. 76, 429-33 (May 2, 1977).

E x e c u t i v e O f f i c e of t h e P r e s i d e n t , ate National. P ~ S P ~ L Plan, J April 29, 1977,

10.

G . T. Mays, A. N e S m i t h , and J. 8. E n g e l , &istribthtis~zand Behavior of T ~ i t i t c r ni n -the Coolant-Salt Technology Facilityy,BRNL/TM-5759

(April 1977). 11.

12.

Greene, Union Carbide C o r p o r a t i o n , p e r s o n a l esmmunicatisn t o

W. A. Woades, Qak Ridge N a t i o n a l L a b o r a t o r y , September 1977. 13.

J. Re M i g h t Union Carbide C o r p o r a t i o n , p e r s ~ n a lc ~ m m u n i ~ a t i ot n o W e A. Rkoades, Oak Ridge N a t i o n a l L a b o r a t o r y , June 1978.

14. E

141 ., .q... ....,.,.,.,.,. i.rc.. ,

16.

J. R. Knight, Union Carbide C o r p o r a t i o n , p e r s o n a l communication t o W. A. Rhoades, Oak Ridge N a t i o n a l Laboratory, June 11978.

$7.

Argonne National L a b o r a t o r y , Reactor Phpics C s m t a n t s , AMJ-580O ( J u l y 1963).

19.

W. R. Grimes, "Molten-Salt Reactor Chemistry," Nuel. A p p l . Tech~101. %,

137 (February 1970).

20.

W. R. G r i m e s e t a l . , "Fuel and Coolant Chemistry," chap. 5 , p. 95 i n ?%e Devetogment S t a t u s of Molten-Salt Breeders, 08K-4812 (August 1972).

22.

P. N. Haubenreich and J. R. Engel, "Experience with t h e Holten-Salt Reactor Experiment," Nucz. A p p l e TechnsZ. 8, 118 (February 1970).

24.

Cantor, Density and Viscosity of Sevarat Molten Fluco~ideM i t e ORNE/TM-4308 (March 1973).

tupeg,

25.

C. F. Baes, Jr., p . 617 i n "Symposium on Reprosessing sf Nuelear Fuels," ed. by P. C h i o t t i , v o l . 15 of fluetea? MetatZurgy, USAECCOW-690801 (1969).

26.

C. F. Baes, Jr., "The Chemistry and Themodynamics of Molten-Salt Reactor F u e l s , " J . .VucZ. Mat. 51, 149 ( 1 9 7 4 ) .

27.

C. E. Bamberger, R. G. ROSS, and C. F. Baes, Jr., "'Phe Oxide Chemist r y of Plutonium i n Molten F l u o r i d e s and the Free Energy of Format i o n of PuF3 and P u P , , " J . Inorg. Nuel. &ern. 33, 776 (11971).

2%.

J. K.

29.

C. E. Bamberger e t a l . , "Absenee of a n E f f e c t of Oxide on t h e Solub i l i t y and t h e Absorption S p e c t r a of PuF3 i n Molten EiF-BeF2-ThP4 and t h e I n s t a b i l i t y o f B ~ U ~ O R ~ U(111) III O x y f l u ~ r i d e s , "J. Inorg. Nucz. Chem. %3, 3591 (1971).

38.

'*Molten Salt Reactor Concept 9 1 1 Quar-terZy Reg~p?-t fop tge~iodEnding J d y 39, 1971 A t o m i c Energy Commission, Bombay, India, NP-119145.

Dawson e t a l . , "The P r e p a r a t i o n and Some P r o p e r t i e s of Plutonium ~ l u o r i d e s , "J. â&#x201A;Ź%em. A ~ o e% . 55% (1954).

142

31 e

6. J. Barton et al., Solubility 0% Cerium Triflusside in Molten Mfxtures 0% Lithium, Beryllium, and ~%oriurn~ E u o r i d e s , " Inorg. Cic.ern. $ e 309 (1970).

36 m

C. F. Baes, J r e 9 p. 617 in "'Sywposim on W e p r ~ c e s ~ i n of g Nuelear Fuels," ed. by P. Chiotti, vol. 15 of A%&?tGaP Metaz~urgy,USAEC-CQNF 690801, 617 (%969)e

37.

C. E, Bamberger, R. 6. Ross, and C. F. Baes, J k c , M~Ztapz-Satt Rea@t o r P F Q ~ Semiannual P ~ P P Q ~ W S SReport f o r P e ~ i ~Ending d February

G. R O S S , C. E. Bamberger, and e. F. Baes, 9%., "The Oxide Cpeemist r y sf Protactinium in ~ o l t e nF l u o r i d e s , " J . h o p + JIucZ. &ern. 35, 433-49 (1973).

29,

48.

43 e

9 9 7 2 , ORWE-4782, p. 72.

.:.. _.iii..... ..

-.=*

45.

A. L. Mathews and 6 . P. Baes, Jr., "Oxide Chemistry and Themsd p a d . c s of Molten EiF-BeP2 SoPution," Insrg. &ern. T 9 373 (1968).

47.

48.

8. E.

51.

W. B. Csttrell e t a l . , Bisassembty and Postoperative E z m i m t i o n of the Aircraf% Reaetor Ezperirnmt, ORNL-1868 ( A p r i l 2 , 1958),

52.

53.

W e D. Manly et a%., "Meta%lurgicah Problems i n Molten F l u o r i d e Systems," in F?agr. Nuel. dEnergy S ~ F IV . !?'eehnol~gy,figineering a d

E. McCoy e t al., "New Developments In Matesiah f o r Molten-Salt Reactors," #Uet. A p p t . ?%dwlot. 8, 156 (February 1970). McCoy, " M a t e r i a l s for Salt-Containing Vessels and P i p i n g 9 " chap. 7, p. 195 in fie Deve~opmePztStatu8 of Molten-Salt Breeders, QRNL-4812 (August 1972).

D. Manly et a l . , Aircraft R m c t o r Experiment -MetalZurgica% Asp e t s , ORNL-2349, pp. 2-24 (December 20, 1957).

s a f e t y 2, 164 (1960). 54.

W. D, Manly e t al., p t u i d h e 1 Reae-tors, pp. 595-604, ed. by James A. Lane, H. G. MaeHkerssn, and Frank Maslan, Addison-Wesley, Readi n g , Pa. (1958).

55.

Molten-Salt Reactor p v l o g m Status Repor.f;, ORNL-CF-58-5-3, pp. 112-13 (May 1, 1958).

56.

J. N. DeVan and R, B. Evans, 111, pp. 557-79 i n Confmence on Co+ rposion of Reactor Ma-ter-Lal.8, June 4-8, 9962, Proceedings, VoZ. I1 I n t e r n a t i o n a l Atomic Energy Agency, Vienna, A u s t r i a , 1962.

57.

F. F. Blanken9kip,MoZten-SaZt Reactor Program Sern&znntcai? P r o p rem Report for Period Ending July 31, 1 9 6 4 , QRn-3708, p. 252.

58.

S. S. Kirslis and P. F. B l a n ~ e n s h i p , M o t t a n - ~ a ZReactor t &.ogrcun Sem-LannuaZ Progress Report. f o r Period katling August 31, 196Bs ORm-4344, p. 115.

144 59.

E. M. Toth arnd %. 0. G f l p a t r i e k , " E q u i l i b r i a of U P ~ R ~ U I Carbides ~I in Ksbten S o l u t i o n s of UF3 and UF4 Comtained i n G r a p h i t e a t 850째K," 6. bizorg. TJtkeZ. &em. 3 5 , 1509 (197%).

60 *

61 e 62 e

W e a. G ~ i m e ~"MaterFaEs , Problems i n Molten-Salt R e a c t o r s , " i n Matee a l s and &eZs f o r High Temperature K7ticlear hhergy Appl.ieatG.ms, ed. by Me' T. Siranad and L. R. Z r a m w l t , The MHT Press, Mass., 1964.

65.

W. W. G r i m e s et al., "Chemical Aspects of Molten Fluoride-Salt Rea c t o r F u e l s , " in F l u i d FueE R e a @ t o ~ se, d , by J. A. Lane, H. 6. XacPkerssn, and F, Flmslan, Addison-Wesley, Reading, Pa., 1958.

66. 67.

.VoZten-Salt Reae-top Program Progress Report fop Peptiod f ~ o mMarch I -to August 31, 2961, ORNk-3215, pe 117.

68.

W. R. G r i m e s , Radiation Chemis-tvy of Molten-SaZt Reactor S y s t e m , OWNLITPI-500 (March 31, 1963) e

Reactor Chemistry Division Annmt BP.ogress Report fop P e ~ i a dEnding c7anuary 31, 1962, ORNL-3262, p * 20.

14 5 ...... ...... &;.&

74.

W, R. G r i m e s , N e V. S m i t h , and 6. PI. Watson, " S o l u b i l i t y of Noble Gases in Molten F 1 u ~ r i d e ~I.; In Mixtures of NaF-ZrP4 (53-47 mole X > and N ~ F - z P P ~ - u P (560-46-4 ~ mole Z>," 6. $hysB &em. 62, 862 (1958).

75.

M e Blander et al., " S o l u b i l i t y s f Nsble Gases in Molten P I u o r i d e s ; 61. In t h e LIF-NaF-KP E u t e c t i c Mixtures," J e Phys. &em. 6 3 , E164 (19591

77.

W. B, Briggs and R. B e Korsmeyer, MoZten-Sali5 Reactor P m g m Semiannual fiogress Report f o r Period â&#x201A;Ź%ding February 28, 2 9 7 0 , CXNL4 5 4 8 , pe 53.

81.

D. L. Manning and G. Mammilntov, Moltm-Salt Raactor f i o g r m Semiaplnuaz Progress R6qmr-b f o r Period m i n g February 29, 1 9 7 6 , ORrn5132, p * 38,

83.

84.

G. J. Nessle and W. R. Grimes, "Production and Handling of Molten ~ l u o r i d e sf o r U s e as R e a c t o r Fuels," fiemieal Engineering Pksgress

R. Grimes e t a l . , "High Temperature Processing of Molten Fluoride Nuclear R e a c t o r Fuels,'* Nttct. &g. Part VII 5 5 , 27 (1959).

Symposiwn Series 56(28) (19601. 85.

86.

J. H. S h a f f e r , Preparation and Hadling of Salt Mixtures f o r the Molten-Salt Reaet& Experiment ORNE-4616 (January 1971 le R.

B. E i n d a u e r , Molten-Salt Reactor P ~ o g r a mSemiannuaZ Report f o r Ending August 31, 1971, ORNL-4728, pa 226.

P%PiOd

87.

W. B. L l n d a u e r , Moltan-Salt Reactor P ~ o g m Semiannuat Repore f o r Period Ending Feb~uary28, 9971 ORNL-4676, p. 269.

146 88.

A. D. Kelmers e t ale, Evaluation of A l t e m t e Secondary (and Te+ t i a r y ) Coolants f o r t h e Motteaz-Salt ~ r e e d e rReactor, O R W m - 5 3 2 5 ( A p r i l 1976 1

89.

C n J. Barton et a l e , Molten-Sal% Reactor &.Og?+%m 3emiannmt gPiog~%Ss Report f o Period ~ ,?%ding ~ ~ b r m 29, r y 196BS QRm-4254, p 0 166.

90.

E. L. Compere, He C. Savage, and J. M. Baker, "High I n t e n s i t y Gama I r r a d i a t i o n of Molten Sodium Flusroborate-sodim Fluoride E u t e c t i c Salt," J . R7ucte Mat. 34, 97 (l9701.

92.

D. M e Richardson and J. E%. S h a f f e r , Mo'&ten-Salt Reactor f i ~ g m Semiannual Progpess Report f o r Period Ending February 28, 2 9 7 0 , ORNL-4548, p . 136.

95.

S. Cantor and Re M. &Jaller, Motten-Salt Reactor B o g m Semknnwt Progress Report f o r Period Ending August 31, 1970, ORNL-4622, p . 79.

$6.

J. B. ~ a t e set ale, Molten-Salt Reaskor fiogram &mianntlal ~ Q Report ~ Q PPeriod fidiazg February 28, 1991 OWML-4676, p. 9 4 .

~ P Q S S

98.

J. B. B a t e s , J. P. Young, and e. E, Boyd, Molten-Salt Reactor -0g m m Sem&znnuail. *ogress Report f o r Period &ding ?WwuapSy 29, 2 9 9 2 , ORWL-4782, p . 59.

99.

J. P. Young, J. B. B a t e s , and G. E. Boyd, Molten-Salt Reacbor &og m %miannual Progress Report f o r Period h d i l z g August 31, 1 9 9 2 , ORNL-4832, p . 52.

.. ,........ -.ys

103.

6 . R e Hudson 11, CQNCEPT-5 U M P ' s Manud, ORML-5470 (January 1 9 7 9 ) .

.._. ....... .

149 ... .,..Y.&.fl..,..

appendix A

COMPARATIVE REACTOR COST ESTIMATES

... ........ . ...... rrrrr,

Table A . I .

C o s t e s t i m a t e s f o r t h e DSTZR, PIJR, a n d c o a l p l a n t s ( T h o u s a n d s of 1978.0 d o l l a r s )

Account No.

Twodigit

Threedigit

Item

DMSK~

PWRh

~ 0 . 3 1 ~ 9 "

Direct costs

Land and l a n d r i g h t s

21 21 1 212 213 215 216 217 218 219

S t r u c t u r e c , and improvements Yard work Tieactor c o n t a i n m e n t b u i l d i n g T u r h i ne b u i l d 1ng Auxiliary building(ยง) Waste p r o c e s s b u i l d i n g Fue 1 s t o r a g e b u i I d i ng Other st rue t uresP Stack (heat rejection)

2 20

R e a c t o r p l a n t equipment N u c l e a r steam s u p p l y s y s t e m

221 22 2 223 224 225 224 227 228

Reactor equipment Plain h e a t - t r a n s f er s y s t e m Safeguards system Radwaste p r o c e s s i n g F u e l h a n d l i n g and s t o r a g e O t h e r r e a c t o r p l a n t equipment R e a c t o r i n s t r u m e n t a t i o n and c o n t r o l R e a c t o r p l a n t m i s c e l l a n e o i l s items

Account 21 s u b t o t a l

Account 2 2 s u b t o t a l

23 231 232

233 234 235 236 237

T u r b i n e p l a n t equipment Trir b i n e g e n e r a t o r (Changed t o a c c o u n t 2 6 ) C o n d e n s i n g systems Feed-heating system Other t u r b i n e p l a n t equipment T n s t r u m e n t a t i o n and c o n t r o l T u r b i n e p l a n t m i s c e l l a n e o u s items Account 23 s u b t o t a l

241 242 243 244 245 246

Electric p l a n t equipment Switchgear S t a t i o n s e r v i c e equipment. Swi t c h b o a r d s P r o t e c t i v e equipment Electrical s t r u c t u r e and w i r i n g Power a n d c o n t r o l w i r i n g Account 24 s u b t o t a l

..... ........... -,/....

2 ,000 10,000 44 000 14,000 24 000 ( i n 215)

NAd 30 000 2 ,000 ~

12 4 ,00 0 ( i n 221 and 2 2 2 ) 4 5 000 63,000 6 000 10,000 10,000 30 000 10,000 6 000 ~

2,000

10, IO3 39,017 12,820 9,297 8,841 4,928 25,952 FA

5,990 hit 10,337 hit Omit hit hit 2 ,203

110,958

Omit

69,111

Omit

3,727 9,873 11,582 10,042 3 ,405 19,822 7 ,779 5,497

Omit Omit hit Omit Chit bit hit Omit

180,000

138,838

43 000

61,943

42,299

12,000 21,000 18,060 2,000 4,000

15,257 15,315 15,496 1,336 3,590

12,022 14,519 14,924 837 3,190

100,000

112,937

87,791

6,000 14,000 1,000 2,000 12,000 19,000

5,749 9,419 70 1 1,770 10,215 15,737

4,081 3,949 72 I 1,879 9,422 10,805

54 000

43,581

30,857

Omit

T a b l e A. 1 ( c o n t i n u e d )

Account No. -0-

digit

Item

Threedigit

25 251 252 2 53 254 255

Yiscellaneons p l a n t equipment T r a n s p o r t a t i o n and l i f e e q u i p m e n t A i r , water, and steam s e r v i c e system C omm un i c a t i on e q u i pmmrn t F u r n i s h i n g s arid e q u i p m e n t Vaste water t r e a t m e n t e q u i p m e n t Account 25 s u b t o t a l

3 000 10,000

2,617 7 ,664

1,606 6 ,034

2 ,000

1,524 1,041

691 898 1,351

1,000 1,000

17,000

12,846

10,580

14,000

21,968

14,003

491 0 0 0

443,128

Omit

26,000 24 000 25,000

25,801 21,878 22,460

14,348 11,285 13,363

75,000

70,139

38,996

49,008 2 ,333 1,338

14,917

5 3 000

52,679

16,109

3 ,000 22,000 5,000

3,180 19,188 4,683

82 4 8,732 180

4 000

2 ,8 5 3

343

34 ,000

29 ,904

10,079

Total i n d i r e c t costs

162,000

152,722

65,184

T o t a l c a p i t a l c o s t s , d i r e c t and indirect

653,000

595 a 8 5 0

Omit

Main c o n d e n s e r h e a t r e j e c t i o n A c c o u n t s 20-26, costs

total direct Indirect costs

91 911

91 2 913

Cunstruct ion s e r v i c e s Tempo PaKy c o r i s t r u c t i o n f a c i l i t i e s C o n s t r u c t i a n t o o l s and equipment P a y r o l l , i n s u r a n c e , and s o c i a l security taxes

Account 9 1 s u b t o t a l

921 92 2 923

Home-of f i c e Home-office Home-office Home-office

engfneering services services q u a l i t y assurance c o n s t r u c t i o n managevent

Account 92 s u b t o t a l

93 93 1 932 93 3 934

F i e l d - o f f i c e e n g i n e e r i n g and services Field-office expenses F i e l d job supervision F i e I d q u a l i t y a s s u r a n c e / q iial i t y control P l a n t s t a r t - u p and t e s t Accounts 9 1 - 9 3

1,192

"Estimated

by M.

L. Myers.

& E s t i m a t e d from CONCEPT 'Selected

%st

accounts.

applicable.

F o r e x a m p l e , c o n t r o l room, a d m i n i s t r a t i o n b u i l d i n g , f i r e t u n n e l s , s e w a g e , h o l d i n g p o n d , d i e s e l - g e n e r a t o r b u i l d i n g , r e c e i v i n g , and g u a r d .

653

,.._ .

.. ... .. i.iiiiii 1

Date 1978,O 1979.0 1980.8 1981.0 E982 0 1983.8 1984.0 1985.0 1986 0 1987.0 1988.0 0

Cost t o d a t eh ( m i l l i o n s of d c d l a r s ) 0

5 10 23 55 143

313 458 596

637 653

aUnit 1; l600-bfWe DMSB power p l a n t at Middletom; c o s t basis is year of steam s u p p l y s y s t e m purchase (1978.0); conStlrUCtioIl peXTli% f S 1978.0; commercial o p e r a t i o n i s 1988.8. b T o t a l c o s t incurred t o date excludes i n t e r e s t and escalation c h a r g e s ,

.... ..

OKNL/TPI-%207 D i s t . Category UC-76

Internall D i s t r i b u t i o n 1. 2. 3. 4-8. 9. 10. 11. 1%. 13. 14-18. 19-23. 24. 25 c 26-30. 31. 32. 33 34 * 35 * 36. B

T. D. Anderson Seymour Baron D. E. Wartlne M. F. &urnan &I. we B e r t i n i IE. I. Bowers J, C. Cleveland T. E. Cole S. c a n t o r J. F. Bearing J. R. Engel D. E. Ferguson M. H. Fontana W, R. G r i m e s 8. H. Guymon W. 0. Hams J. F. Harvey He we Hoffman

P. R. Kasten Milton L ~ e n s o ~ n

37. 38-42. 43. 44-4 8 49. 58. 51. 52.

W. E. MacPhersom FI. E. McCoy L. E. McNeese w. A, Rhoades J. L. Rich P. S . Rohwer PI. W. Rosenthal

B. T. S a n t o r o k n l a p Scott I. Spiewak 54. 55. EE. E. T r a m m e l l 56. B. B. Trauger 5%. J. Re Weir 58. J. E. Vath 59. R. e. Wper ORNL P a t e n t Office 68. 61-62. Central Research Library 63. Document Reference S e c t i o n Laboratory Records Department 64-65 Laboratory Records (RC> 66

53 e

External Distribution

67.

68 *

Director, Nuclear Alternative Systems Assessment Division, Department of Energy, Washington, DC 20545 E. G. Belaney, N u d e a r A l t e r n a t i v e Systems Assessment D i v i s f s n , Id@p%rtmeTlfZOf E R @ r g y , WcPShL?3gt0ns D6 20545

69.

H. Kendrick, Nuellear Alternative Systems Assessment D i v i s i o n , Bepartment of Energy, Washington, BC 20545

78.

Office of A s s i s t a n t Manager f o r Energy Research and Development, Department of Energy, QBO, Oak Ridge, TN 37838 Director, Nuelear Research and Development Division, Department of Energy, ORQ, Qak Ridge, TN 37830 6 . Nugent, Burns and Roe, Inc., 185 Crossways Park D r i v e , Woodbury, NY 11797 R. A. Edelwan, Burns and Roe, Inc., 496 Kfnderkarnack Road, Orad e l l , NJ 07643 FJ. R. Harris, Rand Corporation, I700 Main S t r e e t , Santa Monica, CA 90406 S. Jaye, S . M. ยง t o l l e r COP.^., Suite 865, Colorado B u i l d i n g , Bould e r , CO 80302 W e L i p i n s k i , Arg~gnneNational L a b o r a t o r y , 9700 South Cass h e . , Argonme, I% 60439

71. 72 e 73.

74. 75.

76.

156 77.

He 6 . MacPherson, Institute for Energy Analysis, Oak Ridge Associated Unfversitfes, Oak Ridge, TN 37838

78.

I%. Omberg, Wanford Engineering Development Laboratory, P.0. Box 1970, Richland, WA 99352 E. S t r a k e r , Science Applications, 8400 Westpark Drive, McLean, VA 22601 A. M e Weinberg, Institute for Energy Analysis, Oak Ridge Asssciated Universities, Oak Ridge, TM 37830 For distribution as shorn in DOE/TEC-4500 under category UC-76, Molten-Salt Reactor Technology

79.

80. 81-181.

.. ..............

y.s2