ORNL-TM-6413.txt

 

  
   
  

 

4’04 ORNL/TM-6413

 

Molten-Salt Reactors for Efficient
Nuclear Fuel Utilization Without
Plutonium Separation

J. R. Engel

W. R. Grimes
W. A. Rhoades
J. F. Dearing

 

 

 
 

 

 

Printed in the United States of America. Available from
National Technical Information Service
U.S. Department of Commerce
5285 Port Royal Road, Springfield, Virginia 22161
Price: Printed Copy $5.25; Microfiche $3.00

 

 

 

This report was prepared as an account of work sponsored by an agency of the United
States Government. Neither the United States Government nor any agency thereof, nor
any of their employees, contractors, subcontractors, or their employees, makes any
warranty, express or implied, nor assumes any legal liability or responsibility for any
third party's use or the results of such use of any information, apparatus, product or
process disclosed in this report, nor represents that its use by such third party would
not-infringe privately owned rights.

 

 

 

 

 
 

-

-)

ORNL/TM-6413
Dist. Category UC-76

 

Contract No. W-7405-eng-26

Engineering Technology Division

MOLTEN-SALT REACTORS FOR EFFICIENT NUCLEAR FUEL
UTILIZATION WITHOUT PLUTONIUM SEPARATION

 

J. R. Engel W. A. Rhoades
W. R. Grimes J. F. Dearing

Oy » Date Published ~ August 1978

NOTICE: This document contains information of a preliminary
nature and was prepared primarily for internal use at the Oak

Ridge National Laboratory.

 

 

 

 

 

This report was m;:'c:m account of work
. i S
Prepared by the zmmmxfigixnmfimfikfififlfiafi
OAK RIDGE NATIONAL LABORATORY | ey, expem o it o s iy b
Oak Ridge, Tennessee 37830 | o it ot any et eeoon o
operated by oty vy at s e would o
) UNION CARBIDE CORPORATION
- for the

DEPARTMENT OF ENERGY

DISTRIBUTION OF THIS DOCUMENT I8 UN.

 
ur

 

 

 

 
 

 

 

 

 

 

 
- )

xn

#

iii

CONTENTS

SUMARY ¢ 0 ¢ & 0 o 2N E s 00t eSS

ABSTRACT

INTRODUCTION ......

BACKGROUND

HIGH-ENRICHMENT MSRs ...

S 0 o @ & 8 b o H s & e e s eSS e

S s e s BB BOREEEIBEEEEBE B

* ¢ ® O F O PSS

8PS e s

ORNL Reference Design MSBR ...¢c00... esssesssessnasessens

Reference Design Variations .

.....

Plutonium Transmuter for 233U Production ...eeeecesocecssesss .
DENATUREDMSR .......l‘......‘I...'...._’..

S ® & 9 S S ST S S O 0D S eSO

GeneralCharacteristics ® 0 2 0 QP00 E PP SR OO ESOeNS e e N

Reactor Characteristics .vceceesses

Core Thermal HydraulicsS ...cececececsss:

a0 0" " O e P e

Chemical Processing '.......'....I.j......'.........-I..'.

Balance

Of Plant @ % 9 % &0 00000

MSR TECHNOLOGY STATUS .cceceeecee

REFERENCES

® 9 * & 0 08 " 8 0 e e 0 0

Page

i W= =

11
13
14
14
16
31
33
39
39
40

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 

-t

ot

SUMMARY

Research and development studies of molten-salt reactors (MSRs) for

- gpecial purposes have been under way since 1947 and for possible applica-

tion as possible -commercial nuclear electric power generators since 1956.
For the latter, the previous .emphasis has been on breeding performance and

low fissile inventory ‘to help limit the demand on nonrenewable natural re-

‘sources (uranium) in an ‘expanding nuclear -economy; little or no thought

‘has been given to alternative uses of nuclear fuels such as proliferation

of nuclear explosives. As a consequence, the conceptual designs that
evolved (e.g., the ORNL reference deéign'MSBR)Vall favored enriched 233U
as fuel with an on-site chemical processing facility from which portions
of that fuel could be diverted fairly easily. With the current interest
in limiting the: proliferation potential of nuclear electric power systems,
a redirected -study of MSRs was undertaken in an effort to identify concep-
tual systems that would be :attractive in this situation. It appears that
practical proliferation-resistant MSRs could be designed and built, and
this report describes -a particularly attractive break-even breeder that
includes an on-site chemical reprocessing facility within the reactor pri-
mary containment. -

The point of departufe~for-this~9tudy (as for other recent MSR
studies) was the ORNL reference design MSBR, which in many respects,
reflects the state of MSR technology at the end of the reactor:development

" program in fiscal Year*1976;1:Thisfreactor*Was”chafacterized%by-a'moderate

breeding rati01(®1i07);'aélo# specific inventory of fissile ‘fuel [v1.5 kg/

“MW(e)], a reasonable-fuélfddubiing time (Vv20-years), and -almost no plu-

toniUm’from*thé~fué1*cy¢1é;*5ThisipérfOrmance3waé‘t6*befachiewedithrOUgh

‘the use of ‘fuel highly ‘enriched in 233y and -23%y (v72%) in-d high-power-

‘density ‘core and afieon;sité>f1sgion;pfoduét-c1eanufi,systemvwifih»a 10-day

fuel'brocessing*éjéle&?*TWdiimportant*Stéps5in-inéstoéésSing’CYCIé'fiere

(1) the isolation’of the enriched uranium ‘from, and its subsequeat return

to, the fuel salt and (2) the isolation of 2?%Pa for decay to *®®U outside

“:“the reactor neutron flux to prevent counterproductive neutron captures in

 
 

vi

the protactinium at the high flux levels* in the reactor. Both of these
steps, aldng*with the ready availability of'excess bred fuel, were perceived
“to contribute to the proliferation_sensitivitj of the reference concept.

‘A preliminary studylwas undertaken‘1ate‘invéalendar-yearA1976:to see
~1if the reference MSBR concept could be modified to significantly enhance
" its proliferation resistance. Among the modifications considered were
~* ‘elimination of the breeding gain, a reduction in power density (and spe-
cific power) so that protactinium isolation could be avoided without ex-
i cessive penalties, and ‘several conceptual variations -in the fuel processing
cycle. Reduction of ‘the fissile .uranium enrichment (i.e., denaturing)
was not considered at that time because'offperceived problems with the-
attendant plutonium production. ‘The net conclusion -of this study was -
that, while some enhanced proliferation resistance could be achieved, :

. the reference MSBR concept probably could not be made sufficiently re-.

- .gistant to allow its deployment outside areas that would be "secure"

against diversion of fissile material or proliferation.

In a minor extension of the above study it was shown that, if MSRs
were confined to "secure" areas, they could also be used to produce power
from fission of plutonium (generated by other reactors) and to convert-
thorium to 233U for subsequent denaturing and use at dispersed sites. -
Since the confinement of MSRs exclusively to "secure' sites did not ap-
pear to be desirable, no further consideration was given to concepts
without denatured uranium. /

The current study of proliferation-resistant systems is based on the
premise that MSRs would be attractive for dispersed deployment if they
could operate with denatured uranium fuel, have good resource utilization
,characteristics, and require no fuel reprocessing outsidélthe reactor - .
.primary containment envelope. A number of molten-salt concepts may meet
these requirements, but the one that currently appears most attractive,
is a system with_denatured7fuel and a net effective lifetime breedingnz
~ratio of 1.00. This implies that, once such a reactor-were-supplied with

 

o v}w»*Npt related to proliferation, but a potential technical problem, was

the fact that portions of the moderator graphite in the MSBR core would
have to be replaced every four years because of neutron radiation damage
at the projected high flux levels.

 

 
 

L

vii

a fissile fuel charge, it and succeeding generations of hardware could
operate indefinitely with no further addition of fissile material. Addi-
tions and removals of fertile material — both *3%U and ?°?Th — and other
salt constituents would, however, be required to maintain a stable chemical
composition. :

.- Break-even breeding in a denatured MSR is achieved by making several
changes in the reference design MSBR concept. First, changes were made in
the reactor core size and salt-graphite configuration to lower the core
power density and to enhance neutron resonance self-shielding in the 238y
in the fuel. These changes increased the fuel specific inventory somewhat
(to about 2.4 kg fissile uranium plus 0.16 kg fissile plutonium per electric
megawatt), but they also reduced the neutron losses to fission products
and 2%3pa and captures in 238y to help compensate for the reduced breeding
performance imposed by the presence of the 238y denaturant. In addition,
the lower neutron flux associated with these changes would extend the life
expectancy of the moderator graphite in the core to approximately that of
the reactor plant, thereby obviating the need for periodic graphite re-
placement. It would also substantially ease the graphite design constraints
and allow for simpler geometric shapes. Although the neutronic calculations
indicate that this reactor could operate indefinitely with the assumed
chemical processing system, there is relatively little margin for error.
However, a substantial margin could be provided by allowing the addition
of small amounts of 235U (well within the denaturing limit) with the fertile
238U, and some additional margin probab1y7c0u1d be obtained by adjusting
the nominal core design and/or the fuel processing cycle.

Aside from the core nuclear cohcept, the other substantial change

from the reference design MSBR ié in the area of chemical processing.

The requirement for break-even breeding would imposera need for continuous
chemical processing, but the cycle time apparently could be increased to
20 days (fromrlo days'for the MSBR). However, a more significant change
would_be the elimination_of'the steps. to isolate 233p, in order to avoid
the loss to waste of’plutonifim; Since plutonium, the transplutonium
actinides, and fission product zirconium all follow the protactinium, this
change not only would preserve the plufonium reduired for neutronic sur-

vival, but also avoid chemical isolation and accessibility of proliferation-

 
 

viii

attractive materials. (An additional step would then have to be:provided
in the process to:rémove zirconium on some reasonable time schedule.) The
change'actua11y<would eliminate part of the reference flowsheet since -
‘the extracted protactinium and its companion nuclides would be returned:
directly to the fuel salt. With the exception of the zirconium-removal
step, the modified process would involve the same chemical unit operations
"fpr0poséd’for'théfreférence MSBR system. Thus, this process should be no
more difficultftO'develop'and”implement-thafi that for the reference concept.
| ‘Preliminary study suggeste that no changes to the reference design
MSBR other than those’described‘abdve'for-the core and chemical plant
~ would beé required to transform the MSBR into' an attractive proliferation-
resistant concept. It appears that afcommercial-prototypé of such a '
 system could 'be developed and in operation in" about 30 years 1if a de- .-
-velopment effort were established.
MOLTEN-SALT REACTORS FOR EFFICIENT NUCLEAR FUEL -
UTILIZATION WITHOUT PLUTONIUM SEPARATION

J. R. Engel W. A. Rhoades
W. R. Grimes J. F. Dearing

ABSTRACT

Molten-salt reactors (MSRs), because of the fluid nature
of the fuel, appear to provide an attractive approach to ef~-
_ficient fuel utilization in the Th- 33U cycle as well as a
means for limiting the availability of plutonium and the
general proliferation risks associated with nuclear power
generation. :

High—enrichment 233U _systems could in principle, be oper-
ated with positive breeding gains to effectively eliminate
plutonium as a nuclear fuel. However, such systems would be
proliferation sensitive. Concept modifications (short of de-
‘naturing the uranium fuel) can be conceived to ‘enhance the

- proliferation resistance of high-enrichment MSRs, but it is
doubtful that sufficient enhancement could be achieved to make
the systems suitable for deployment other than at "secure" sites.

Denaturing the uranium in an MSR introduces some plutonium
into the fuel cycle and generally degrades its breeding perfor-
mance. Nevertheless, a denatured MSR with full-scale on-site
fuel reprocessing appears to be capable of break-even breeding.
In addition, the plutonium (most of which is consumed in situ) -
would be of poor quality and would never be isolated from all
other undesirable nuclides. " Thus, such systems would provide
for efficient utilization of uranium resources in a prolifera-
tion-resistant environment while limiting the amount of plutonium
(and transplutonium actinides) that would have: to-be handled as

‘waste. i

The deve10pment of commercial MSRs by early in the 21st
century appears to be technologically feaaible.

"”INTRbDUCTION

The ‘interest in limiting the distribution and availability of ex~
plosives—usable Special nuclear materials "(SNM) , particularly plutonium,
along with a recognized need for optimum utilization of nonrenewable
'energy sources, ‘has led to a reexamination of the Molten—Salt ‘Reactor
:(MSR) concept as a potential candidate for resource-efficient nuclear

electric power generation within these constraints. Prior studies of

 
 

this concept had established it as a neutronically feasible nuclear
breeder in the Th—233U system, but its proliferation resistance was not
considered. In the current. study, an effort is being made to retain
favorable nuclear performance of the reactor while enhancing its pro~
liferation resistance to a level that may make it attractive for wide-

spread deployment as a nuclear power system.

The criteria for judging the proliferation resistance of a given
nuclear power concept have not- been fully established but some of the
properties of the "ideal“ nuclear system are readily apparent. First,
such a system should avoid the isolation of plutonium (of whatever iso-
topic composition) as a pure material anywhere in the reactor cycle, in-
cluding the fuel cycle.; Second the system should 1imit to the extent
possible the inventory of SNM at explosives-usable isotopic compositions,
regardless of its chemical impurity or unavailability. Finally, the
system should provide reasonable safeguards for any: SNM that might be
transformed (e. ges by isotope separation) into material that could be used
for explosives.‘ Another factor that has not been heavily emphasized is
that, since the current generation of. light—water reactors iS‘producing
a substantial amount of plutonium, there may be some advantage in a system
that could in an appropriately safeguarded manner consume that plutonium

to obviate the need for its 1ong—term, safeguarded storage.

A variety of molten-salt reactors may be described which would have
most of these properties in varying degrees. The basic reference design
MSBR, ! developed at Oak Ridge National Laboratory, could for all practi—
cal purposes eliminate plutonium as a nuclear fuel. However,lsuch a

system would require highly enriched uranium, a comparably attractive

nuclear explosives material, as a fuel. If appropriately safeguarded fa-
cilities could be provided, MSRs could be used to transform plutonium to
2??U (which can be‘denatured)_while_efficiently using the plutonium fis-

(sionfienergy.: Such systems could range from ?33

U fuel factories, which
would require continuing plutonium fueling, to MSBRs or denatured MSRs in
which plutonium might be used only as a startup fuel. . But possibly the

most attractive proliferation-resistant MSR concept is a denatured ?°°U
-j

»)

£

system with a very limited internal plutonium inventory. Current studies
indicate that such a system could produce all 1ts own fuel requirements

and have otherwise favorable technological features.
.. BACKGROUND

The study and development of MSRs was begun at ORNL in 1947 as part

of the U.S. Aircraft Nuclear Propulsion Program. This effort led to the
~construction and operation”of_a-Z,SfMW(t) MSR.[the Aircraft Reactor Experi-

ment (ARE)] in 1954. Although the effort to develop an aircraft propulsion
unit was subsequently abandoned, the potential oi.MSRs,for{civilian power
production was recognized and a development program directed toward that

goal was established in 1956. ' This effort led to the design, construction,

and operation of the 8-MW(t) Molten-Salt Reactor Experiment (MSRE). Cri-

tical operation of the MSRE spanned the period from June 1965 to December
1969, during which the reactor accumulated over 13,000 equivalent full-

power hours of operation and demonstratedfremarkably high levels of opera

»bility,_availability,4and.maintainability,zn_Ihe reactor was fueled

initially with a mixture of 235y and %3%y which was subsequently removed

- (on site, by fluorination of the salt mixture) and replaced by 233U thus

making it the first reactor to .operate at significant thermal power with

this fuel. During the latter stages of reactor operation, a few hundred
.grams of plutonium was added to the reactor to demonstrate its compati-

,bility with the salt mixture.,,,,_.uz

Subsequent to the operation of the MSRE, some conceptual design work

vas _aconutziseed. toward a Molten-Salt Test Reactor and a commercial-size

Molten-Salt Breeder Reactor (MSBR) However, most of the program effort

- was directed toward further development of MSR technology., Emphasis in
ifithe design study was on moderately high breeding performance and a minimal
_specific fissile inventory for the: system. These objectives led.to a

r‘and a compound doubling time of %19 years.gji 3

.. The apparently favorable characteristics of the MSBR attracted some
industrlal and utility interest; this led to the formation of the Molten-

Salt Group, headed by Ebasco Services, Inc., and including several prominent

 
 

F“U S. corporations. This group carried out some design studies and as-
sessments of the ORNL work (under subcontract) as well as some indepen—
dently funded studies. - | T -
All AEC-supported work on the MSR concept was interrupted in early
1973; the program was terminated and all subcontracts were canceled. The
technology development effort was resumed in early 1974 (no conceptual
design work) ‘and ‘terminated again in’ mid-1976. ‘One result of that effort

was a comprehensive program plan for the development of MSRs. - The cur-

““rent study is part of the Department of Energy s Nonproliferation Alterna-

fi'tive Systems Assessment Program, ‘which was established in support of
ALPresident Carter's Nuclear Policy Statement of April 7, 1977.
Molten—salt reactors, in common with essentially all fluid fuel con-
'fcepts.'have’ainumber'of'characteristicsrfihichamay'proveivaluable’from‘the
"standpoint of nonproliferation of miclear explosives. Since the fuel'is
“a fluid, essentially all fuel fabrication ‘and refabrication steps are
eliminated from the reactor fuel cycle. Thus, at least in principle, it
‘should be possible to carry out completely remote operations within the
primary containment of the reactor system. This would'éliminafe=a11§direct
access to the fuel constituents. | . e
~Since the fluid fuel also contains fission products, thé entire pri-
‘mary circuit (including the fuel processing facility) is highly radiocactive
and ‘therefore not easily modified for diversion of fissile materials.’ Any
such modification would require remote procedures which, even with exten-
sive preparation and preplanning, would be difficult, time consuming, and
expensive. Clandestine modification of the facility would be'essentially
impossible because of the high radiation 1evels inside the" primary con—
‘tainment.‘“ ' | ' o S | A
‘Molten-salt reactor systems as a class, particularly those treated
- here. ‘have many features in common. All are thermal reactors’ with ‘unclad
graphite as the neutron moderator and ‘all use the same nominal salt mix-~
‘tures and the same conCeptual'balance—of-plant'design. DifferenceS'ambng
concepts are primarily in the details of the fuel-salt compositiOn“(e ges
‘uranium concentration and isotopic composition) and in the on-1line fuel—

":cleanup concept.
 

-j

-

”

fblanket region.

HIGH-ENRICHMENT MSRs

The principal advantages of high-enrichment MSRs are their favorable
nuclear performance in thermal spectra and their near-complete avoidance
of plutonium; their principal disadvantage is the need for "secure" siting
due to the proliferation attractiveness of the highly enriched uranium
fuel. In the equilibrium fuel cycle, with no 238U in the initial loading,

the fuel contains a.small amount of 238py . and almost no higher actinides.

f-flfORNL Reference Design MSBRE

Prior concepts of high-enrichment MSRs are typified by the ORNL
reference design MSBR ‘shown schematically in Fig. 1 and described in
some detail in Ref ‘1. This design (breeding ratio =1, 07) Tesulted from
an effort to restrict the reactor fissile 1nventory [1 5 kg/MW(e)] in
order to maximize the conservation of uranium in an expanding, but ulti-
mately limited nuclear economy Somewhat higher breeding ratios could
.have been obtained at the expense of higher inventories and correspondingly

longer fuel doubling times.

Reactor system

The primary feature in the MSBR design is a high-power—density, well-

thermalized, graphite-moderated reactor in which a single molten salt con-

taining both fissile and fertile material serves as both the fuel and

blanket fluid. The two major neutronic functions (energy production and

breeding) are achieved with a low fuel inventory by varying the fluid

~fraction from about 13 vol A in the core region to about 37 vol #Z in the

The fluid fuel consists essentially of a molten mixture of "LiF and
BeF2 containing appropriate quantities of ThFu and UFQ in a homogeneous
solution. The molten fuel is pumped from: the core to heat exchangers where
heat generated by fission (and other related nuclear processes) is trans-
ferred to a molten secondary (or‘coolant)rsalt, a eutectic mixture of NaBF,

and NaF.* The secondary salt transports the heat to the steam supply

 

*This mixture has frequently been called "sodium fluoroborate."

 

 
ORNL-DWG 68-1185-ER

  
  
   
 

SECONDARY
SALT PUMP

PRIMARY B
SALT PUMP ' o NCBF4-N0F
COOLANT SALT

 

]

 

B
—

 

 

 

 

 

 
   

 

 

 

 

PURIFIED 704°C md
SALT

GRAPHITE
MODERATOR

REACTOR

HEAT
EXCHANGER

 

 

 

 

 

 

 
  

 

566°C ~ 4m

 

  

CHEMICAL
PROCESSING
PLANT

—
TLiF -BeF, - ThFy - UF, |
FUEL SALT -—

 

 

 

| . STEAM GENERATOR

 

 

 

 

 

 

 

 

 

 

 

—/

 

 

 

 

 

 

 

 

 

TURBO-
GENERATOR

 

- STEAM

Fig. 1.;uSing1e-f1uid, two-region molten-salt breeder reactor.

 
 

system and serves to isolate that system from the primary fluid, which
is thereby confined to the reactor primary containment .system. The
secondary salt also serves to intercept tritium migrating through the
heat exchange system toward the steam circuit.

The high degrees of radiological, chemical and thermal stability
of the inorganic fluoride salts and. their low vapor pressures permit the
operation of MSRs at relatively high temperatures (the nominal reactor
outlet salt temperature is about 975 K) and’ correspondingly high~tempera-
ture, high—efficiency (nominally 447%) , steam—electric power cycles. In
fact, the high melting temperatures of the salts (e.g., the liquidus
temperature of the fuel salt 1is 775 K) require that these reactors be
operated near the higher_portion of the usual temperature range for fission
power systems. This high—temperature operation requires the use of high-
temperature design and systems technologies and also allows the use of

established high-temperature steam~power technology.

Fuel reprocessing

The fuel processing piant;'or fisgion-product-cleanup system (Fig.
2), of the reference'design.HSBRiis conceived to operate continuously on
a small side stream of molten”fuelrs’sa This processing plant removes
fission product"poisons fdr discard as waste. In addition, it removes
233pa from the fuel mixture and accumulates it within the processing plant
where it can decay to high—purity 233U without further exposure to neutrons.
(Minimizing protactinium 1osses through neutron capture is particularly
important at the high power density of the reference design MSBR and much
less important: in designs ‘that Operate at lower power densities )

All the fission product species do not go to the processing plant;
~ krypton and xenon are removed by sparging with helium in the reactor. The
- seminoble and noble metals rapidly deposit on surfaces within the reactor
vessel and the primary heat exchanger' of these elements, only niobium ap-~
pears to plate preferentially on the surface of the ‘graphite moderator.
Tritium diffuses through.the heat—exchanger tube walls into the NaBF4-NaF
coolant, where most of it is retained.

Most of the separations are accomplishedrby selective extractions of

cationic species from the molten fluoride fuel into bismuth containing

 

 
ORNL-DWG 78-572
. "ADD Li: :
i 83 g-molesdey

"

   
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 
    
      
    

 

 

 

 

     

 

    
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

e e e e e e e e

r.—‘-.* Do -3
' Too1 e
r__.mr TO RECYCLE 1 * T | 3.2...‘.”!’:..;....__.:
. | - g ) :
. RE*2 - }
MAKE UP ure ‘ ) _ 7 ? STRIPPER r_!_il_l%tl’l);z:c_'-“.___q |
74 8 g-mole DeFz AND REDUCTION| : . 3 . o = . 9
1003 g-mole TaFq/day . - -~ l Y : 7 i © RE*? |
" J | beeod 8 L _omaworr |
: s . : ue 0.02€cmissec | .|
- _ FUEL SOLVENT DISCARD + | 42 emd/ssec , '
. REMOVALL™ O Haem3/ssec r_———!l e . T ; o ' b
ey i TRANSFER] : : oo i, 3
e, B . ' : wer et Ao ge P oot [ 5 v-motesaer '
ExTrRaction| - _ ! ] T - . o - o —ea 2 !
: e i e e | g .- RE3* i en. B L
: 'L Bi-10-RES*-mpRe : 4 — ! ACCUMULATION | 10.28 cmdmec
ytdyte ———r—f1'!—‘.”.—" ———— o —————— — —————— - - ———— 1 . ‘ . N3 ‘) 1 q
VALENCE. | ‘ ‘ ‘ : 1 1. T } o
ADJUS TMENT - T ,r - T . : { ] . * ' : 1 :
P l I . ) ' . * o ) . 1 .
Yo RE,ACTOH: . 233py -t SECONDARY| u'.‘::‘m':CT ' : . e = e o e - ) : - - : :
3 emimec XTRACTION N -l L 1wse Urdey | } | M | pmrmmem e MR
. . . . ' i STRIPPER] ) ‘  TORECYCLE} |
L . i . TI™ I Ve 3% o e REde 2 t
: - : L @i-ts Th-y-Pe-2r sToma - b 2080 T i '
- LWL Ta-y-p GE AND DECAY : . T . | DRAWOFF Y !
(_EmoM AgAcTOR, MAETRA | o e LIF ~ThFa-ZrFa=Pafs 1oV s e | 023 embnec JrLuon B
T 9% cmdisec e . V- ' in _2¢ 28 2 ‘ 1 ADD L,  Y0Sem¥ec WF, Hy" . .
Fe—> ATION | » - } e’ 371 @-moles Aoy ; . S : 1
. : 1 pr—— PATCH DISCARD FINAL { § :
: - ouR - -
. R WaTion | pe stonage sauy [7F =220 dde o, ! , RIS
LEGEND ‘ : e ' ascmdrsec < f2 | : FOR ULTIMATE '
———— FUEL OR FUEL SOLVENT 1 DISPOSAL .
——— - BISMUTH - P 1 | WASTE SMT :
—e-e- UFg ) e e —————— e e e e e B e i o e e e e e e it ——————— -
wemvaen LICE ) : S : L_ : g - "_'
cpt— WASTE SALT ) . - e e e — ——— ——— e o o ——— . kY S = o o i S o 2 2 T o o — e e e —————

.

Fig.2 Flowsheet for fuel proéessing plant in 'reférence. desi’gn MSBR.

 
-}

properly adjusted concentrations of litbium. Beryllium is not extracted;
Zr, U, Pu, Pa, the rare earths, and Th are extractable in that order.®?®
Such reductive extraction processes from fluoride fuel can effectively
separate uranium from protactinium (but not from zirconium) and protactin-
ium from the rare earths"and-thorium. Rare—earth fission products are par-
tially extracted from molten fluoride mixtures by bismuth containing
moderate concentrations of lithium, but they areraccompanied by an appre-
ciable quantity of thorium. Separation of thorium from rare earths (and
from Y, Ba, Sr, Cs, and Rb, which behave similarly) must be accomplished
by transferring all these elements (except thorium) to molten LiCl from
the bismuth-1ithium alloy 5,10,11

Uranium can be separated and recovered by reductive extraction, but
fluorinatiOn.to UFs isymbre effective and convenient; The UFg and F; are
absorbed in a sufficientzquantity of purified fuel solvent containing
UFr.s’6 Uranium in thisrsolution is reduced to UF, with H;, and the re-
constituted fuel salt is returned to the reactor after final cleanup and
adjustment of the average uranium valence’to about 3.99; Bry, I» (and
probably SeF¢ and TeFg), which are volatilized with the UF¢, pass through
the sorber and must be removed from the off-gas stream.

A small processing plant is sufficient, The reactor fuel passes
through the plant every ten days with a processingfrate of 55 em®/s (0.87
gpm). Table 1 summariaes the removal methods and cycle times anticipated
for such a plant.® The several separations required are well demonstrated

in small—scale experiments, but engineering—scale demonstrations are still

,largely lacking, and materials to contain ‘both molten fluorides and bismuth

;alloys seem certain to pose some problems.

fNonproliferation attributes

Once placed in operation, the reference design MSBR would require no

sthipments of fiesile material to the reactor and only occasional shipments

" of bred excess 233U to fuel other reactors. Accordingly, 1t would present

a very low, ‘and perhaps acceptable, profile toward diversion by subnational
or terrorist groups. However, as far as weapons proliferation — a national
decision to exploit the machine to produce nuclear weapons — such a reactor

has pronounced and obvious weaknesses. The uranium within the fuel is

 

 
Iable

1. Methods and cycle times for removal of fission products and
" salt constituents in an MSBR processing plant

 

 

 

Adapted from Ref. 5.

-

Group . Component thgzal Primary removal operation .

Noble gases Kr, Xe 50 sec Sparging with inert gas in reactor fuel
T _ circuit : _
Seminoble and noble Zn, Ga, Ge, As, Se, Nb, 2.4 hr Plating out on surfaces in reactor vessel_

metals Mo, Tec, Ru, Rh, Pd, Ag, and heat exchangers - ‘
t cd, In, Sn, Sb, Te :
Uranium 233y, 234y, ZQSU; 236y, Volatilization in primary fluorinator'
’ 237y - returned to carrier salt and recycled
‘ to reactor
Halogens Br, I. 10 days Volatilization in primary fluorinator
' - followed by accumulation in KOH solution
_ in gas recycle system sy L
Zirconium and 2r, 23%pa 10 days ,Reductive extraction into Bi-Li alloy f
protactinium followed by hydrofluorination into
o : ‘ , Pa decay salt ‘
Corrosion products Ni, Fe, Cr 10 days- Reductive extraction into Bi—Li alloym
S ' ' - followed by hydrofluorination,tnto Pa
, _ decay salt o - : _
',Trivalent rare earthsb Y, La, Ce, Pr, Nd, Pm, 25 dayse _ Reductive extraction into Bi-Li alloy‘
' ‘ Gd, .Tb, Dy, Ho, Er followed by metal transfer via LiCl
, 7 : into Bi-5 at. % Li solution .
' Divalent rare earths Sm, Eu, Sr, Ba 25 daya® Reductive extraction into Bi-Li alloy
and alkaline earths ’ followed by metal transfer via LiCl
- ) into Bi—5 at. % Li solution o
. Alkali metals Rb, Cs 10 days Reductive extraction into Bi-Li alloy
L o o followed by accumulation in LiCl
Carrier saltfu f Li, Be, Th 15 years Salt discarda ‘ ‘

Y is not a rare earth but behaves as the trivalent rare earths. .

Effective removal time —-varies for the different elements.

ot -

 
11

clearly usable material for weapons, and its removal in relatively pure
form by fluorination could be accomplished with little difficultyffiy use

of the available processing system. Of course, snch-an ac;ion would be

an overt andlqbvioue_treaty violation (the reactor could no longer furnish
power), but given suitable other preparations the fiwarning time" could

be quite short. Less obvious (and probably more‘insidious) routes for
_prpliferafion are,'in,principie, avaiiable. The reference MSBR produces
more *3°U than it requires; this 233U_is_genereted;in‘quite;pure form in
__.the'prptac;iniumraccumnlatipn system and is aveilable_via flnorination with
the installed processing gear. ‘Attemprs,tp remove it secretly should be
obvious“foen inspector, but succesefniirenovai;would beiundetectable for
a:moderately}long_periodir:Itfiis probably eaey_to underestimate the dif-
ficulties in such scenarios, ,The preeence_panppreciable;quanitities of
?E?U and of more than traces of fission preincts4will e&d to the difficul-
.ties, but a well-planned. end determined effortrcould obviously surmount

| them. As a consequence,_theureference'MSBR nould ‘seem more Vvulnerable

than most reactor types to rapid results from an overt proliferation action

and would offer significant opportunities for covert action.
Reference Design Variations

Because of the perceived proliferation sensitivity of the reference
design MSBR, a briefastudy%g.waS‘undertaken’infithe fall of 1976 to deter-
mine whether the basic concept could be modified to make it sufficiently
- proliferation resistantafor;widefdeploymentras a power: producer. The re-

vquirementwfor,a:positiveebreeding1gain?was,eliminated,;but;the high-

: enrichment:fuel;cdmposition;was-retained:to completely avoid the need to
deal with plutonium, Theflonlyuother_changefconsidered:in the reactor was
a -lower power density (higher;fiSsile;epecific-inventory);to,reduce the
-zzsignifieanCe_bfaneutron?absorptions:in;%3?Paf(if;Pa;isqletion_were;aban—

'é;doned);and;;afieliminategthe=needkfpr?periodic:replacement{of;moderator

n,&graphite*in'the~reaetor*corer;=Five variants of the basic system, including

the fission—product-cleanup concept, were considered.x.-f»r
-:The first-wvariation. modified the ‘reactor performance capability and
eliminated the breeding of excess fissile material. Such a system would

have all the proliferation resistance (or sensitivity) of the reference

 
12

concept but would lack the potential for" continuous removal of fissile
fuel while maintaining reactor operation. J o

' The second variation eliminated all fluorination steps'—-the most
7proliferation-senSitive’procedure in the entire fuelécleanup“process.‘“
This WOnld'preVent the isolation of 2%%Pa and would require more isotopi-
caiiy'separated37Li; since uranium removal prior to fission-product clean-
:up'would:belaccompiished by rednctionflwitfi‘lithium.‘ It”appeared’that‘fuel

'self-sufficiency could be maintained in such a system with a reduced
' reactor power density (to limit Pa losses and reduce the relative poison-
ing effect of other fission products) and a significantly longet fuel
;'processing cycle. The longer processing cycle would also reduce the re-
'quirement for "Li. The elimination ‘of the fluorination steps was felt to
"represent a significant increase in proliferation resistance. -

The third variation involved a major change from the nominal fission—
product-cleanup concept; it was proposed to substitute a CeF3 ion exchange
system for all the chemical fission—product-cleanup operations. (Gas R
'stripping to remove xenon and other volatile fission products would be
retained.) Such a system would remove only the rare éartfié*{by'sabst1;
tuting Ce, which has a lower neutron cross. section) and leave a variety
of other fission products in the salt. Some degradation in breeding per-
formance would be experienced, but it appeared that self-sustaining opera-
tion could be achieved at the lower: core power density. Since this
process completely avoided separation of the fissile material, it appeared
to be significantly more proliferation resistant :than the-reference“concept.
However, the ‘technical feasibility of this approach has not been demon--
strated, and substantial research, development, and demonstration -(RD&D)
would .be required to reduce it to practice'if.it is feasible.. .

The use of some form‘of'vacuum:distillation for fuel cleanup was
proposed as a possible fourth approach to enhance the proliferation’resis—
tance of the reference reactor concept. Although such an approach would
eliminate many of the proliferation-sensitive stepe, it was not clear 'that
it would be workable with a salt containing ‘thorium. The technological
-uncertainty of this approach tended to rate it relatively low among the
- possible alternatives. ' ' ' ’
 

fiyiinventory of product. Moreover, ‘the MSR permits recovery of ‘the

13

.The final alternative considered was the elimination of all on-site
cleanup processes other than physical removal of noble gases. The poten-
tial feasibility of this approach was based on some earlier studies of
high-performance converter MSRs in which the unprocessed fuel charge was
- simply. replaced every few years. It appeared that, if reactivity varia-
~tions-could be managed, such a system might require replacement of: the
fuel charge only two (or possibly three) times during the life of a
reactor plant.  Although such changes would require the application of
additional safeguard measures, the infrequency of the changes was judged
to make this approach reasonably acceptable.

Although some of the proceSSing*modifications to the high-enrichment
concept appeared to be clearly technically feasible and all provided some
enhancement of the proliferation resistance of the reactor, it did not
appear that the antiproliferation gains were of sufficient magnitude to
‘justify an extensive effort to develop the reactor and the associated
fuel-cleanup system. Consequently, nondenatured MSRs for power generation

at dispersed sites were not considered further.
Plutonium Transmuter for >3y Production

At may,be,that‘any high-enrichment MSR would have to be located at

a site where special safeguards would be in effect and thus special-purpose
MSRs might also be acceptable. Of particular interest in this regard would
be MSR systems that consume plutonium and higher actinides (produced by
other reactors) and producegga3Ugforgdenaturing and subsequent utilization

at dispersed sites. o |
e Thermal or near-thermal reactors (Whlch include MSRs) are inherently
h"filess efficient” burners of plutonium than are fast reactors and are at ‘some

‘idisadvantage in'"fuel—factory applications. However, MSRs ‘have minimal
para31tic absorbers in their cores need neither head-end reprocessing
steps nor fuel element refabrication, and have a much smaller in—process
233U B
product .as8’ soon as it is produced' hence, very little of ‘the" product-—
whose greatest value is as an export commodity -is consumed by fission

fbetween replacements of solid fuel elements as in ‘the fast reactor system.

 

 
 

14

“Thus, ‘any advantages MSRs might have as' "fuel-factories” would be.related
“to-their fluid fueli® o oo e
© "The net production capability would be a major,Vbut~notrthe only,
“eriterion for7evaluating‘"fuel factory" options. ' Other ‘significant cri-
“teria would include the technological feasibilitj ofrthe’concept,?indus—
tria15acééptability,VcommercializatiOn*potential,?safety and reliability,
licensability, time ‘to commercialization, and the probable net cost of
the produCtQ*‘Molten—salt reactors have not:been seriously considered

Vheretofore”as?safeguarded-producersflof\233U;>perhaps’they should be.
'DENATURED MSR .

. MSR systems containing substantialmamountstof‘?agU have:not been
considered,in_most prior studies because of the perceived difficulties
ingdealingawith the plutonium that would be produced.f_lnnaddition, such
. 1 gystems would not be compatible with the high breeding performance and
low inventories that have been among the traditional system goals.. How-
ever, with the current emphasis on proliferation resistance and ultimate
‘resource utilization in fission energy systems MSRs fueled with denatured

uranium may have significant overall technical advantages. The denatured
MSR" (DMSR) described in the following subsections is based on-& preliminary
JCOnceptual“study of this';ystem.' It is anticipated that a more precise
and detailed description will be evolved as the study is continued.’

General Characteristics

- The principal characteristics desired in a DMSR are (1) that it meet
"to the maximum extent practicable the currently perceived requirements for
-‘resistance to proliferation of nuclear explosives and (2) that it provide
for a very ‘high level of resource utilization._ ) . .

| At equilibrium,+ the principal fissile material in the denatured
system is uranium with 2330 and 235U in a ratio of about 10 1. Sufficient

 

Their efficiency. as net energy producers would be an. advantage in
comparison to accelerator—driven fuel producers.

‘ +Isotopic equilibrium for fissile uranium is effectively reached in
-3 few years and is independent of whether startup was on 235y oy 233y,

 
15

238y 4g present in the mixture to dilute the 233U by 6:1 and the 2%°U by
4:1. - Additional denaturing is provided by the 23U and 23U in the steady-
state mixture to achieve the-preferred'dilutions for nonproliferation.

Although substantial plutonium is produced from the 238y,

the high neutron
cross sections of the plutonium isotopes and the fact that all plutonium
-.1s retained in the fuel salt keep the total plutonium inventory relatively
low; about- 10% of the fissile material is plutonium (239 and 241 isotopes).
The long effective exposure time of the piutonium results in the buildup
of substantial amounts of 2*%Pu and 2%2pu. Although these isotopes have
significant fission cross sections (particularly at high neutron energies),
- they also undergo spontaneous figsion (i.e., produce neutrons), which tends
to detract from their value as explosives materials. In addition, there
is no provision for the isolation of plutonium from a number of other
radioactive and otherwise undesirable nuclides. One other potentially
attractive material is ?aaPa,-which:is present to about 84 kg in the fuel
salt at steady state. If this material could be isolated from the rest
of the fuel, it would eventually produce high-purity 2°3U, which would be
proliferation'sensitive; "HoueVer,xprotactinium'isolation is not part of
the concebtuaIISystem and'modification-of the‘system to provide for such
”’”isolation would be difficult, time consuming, expenéive, and readily de-
tectable. - "

' Utilization of all natural resources in the denatured system appears
“to be”quite:favorable;”'Significant"amounts'of“7LiF‘(and'hence beryllium

~and’ thorium fludrides)* must be continuously removed from the fuel salt

" as' 'Ll 1s’ added in the fission—product-cleanup system; however, these

materiale could be’ recovered by a variety of aqueous processes if it were
-economically attractive to do s0. ‘The effective breeding ratio can be
 maintained at 1.0, so that, after the initial fissile loading; no fissile
“‘materfial need be added or removed for the life of the plant, howeVer, thor-
“"jum and 23%U must be-added’ continuously to maintain ‘the concentrations of
thése nuclides. At the end of plant life,'only a small’ amount of addi-

" tional’ uranium would have ‘to be added to that recoverable from the old

 

' These materials must all be included as potentially limited natural
- resources.

 

 
16

“plant (to substitute for plutonium that is not recovered) to start up
‘a new plant. Alternatively, the entire salt charge from a retired plant
(including in-salt fission products,'plutonium, and higher actinides)
icould be transferred to a new plant with no new fissile addition and no
plutonium left over for storage or disposal.,

The basic reactor flowsheet for the DMSR is essentially the same as
that for the reference design MSBR. The only differences are in the core
:.configuration, details of the fuel-salt composition, and the fission-
'-product-cleanup*(chemiCal.proceSSing)~system, Thus the primary-system
 temperatures, pressures, and major flow rates, as well aS=the;entire

~secondary system and balance of plant, would be the same as for the refer-
-ence plant. The remainder of this section is devoted to those portions of

the denatured concept that have not been described previously.

Reactor Characteristics

The principalicriterion for an attractive DMSR is survival in a
neutronic sense. It is axiomatic that adding 238y to a thermal spectrum
V‘MSR degrades its overall breeding performance because the plutonium that
is produced has a lower effective fission neutron yield than 233U in such
a system, In addition, it was recognized that protactinium isolation
would not be acceptable and that neutron and hred 233y josses due to
neutron captures in 233pa would have to be accommodated. Thus, the nuclear
design problem became one of halancing a low core power density (to limit
protactinium losses and graphite heating) against a higher fissile inven-
‘tory in a‘oore of reasonable size and balancing a more heterogeneous _

(lumped) core‘(to limit neutron absorptions in ZSBU) against potential

N cooling problems in large moderator elements,

 

| One of the first requirements established for the DMSR was the need
_for break—even breeding. This requirement probably applies to all denatured
fluid-fuel reactors and to any other systems in which the entire fuel

charge has one homogeneous composition.* The actual "critical point" for

 

o In such systems it is not possible to upgrade the average core en-
 richment by removing below-average (depleted) fuel and adding near-average
(but still denatured) material.

 

 
 

17

operating feasibility occurs when the fractional rate of production of
fissile isotopes equals that of consumption of fissile isotopes with ap-

t propriate consideration of the rate of burnup of fertile material. At
this point it becomes possible to sustain reactor operation indefinitely
by additions of denatured fuel. For denatured feeds containing 137 233y
in 238U and ZOArzasU in 238U the minimum acceptable MSR conversion ratios
are 0 98 and 0. 97 respectively. However,‘such systems would be signifi-
*:cantly less attractive than a true break-even reactor in that they would .
require transport of substantial denatured fissile material to the site.

A DMSR must have an effective fission-product-removal system and must
use the plutonium produced from 23°U efficiently to achieve break—even
fbreeding over its lifetime. The plutonium and protactinium, as well as
uranium, must be removed from the fuel before rare-earth and other fission

products ‘can be removed. Accumulation of 233Pa for deCay outside the
| reactor (as was planned for the reference MSBR) could not be permitted for
the DMSR since it would make high—quality 233U available with moderate
ease. It is convenient to remove plutonium and protactinium together from
the fuel and to reintroduce them immediately to purified fuel solvent for
return to the reactor. Such retention of 233Pa in the reactor tends to
" lower the tolerable neutron flux (and the power density) to limit losses
of 233Pa by neutron capture.' This decreased power density increases the
fissile specifie inventory for the system but also has some favorable
effects. |

1. If the neutron flux must be reduced, it is reasonable to reduce -
it to values that limit irradiation damage in the core graphite such that
the graphite lifetime is equal to that of the reactor, thus eliminating
‘the need for scheduled moderator replacement. JV .

) A the lower neutron flux, the xenon poison fraction for a given

‘jxenon concentration s reduced thereby possibly eliminating the need to

iI;imPregnate the graphite surfeces to reduce their permeability t° gases.

3, The attendant lower graphite power densities lead to lower tem-
="perature rises in the graphite, thereby substantially easing the design
constraints on moderator elements.

4, The poison fraction associated with the shorter-lived fission

products is somewhat reduced, providing slightly more margin for operation.

 
 

18

Core configuration

 

| Consideration of the above factors led to the selection of a nominal .
reference reactor concept with the characteristics described below.:__

y 1. A cylindrical reactor about 10 m in diameter by 10 m high in- -
'cluding the reflector. The core size is determined primarily by the

“n)_neutron damage flux to the graphite With little influence (at these 1arge

B sizes) from criticality or conversion ratio._ Hence, effective flux flat-
tening in the core might allow selection of a smaller reactor size or a ;
| longer graphite life with minimal reactivity penalty. | -

2. A nominal fuel fraction in the core zone of about 134 subject
to optimization and minor spatial variations (axial and radial) for flux
flattening. -

-3, Absence of a high—fuel—fraction "blanket"'zone, comparable to
‘the 37A salt zone surrounding the core in the reference ~design MSBR. This
'zone was used to help limit neutron leakage in the original breeder con-
cept.

4, Simple cylindrical design (25 cm OD) for the graphite moderator
elements with relatively large—diameter (v5-cm) central fuel passages.
Refinement of the design might lead to modification of these properties.

This basic reactor design appears to meet the neutronic and thermal-
hydraulic requirements of the system while providing latitude in several
areas (core size, fuel fraction, and moderator-element size and shape) for
adjustment of the system performance to cover uncertainties. |

In addition to the above features, the reactor would include salt
" inlet and outlet plenums (between the core and reflector) at the bottom
and top of the core that would be characterized by high fuel—salt volume
'fractions. A similar, though smaller, salt zone would be present between
 the core and reflector in the radial direction to accommodate the differ-
ential thermal expansion between the metal reactor vessel and the graphite
moderator. (The reflector is attached to the vessel 80 that it moves
outward as the vessel expands on heatup ) The effects of these zones are

included in the conceptual design.

 
 

 

19

Neutronic properties

 

Nuclear composition and the basic fuel cycle. The reference graphite
and fuel characteristics and compositions are shown in Tables 2 and 3.
The isoéopic:composit{on &f the actinide component of the fuel at equilib-
rium depends on the refueling policy, the removal process, and the flux-
averaged cross sections. The fuel circulation is rapid, so that fuel
everywhere in the core can be assumed to have one composition.

After startup, the basic refueling policy is to add thorium con-
tinually in the amount required to hold the concentration constant and
to add 23%U as required to satisfy the'"denaturing‘inequality;"'Nzgau-3
6N233U + 4N235U » where N refers to nuclear number density. The actual
amounts fed at equilibrium, assuming a Q.75 capacity factor for a 1000-
MW(e) plant, are 601 kg of thorium and 116 kg of uranium per year. Thorium

Table 2. Reférence characteriétics'of fuel
salt and moderator for a denatured MSR

 

 

Characteristic : : Value
Graphite moderator density, Mg/m? - 1.84
Fuel-salt density, Mg/m3 3.33
Salt volume in reactor vessel, m® -~ 80
Salt volume outside reactor vessel,‘ma ‘23

Core salt-volume fraction -~ - ~ - . 0,129

 

Table 3: Nominal chemical
~+ .composition of:fuel salt:

 

 

. Material = ' .  Molar percentage
BeFp ... 160
Fission products o Traece T

 

aX'refefs éd all aétinides.

 
 

represents 84% of the total feed on either a molar or a weight basis, and

- elither depleted or natural uranium could be used with only insignificant
differences. (Pure 238U was assumed in these studies.)‘: | ,
oA fission—product-cleanup Pprocess much .like that. described for the
reference design MSBR.(see,also_Iable_l)risfpresumed to operate contin-.
uou81Y=t9_remove,mate?ials‘f??@-thesfuelpéaip- A 20-day processing cycle
was assumed for the denaturedvsystem_(vs%lo.days;for*thehreference'MSBR),
so that effective removal times from-Iable'l;are approninatelyfdoubled for
those elements*_whose_remova1:is a;function_of theiprooessingfcyclegV_Other
differences‘from‘the,reference;cyclerthatfarise}from,changes in the nominal

e +
reprocessing concept are: .

1. The ?%%Pa remains with-the fuel:salt. indefinitely rather than
being isolated on the nominal 20-day processing cycle. - !

2. The transplutonium actinides are recycled into the fuel salt.

3. Selenium and tellurium are’ removed with the halogens on the
nominal 20-day processing cycle rather than plating out on metal surfaces
on a very short cycle. L

4. Fission-product zirconium, because it requires a special separa-
tion operation, is removed on a rather long (%300-day) time cycle.

5. The fuel carrier—salt replacement cycle 18 about 7.5 years.

The breeding and burning of fissile fuel proceed approximately as
shown in the nuclide charte (Figs. 3 to 5), which illustrate the Th-U
U-Pu, and transplutonium chains in the EM?R,_respectivelx.__Although the
actual branch fractions depend'on“tfie“flUX'level“as well as the energy
distribution of flux, these simplified chains indicate the potential for
a mixed-fuel breeder. The data shown on the figures indicate a total of
- 2.36 neutrons absorbed and 2.51 neutrons produced for each thorium atom
consumed in the 2%2Th chain, while the 2°%U chain has a "cost" of 3.20
neutrons and a yield of 2.88. From this, we .can see that a combined

neutron yield gives a small'surplus‘to”accounthfor’nbnactinide losses.

 

Halogens, corrosion products, trivalent and divalent rare earths,
alkaline earths, and alkali metals.

TThe modified reprocessing concept is described in more detail in a
later section. : -

 
 

 

 

 

ORNL-DWG 78—10031
238p,,
| g 2.12d
g 237N nyY o 238
g~ |6.75d
nl 7 l~-t[ 914 nl 'Y 235 nr 7 236 nv 7 2
> > » —_—237
10.6% ' V=% Y U
S . o R
27.0d " H eV
98.3% - v =2.42
- 7 16.66h
C233p, MY o34
- “Pa 17% Pa
g~ |22.2m.
sy,

 
 

99.7% ;.

Fig. 3.

Simplified thorium—~uranium nuclide schematic.

1¢

 
 

22

" ORNL-DWG 78-10033

 

= x_. ———=——= FROM #Z¢m =
i . _
28py e 00 29y, MY 20p,
AN 37.7% N
62.3%
n,
238Np u = 288 :
f~ 12354
239~p
p
g~ | 23.5m

 

 

v=275

Fig., 4. Simplified uranium?fiifitonium nuclide schematic.

Since the feed material is 84% thorium,;the net neutron yieldeis

2.88

Y = 0.84 =% 4 (1 —-0 84) 3—53-= 1. 04 .

2 3

The branch fractionsafld*the-Th/chhein ratio are both sensitive to
the neutron energy spectrum; ae discfiSsed later. The above equation shows
that the effect of the 238U chain is an important loss of reactivity and
that efficient use of the resultant smaller neutron yield 1is required.

The overall effect of the higher transplutonium actinides is of par-~
ticular interest. .The DMSR .is unusual in that these nuclides are recycled
indefinitely as an alternative to includiqg them with the waste stream.

This reduces the long~term waste problem, but it can have a significant

 
e "““"‘3%'9?"" TN U
S S N f.n,Kaa,%'

3
e
t
|
Q
s
|
N
O
3
» 5

""'msd 65%

Cm_'l.'_'.‘.. 2“5cm——>2“sc

ORNL-DWG 78-10934

vz 33

nl f\ss% L

DY oare, Y e DY 2e0n

 

L . g 1oan

 

 

5

cere e g SATE VY
‘.

 

 

|
/

ade Sy P ey
N %% . .

Dok
v=294 °

 

v=45

-\
SO
S

 

 

& 311d

52%

- 64m

 

. 490

16%

Zfléakm_zsoak
48%

g~ 3.22h

LY _asp Ny . 2851 262
> Cf , wa Ct

a] 2.65y

 

TO 2480"1("__- —.—. —_ —J

Fig 3. Simplified transplutonium nuclide schematic.

 

 

£c

4 g e T AT
 

24

effect on the neutron yield of the system. Data taken from a 200-year
operation study show that each atom of 240py produced from 23%py is joined
by 0.11 atom produced by o decay of 2““Cm. 1If the additional 2I+°Pu, 241py,
and 2"2pu reactions are. ‘taken as-a- part of the transplutonium effect, we.
can characterize: the total effect as follows: For eachlabsorption in
Zh2py calculated,without the transplutonium chafn, 4.0'additional absorp-
tions, 1.0 additional fissions, and 3523additiodal fission neutrons are
born. The net result is a loss of O.Sgneutron fier "nornalfabsorption in
262p, : Lo [ ] -

The fiasions in 2"50m ?“lpd? and 2“7Cm, in descendiné order, are the
largest neutron contributors associated with the higher actinides. At the
low power density of this system, the a decay of 24%cn leads to an impaired
neutron yield compared to that at higher power densities._ Also, the
8~ decay of 2*!Pu becomes .a nontrivial loss of fissile material.

Neutron absorption in %33Pa represents a significant loss of reac-
tivity in this concept, since each atom would otherwiseidecay to a fissile
233y aton yielding 2.2 neutrons directly for each‘absorption. Each ab-
sorption -in protactinium leads to another in 23"U before - ‘a. fissile material
is finally produced. Higher power density would make this situation worse.

The'nonfissioning capture in 235U is similarly unprofitable. A total
of three additional-captures are*required to produce a fissile nuclide,
239py, Some of these chains would take many years to develop fully, for
example, 238y would saturate with a time constant of approximately 30
years. Even so, the full equilibrium value would eventually be reached.

Consideration of all these factors leads to the equilibrium fissile
inventory of the reactor. The total inventory of 233y + 235U is thus
2.4 kg/MW(e), while the fissile plutonium* (229py + 2"lPu) inventory is
0.16 kg/MW(e), |

Neutronics results. The concentration, absorption,“and fission data

 

corresponding to the fully developed breeding chains in the DMSR are shown
in Table 4. More than 98% of all fissions take place in 233U 235y, 239py,
and 2*1Pu. The U/Pu fission split is 5 to 1, but the plutonium neutron

 

The total plutonium inventory is about 0.37 kg/MW(e), S0 only about
437% is highly fissile.

 
25

Table 4. Nuclide concentrations and reaction rates
in the DMSR after long full—power operation

 

 

 

' Nuelid Concentration® -~  Neutron - Fission
clide -(x1029)‘ , absorption fraction
232y 3,211.0 1 0.32775 0.00248
233p, 2.12 0.00396 0.00001
233y | 54.7 0.32284 0.75133
234y C 24,0 - 0.03420 0.00043
2355 6.07 0.03403 0.07272
236y - 10.0 0.00610 0.00008
237p ) 2.01 0.00607 0.00005
238y5. 348.0 ' 0.06769 0.00119
239p, . 2.69 . 0.06723. 0.10896
240py, 1.63 0.02538 0.00006
21»1?1‘l : ‘ 1.26 _ 0.02435 0.04687
2“2Pu‘ o | 3.43 0.00635 0.00006
Transplutonium o 0.02605 - 0.01577
Total actinides 0.9520 1.00000
Fluorine 47,800 0.008
Lithium , . 22,400 - 0.007
Beryllium _ 5,010 0.001
Total fuel salt | | 0.968
Graphite = | | 92,270 0.020
Fission products 0.004

 

- Total , . - o : 0.992

 

Nuclei per cubic meter of salt or moderator.
bAbSOrption per neutron born; leakage is 0.008.

Includes some 2“9%Pu, - 2%1py, and 2*2Pu produced from o
. decay of | l”’Cm.-_ Lk ' '

yield per fission is significantly higher.i Neutron leakage is only
small loss in this system, and captures in nonactinide nuclides are also

f_loy The neutron utilization can be summarized as follows

“:Absorbér*tjge*f> L “-*Absorgtioni(zzgfir'

Actinides v Lo 005020
Nonactinide salt nuclides 1.6
Fission products \ 0.4
Graphite =~ o 2.0
Leakage 0.8

 
 

26

The depression of thermal flux in the fuel is of some interest because
it governs the allowable size of the moderator 1ogs from a neutronics
standpoint. . If the flux depression is large, graphite and resonance cap-
ture will be enhanced._ Table 5 shows that flux depression would not be

excessive in the reference core design. _

TableiSQ Fuel disadvantage factors

 

 

“Neutron energy . Inner fuel  Moderator Outer fuel
- group o zone - zone
pl-(fast) 118 . 0.98 1.11 .
2 (resonance). = ~ 1.00 . - 1.00 : 1.00

 

The spatial peaking factors for both power density and fast-neutron
flux (E > 50 keV) have gignificant effects on moderator graphite lifetime
in MSRs, particularly in the low-power—density concepts ‘where a moderator
lifetime equal to that of the reactor system is desirable.’ The power-
density distribution primarily affects the graphite temperature, which in
turn affects the amount of graphite damage for a given neutron fluence; the
neutron flux directly affects the carbon-atom displacement rate as well
as the temperature. The peak-to-average values for:both power density
and neutron flux are the same in the nominal core design, both in the
radial and axial directions; the values are 1.69 and 1.35 for radial and
axial directions, respectively. The core average neutron damage flux is
3.1 x 10?3 neutrons/cm -sec (E > 50 keV). If a fast fluence of 3 x 1022
neutrons/cm2 is agsumed as the 1imit of useful msderator life, this value |
would be reached in the highest-flux region in 13 equivalent full—power
years (17.3 years at 75% capacity factor) Less conservatism in defining
the upper limit for damage fluence and flux flattening may allow an exten-—
sion of the useful graphite life to the desired 30 years at 75% capacity
factor. ‘ '

— Startup'and control. The startup of;the‘denatured;system can be

 

accomplished with either 233U or 235U at the appropriate‘denaturing level.

 
 

27

The effect of the denaturing is such thatreithgr;fuel_will give approxi-
-mately the same perfo?fiénée. .The initial.?eéc?i?ityflis’very senéitive to
~the initial fissile loading, as shown by Fig. 6. The calculated f;ssile
~loading required to achieve initial criticality_and_qve;come_equilib:ium
fission product loading ie 2371 and 3115 kg for %0 and ?%°U, respectively.
Figure_6,also‘shows,that_a_2%rerror_in the griticality calculations could
_;be,cqmpensatedfby.a_SZ;éhange.in fissile loading.

- After startup,. an increase in .reactivity on the order of‘Z.SZ will
occur due to the greater effect of buildup of new fissile material over
.. -that of fission products. Short-term reduction of fgactivity could be
. acqqmplished}by-withholdifig uranium from_the_inpufi étream, A short-term

~ increase could;be,accomplighed‘by redficing;hefithorium.cqntent,although
the long-term effect of this action might bé legs_fissile production.
Thus, reactivity increases would more likely bg:providéd by smali fissile
additions. B | |

ORNL - DWG 78-10932

 

0.100 [

W

0075k

0050 .. &

REACTIVITY CHANGE .

 

 

 

 

' o ECEE | 20 - 30 40 50
TLwt T FRACTIONAL CHANGE IN'LOADING OR -+ 7+ o
... . MAKEUP ENRICHMENT (%) '
- Fig. 6. Effect of initial loading and enriched makeup on equilibrium
reactivity. | - |

 
28

" Long-term breeding and nuclear design flexibility. ' The meutronic
:jcalculations indicate that the DMSR would start and run for ‘the life of the
moderator on fuel which 1t manufactured’ internally. However, more is ‘ex-

5pected of it. In this scenario, it ig intended that the fuel be- recycled

‘°”Findefinitely in a succession of new reactors as the useful life of the old

;fones ends. This would eventually lead ‘to a buildup of ‘the "trash" nuclides
(236U 237Np, 238py  2%2py  and the various americium and ‘curium nuclides).
- of these, the data used for 2%%Pu and the various ‘americium and
curium nuclides must be described a8 estimates and are perhaps subject to
errors of 30/ or more. If these chains develop as predicted ‘the ‘ultimate
’effect'would be a slow'approach“after‘msny‘yearS“to an'absorption fraction.
of 0. 0633 (including plutonium ‘&nd’ transplutonium ‘effects) due to ‘absorp-
tion in 23®pu, with a yield of 0.0377 for a net loss of 0.026. Present
"Ecalculations indicate ‘that a system using “this fuel ‘would be no more than

barely critical if the calculations were accurate. S |
What can be done if these predictions are true? What if reactivity
is even lower than predicted? Potential alternatives for increasing the
overall system reactivity include (1) altering the spectrum to improve
neutron production, (2) enriching the 23°U added, (3) altering the fuel
salt processing concept, or (4) adjusting the denaturing limit to reduce
the 23%¢ additions somewhat. The poténtial for improvement by spectrum
modification seems attractive. Certainlyfithe fission neutron yield is
sensitive to_the energy spectrum. To illustrate this point, Table 6 shows
a three—énergj?group structure used in soue:of the analyses. Absorptions
in groups 1 and 2 show a net loss of neutrons, while there is a gain in
group 3. Thus the neutron yield is sensitive to the ratio of group 2
(resonance) absorption to group 3 (thermal) absorption and thus to the
fuel/moderator ratio. The relative importance of group 1 (£ast) absorp-
tions is small because both absorptions and productions are much smaller

than for the other two energy: groups. Some additional information on the

spectrum effect may be - obtained - by- intercomparing the group-average neutron

‘absorption cross sections and the effective neutron yields for some of the
heavy-metal nuclides in this three—group structure (Table 7). For example,
it is clear from a comparison of the Th/zsavecross-section ratios in the

resonance and thermal groups that the ratio of Th/%3%U neutron:ahsorptions

“th

 
 

29

- Table 6. Three-group-neutron structure and
reaction rates for the postulated DMSR

 

 

 

 

 

'Ener"; ' Flux Relative  Fission Net
-Group rangy - volzme' . neutron- neutron fission
o g_ , absorption  production source
1 14.9-1.00 MeV 21 0.008  0.005 0.69
2 - .1.000.55 eV 223 - 0.378 0.197 0.31
_3: 0.55—0.005 eV 145 0,606 | 0.798 0.00
Total 389 0.992 1.000 1.00
Table 7. Selected cross-section data for
fissile/fertile nuclides
Group 232y 233U | 235y 2aafi 239p,  24lp,  233p,
1 (fast)? - T
Ua - 0.19 2-2 ’ 105 0.51 o 2.1 2-0 0-81
nb 1.2 2.6 2.6 2.4 3.2 3.1 2.1
2 (resonance) | | | ;
Oa C 1.6 51 25 5.8 28 39 53
n 0.00 = 2.1 1.6 0.00 1.7 2.4 0.00
3 (thermal) PP - . '
Og 3.0 250 274 1.2 1400 1000 16
0.00 1.8 2.2 0.00

n 0.00 ‘2,3 2.0

 

See also Table 6, ; |
Defined here as vc /g .flf"‘*

},would be increased by reducing the resonance flux in relation to the ther-
‘mal flux.; The _same would. be true of the ratio of ?33U/Th absorptions.‘ In
this system, almost -every neutron absorption in thorium also results in a
neutron absorption in 233U, thus, ‘an . increase in the absorption effective-
nessigfm%??U with.reducedxresonanceafluxlleads‘to_a\lower allowable inj
ventorybof§%§3UAIelative;toLtfiOrium.;fiSince the required;%saU'inVentOrY

is governedfprincipally byfthe'emount-of 2331‘I_present, this also leads to

a lower 23-aU_loaLd:I.ng._ Both of these effects work to increase the relative

 

 
 

30 -

importance of the more . productive Th-233y chain (as measured by the higher
weighted—average value of n for 233U in this spectrum).

All these factors tend to make the neutron yield larger when resonance
flux is reduced by increased moderation. Acting contrary to this trend is
the tendency of ‘the resonances in 238U to capture more as the concentration
is reduced. This effect does not dominate, however. A thermal spectrum
results in more- absorption in nonactinide salt nuclides and graphite.

Also, it is necessary to increase the moderator volume fraction to make
the resonance flux lower. These effects result 1n more parasitic absorp-
tions, which tend to offset the beneficial effects of the more-thermal
spectrum,

In the reference MSBR, a "blanket" with a relatively high salt frac-
tion and a harder spectrum was used ‘around the optimum-spectrum inner core.
This“tended to increase reactivity,” with the fissile material being pro-
duced—in-a;hard spectrum.with.lowmparasitic capture and consumed‘inva
softer central spectrum. Although the resulting core would‘be more com-
_plicated (and difficult to manufacture), the alternative might be accep-
table if one were required to provide the added reactivity. o

Figure 6 shows the effect of enriching-the 238U makeup. lhehamount
of ?3°U added would remain’as before, but some 233y would be added. If
the material were enriched to the nominal denaturing limit, V1% of reac-
tivity could be gained. A 50% enrichment:vould yield 7% of reactivity.
This would require special protection of the material added but the
amount would be only 155 kg of fissile material per year. Enriched 235U
could also be used with somewhat inferior results.

Since fission products constitute only a very small reactivity loss

in this concept (cf. Table 4), the reactivity gain that could be realized

o by modifying the fission—product-cleanup process probably is insignificant.

| However, in the equilibrium fuel mixture, there is significant" poisoning

associated with neptunium, plutonium, and the transplutonium ‘actinides.

o Thus removal ‘of some of these materials,' possibly between movements of

. the salt'from one*reactOr plant to another, could effect a significant .
extension in the useful life of the fuel charge. (In the 1limit, the "
"'entire fuel charge could be consigned to storage or ‘disposal at the end

'of life of a given reactor.) It seems apparent, however, that this

 
 

31

approach would have an unfavorable effect on the antiproliferation at-
tributes of the concept. . _u‘ d

~The final option — reducing the denaturing ratio — may be inferior to
the other three from an antiproliferation viewpoint, although it would not
add to the fuel cycle cost as would enriching the feed material. Allowing
the 233y denaturing factor to drop to 4 as for ??SU would produce a O.7Z
increase in reactivity. . Further reductions in the ZS?U loading Would also
improve the reactivity but would decrease the proliferation resistance of
the system. . 4 L

In summary, it appears that an attractive, proliferation—resistant

DMSR with break-even breeding is neutronically feasible and that suffi-
~cient latitude and alternatives exist to ensure its technological success

in this area.

Core Thermal Hydraulics

The reactor core thermal-hydraulic features, particularly with respect
to graphite temperature and xenon transport to the graphite, were major
considerations in the reference design MSBR. Although the design con-
straints are considerably relaxed in this area for the DMSR they remain
'31gnificant from the standpoint of overall technological feasibility of
the concept.

Because of the relatively low power density of this reactor concept,
simple core configurations which were not possihle in the MSBR reference
design may. be considered Three simple designs were considered: (1) a
core made up of Spaced graphite slabs, (2) a core made up of stacked hex-
agonal graphite blocks- with circular coolant channels, and (3) a core con-
sisting of a hexagonal array of graphite cylinders With central coolant
channels. ", o .“,.

Constraints which must be considered in selecting a core design in-
clude maximum graphite element temperature, local ‘salt volume fraction, and
the 238U self—shielding effect, which imposes a minimum limitation on the
coolant channel dimensions. The temperature rise between the coolant
channel and the hot spot im- the graphite moderator element is especially

important because of the strong dependence of graphite dimensional change

 

 
 

32

with temperature: The salt volume fraction and the 23°U self-shielding
effect strongly couple the thermal-hydraulic and thefnéutronic'core designs.
' These combined constraints appear to rule out the possibility of a
" graphite slab core configuratiom. Mechanical problems, especially the
" loss of coolant chamnnel geometry due to shifting'of'the}stacked hexagonal
blocks, rule out the second option. The third design seems to fill all the
”Erequiremente'an& is also very appealing because of its structural simpli-
" éity — which ia:importantjin'a core expected to last the 1life of the plant.
The outer diameter of the cylindrical graphite elements would be V25 cm
and the diameter of the inner coolant chanmel %5 cm. This yields a salt
volume fraction of 13% and equal coré salt temperature increases of 140°C
in the central and outer coolant channels, Figure 7 shows the basic core
geometry and the two types of salt flow channels.(the central and the outer
channels) which are formed between the moderator elements. The 30° annular
section of moderator element used in the thermal analysis is also shown in
/Fig._7. If the heat transfer from the surfaces of this element were uni-
form and characterized by a Dittus—Boelter correlation film heat transfer
coefficient the maximum temperature rise in the moderator at the center
of the reference core would be %60 C. The heat transfer, however, is B

obviously not uniform to the outer channel because (1) the salt (which wets

" ORNL-DWG 78-10035

   

FUEL SALT

. MODEL SECTION FOR
THERMAL ANALYSIS

GRAPHITE

Figt.7. Reference corewconfiguration for denatured;MSR..

 
 

-t

-

33

graphite poorly) will not penetrate all the way to the point of contact of
the moderatorgelements, (2) the salt velocity near the point of contact
will be greatly.diminished, and (3)_regions of{low salt velocity.willrhave
temperatures greater than the channel average because >90% of the power;is
generated in the flowing salt. In addition, the Dittus-Boelter correlation
may not apply, because a thin film of helium may exist on the graphite
surfaces. N , 7 4 -

In the absence of information on salt heat transfer coefficients,
penetration depths, and turbulent velocity profiles, an estimate (probably
conservative) of the moderator temperature structure was obtained assuming
a salt film heat transfer coefficient of 0 within 15° of the point of
moderator contact and a salt film heat transfer coefficient equal to 80%
of the value obtained using the Dittus-Boelter correlation elsewhere.

‘With these boundary conditions, the heat:conduction.equation in
cylindrical finite-difference form was solved in the 30° graphite section
using the method of successive over-relaxation. ‘Constant heat generation
and thermal conductivity within the graphite were assumed. . This analysis
yielded a maximum graphite temperature 80°C above the salt temperature at
the core center and a maximum graphite temperature in the core of 740°C
at an axial location 2.1 mfldownstream,of the core midplane.

The hydraulic diameters of the central and outer channels are 5 and
2.6 cm, respectively, which means the central,channels will need to be

orificed to more nearly equalize the salt velocities and hence the salt

ntemperature rises in two channels. This could possibly be achieved by

Vmachining small channels in the graphite near the inlet and outlet ends.

The possibility of spacing the moderator elements to eliminate the .prob-

T_lems caused by low heat transfer and low salt velocity near. the contact
- points has been investigated, but at present it ‘appears this would entail
;?a salt volume fraction significantly greater than 137% to be effective.

-;Chemicaerrocessing"

'_,_ Unit processes and operations generally similar to those in the flow-

'sheet for the reference MSBR can be used to process fuel from the DMSR.

Processing for the latter reactor has not yet_been analyzed in detail,

 
 

34
but it is clear that the flowsheets must differ in some important aspects.
The fuel volume in the DMSR must be considerably larger, and, although the
cycle time can probably ‘be appreciably greater than 10 days, the processing
‘plant will be somewhat larger than that of the MSBR. 'The DMSR will contain
“a considerable quantity of plutonium which must be retained within the
reactor circult. The MSBR system in which ?2%Pa was accumulated outside
the reactor core and allowed to decay must obviously be abandoned ‘since
such a system would furnish weapons—usable 233y 4 upon treatment with Fa.
Since’ protactinium and plutonium, along with uranium, must bé removed from
" the fuel solvent before;yttrium'and the rare-earth fission products can
be removed the DMSR must contain a system which provides for removal of
plutonium and protactinium and minimizes proliferation opportunities by
immediately reintroducing them to purified fuel ‘solvent’ for return to
the reactor. Such a protactinium-plutonium reintroduction circuit has the
" considerable disadvantage cbmpered‘with the MSR plutonium accumulation °
system that it also reintroduces fission product zirconium to the puri-
fied fuel solvent. However, the protactinium-plutonium reintroduction cir-
‘cuit has the advantage — insofar as waste management is concerned — that ' .
it also reintroduces americium, curium, californium, and plutonium to the
reactor fuel and permits only very small losses of any transuranium
elements to the waste streams.*

It seems apparent that the DMSR'can_manage the noble-gas and the semi-
noble and noble-metal fission products in the manner and with the same
removal times described earlier (see Table 1) for the MSBR. Operation of
the DMSR with 5 to 10% of the uranium present as UFj, as‘seems-feasible,_
would apparently result in essentially immediate reduction of fission

=2 and their complete reten-

product selenium and tellurium to Se~2 and Te™
“tion (with little or no interaction with the Hastelloy N) by the fuel.

Any other seminoble and noble-metal fission products:that appear appreci-
ably in the fuel stream to the processing plant could befeffectively re-

moved by a simple wash with bismuth containing no reducing agent.

 

W course, it is not known whether solid LiF-BeF,-ThFy containing
fission products can be considered a suitable form for disposable waste. *
It does seem certain that very low levels of transuranium nuclides will
offer some advantages whatever the waste form. o

 
-

 

35

The DMSR processing flowsheet, shown as a simplified block diagram
in Fig. 8, would recover about 997 of the uranium by fluorination to UFg
and would reintroduce it to purified'fuel solvent as proposed for the MSBR.
The quantity of UF¢ to be produced and absorbed per unit time would be
several—fold larger than'that fordtheEMSBR. Also, if the DMSR were operated

-with 10% of the uranium as UFj3, the quantities of Ser and TeFe¢ to be

recovered by the off-gas treatment system would beimarkedly increased.
Fission product zirconium is produced in high yield, and its removal
from the fuel is highly:- desirable. Although the zirconium isotopes are

‘not important neutron sbsorbers, any contained zirconium must be reduced

with expensive L1 and reoxidized each time the fuel is processed. It
should be possible to remove zirconium (on a cycle time of about 200

days) by partial extraction —-along with a portion of the uranium, plu-
tonium, protactinium; and transuraniumgelements —ain bismuth containing

a small concentration of‘lithiumffollowedfby selective and essentially
complete reoxidation of plutonium; protaetinium, and the transuranium
elements into purified:fuel solvent in a muitistage operation.* The
pregnant solvent from this operation serves as the absorber solution

for the UFs. Sinee the zirconiunrbearing bismuth solution cannot be
completely freed from the 238p.233%y mixture by selective oxidation, the
zirconium and uranium must be transferred by hydrofluorination to a '
waste fluoride salt and the uraniumtrecovered‘as UFs by fluorination before
discard of the waste salt at a rate:corresponding to about 4 moles of
zirconium per day. A simple method for zirconium removal on a much shorter

cycle time'would be very desirable and may be possible.+

1,;7‘;

 

This reoxidation of plutonium and protactinium must be essentially
quantitative since any of these elements (and the other transuranium ele-

ments) that remain. with the zirconium are’ consigned with the zirconium to
waste, - . L o i

TZirconium is known to form a very stable intermetallic compound (Zr-

| 'Pts) with platinum.13 This ‘¢ompound should form when fuel containing 107

of the uranium as UF; 1is, exposed to platinum, and the ZrPt; can be decom-
posed to dissolved ZrFs. and solid platinum upon hydrofluorination in the

- presence of molten fluorides. It appears that. neither wuranium nor thorium

would be removed with zirconium from the fuel mixture by platinum,'® but
there is no information about protactinium, ‘plutonium, or other trans-
uranium elements.

 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

—tm- YFE
; enppomgepe LICH -
e WASTE SALT

Figl Z 8.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

denatured MSR.

 

 

 

 

 

 

 

 

 

Preliminary flowsheet for fuel reprocessing plant in a

¥

o A _ o ORNL DWG 78-576
: C P, e i e Ce AND R |, _ iy
- ‘ e—=- - == WF 70 ACCUMULATOR [~+~*3 200 L1
o T T »
- . a .
—— 1 - - REVE  Je——— g ——upy
: R ' N B 1 b .
_ , - , unuu'non ‘ 1
N vALENCE] | ' - ace ki 1
. . " UsST- W' UFfs 1 : 1 t
. . ADJ . REOUCTION AB!ORPTION , ! ——t L ! !
i MENT Hp | - ] P : _l | i
o T T T ) ADD Li | . e e e e e '
REMOVAL| Lma i }‘ ‘ | RESE . '
= ' - . i ucu C|STRIPPER L., ! |
; : W . *- . - —————y : T . > R ; t
" 1o . mmm e e Xy = T d 4 ;
R | 1 -F s 1l T | ;-
‘ =X - 2 T mmsrtn ! Lo 1
PARTITION EARTH | e 4 '
b , EXTRACTION ' i | | O Jeta¥asomeiogin s v
i _ L—-_.—'——-f_ ‘ L |- *_._-—_... + 1 : - !
: T &- : — - : T S ». o ) : %
1 Pa 1 | ¥ + e ' 1
g | XTREE EXTRACTION] A : | SR | wees 0T vorme-l 11 1
i J i EXTRACTION _xffliflon { - i SOLVENT 4 . > STRIPPER : | FLAl!r?on:‘N- L
. ) - : W e e 3 —d =T !
T e e T 1
L. FROM - 4. . ———————e - mrmem— e T ‘
REACTOR | ' L~ I
I [ pmmany | .y o iK--z | e i
|Fwommaton] F2 | Secomanv e+ pvoro— & | PFPOQNDEWQSEE AT e |
b C FLUORIN= | ety |Taron L1 ULTIMATE DISPOSAL
‘ fa,| ATOR | HFMp | | ———t . !
. - - -2 | e e -
R o F 1 N " v :
- f’ M Y. -‘—.— - - o o v e o e e San Sk Sl D SN EPe WAL L e SN SR MW S R PR RN e e ke g
LEGEND ? ; oo _ >
-‘-.—rufl.onmnsowenr —h * - R = —m————
———— SMUTH : A i ot v e 1 o e e i —— i - -_— - -

e e, ————

9€

 

 
 

37

If the partial reductive extraction of zirconium were used, the fuel
salt would then pass‘topa_multistage extractor;where»the balance of the
zirconium, uranium, plutonium, protactinium, and transuranium elements
would be recovered by extraction into a bismuth~lithium alloy at somewhat
higher lithium concentration. By use of about six countercurrent stages
with lithium in bismuth maintained at about_Z,lSAX-lO’? atom fractionm,
the protactinium losses can be kept to_completely,negligible values and
the -plutonium losses can be made very satisfactorily low.* The pregnant
bismuth.(containing U, Zxr, Pu, Pa, etc.) would be sent to the UFs5 re-
duction and final valence adjustment stages, wheré the values would be.
recovered in the fuel for return to the reactor. The fuel solvent
(LiF-BeF2~ThF, containing a very large fraction of yttrium, the rare-
earth, alkaline-earth, and alkali-metal fission products) from this
extractor passee to the rare-earth extraction column.: -

‘The process for removing yttrium and rare-earth, alkaline-earth, and
alkali-metal fission products from the fuel solvent is the same as that
proposed. and described above for the MSBR. dThe,effective removal rates
of the several fission products depend upon the. element removed and upon
the size, flow rates, and number of effective stages in the rare-earth
extraction, - transfer, and stripping systems. However, it appears that
by processing 5% of reactor inventory per day (a number that may prove
. .uneconomically-large), the rare earths and barium could be removed on a
cycle time well below 100 days. Such removal would require discarding
about 100 moles of lithium per day through hydrofluorination of the rare
earths into waste salt. Cesium could be removed with a cycle time of 100
days by discarding about 100 moles of LiCl per day. )
| ~Since: the quantities of uranium, plutonium, zirconium, and trans-
*uranium elements that must be reduced and reoxidized are much larger than
in the MSBR, ‘the use of lithium by the DMSR will be relatively large. On
‘La 20-day processing cycle,rabout 2000 moles would be required as reductant

';; each day (with most of this entering the fuel) This corresponds to- about

 

- *It appears that protactinium losses could be kept to less than 25
,g/year and plutonium losses to about 100 g/year in the combined zirconium-
removal system and the main extractor. . .

 
 

38

0.05'm® (1.8 £ft?) of purified fuel salt that must be removed each day.*
About 280 moles of ThFy, 7 moles of 23°UF,, and 430 moles of BeF, must be
added each day.' These removals and additions constitute replacement of
Jthe ‘fuel solvent (LiF-Ber-Tth) once per 7. 5 full-power years of opera—
tion: o o | L R
" Removal of radidactiVe'speciesifrbmithe several exit gas streams
could presumably be accomplished ‘in ‘the manner proposed — though not yet
developed in detail — for the feference\design'MSBR.--Krypton'and xenon
isotopes, along with small: quantities of salt, radioactive particulates,
.and€traces of radioiodines, must be removed and_:ecovered,as'wastessfrom

the reactorisbsrgingacireuit‘:'Tritium”must'be recovered from the second-

 

ary coolant. Insofar as practicable, the several streams containing HF-
.and Hz would be combined for recovery of the HF for recycle through the.
system for generation of F2. It is clear that essentially-complete
'recovery of radioiodine and radioselen1Um and tellurium from the gases
passing the UF¢ absorption system will-prove"to-be'difficult.ff*

- fAil in all it is certain”that; even if all the gystems indicated =
above prove feasible, a great deal of development isirequired=5efdre the I
fuel processing plant could be designed in detail, ~Indeed as indicated
in a subsequent section of this document,: design of the processing plant ’
will be further complicated by the paucity of materials of construction |
that are adequately stable toward both molten fluorides and molten bismuth

alloys.

 

*This salt contains essentially the proper quantity of LiF, BeF,, and
ThFy along with some rare-earth, alkaline-earth, and alkali—metal fission
products and virtually no fissionable or transuranium isotopes. It con-

- siderably exceeds the quantity needed for the hydrofluorination of waste

- materials. It may be that an appreciable fraction of this could be stored
“and used for startup of additional DMSRs. ‘Alternatively, and especially
1f solidifed' fluoride cannot be considered an adequate disposable waste, it
may prove economical to recover at least the ’Li from the salt during its
conversion to suitable waste.

+The very short cooling time for this fuel will, of course, intensify
' the iodine retention problem though the abséence of complications from. or-
ganic solvent—iodine interactions should be of some benefit. - '
.. to'the structural alloy question

 

39
Balance of Plant

' As indicated earlier, the purpose of this study is to examine the
features of a DMSR that would differ significantly from those of the
reference'design MSBR. Since the fuel salt for the denatured system would
have essentially the same thermal-hydraulic properties as the MSBR fuel
salt, there is little if any basis for considering changes to the reference
system other than those described above for the reactor itself and the
the fission-product-cleanup system. Hence, the remainder of the primary-
coolant (fuel) circuit the entire secondary circuit including the second-
ary salt the steam system, and the plant auxiliaries would be essentially
as described for the MSBR. One possible exception to this is the shutdown
cooling system and relatedrequipment, which might be simpler for the de-
natured reactor because of the lower fuel power density. Other differences
might appear if a detailed design were developed ‘but the reasons for such
. changes would involve engineering judgment, safety analysis, and/or eco-
nomic choices rather than: basic differences in the reactor concepts. As
a consequence, most of the design study work that was directed toward the

MSBR balance of plant could be applied to a denatured aystem.
' MSR TECHNOLOGY STATUS

A comprehensive review 7 of the status of molten-salt—reactor tech~
nology was published by ORNL in August 1972 This document was comple—
‘mented by an AEC review of ‘the statusls which also identified a number of
technical issues needing solutions before an MSR could be successfully
“built and 1icensed., When the technology development effort was resumed in
_ 1974, work was directed toward several of these issues, including the pri-
mary-system structural alloy,;chemical processing technology, and tritium
'management.; Significant Progress. was made in these areas with 1aboratory
demonstration of the requirements for an apparently satisfactory solution
17518 and an engineering—scale demonstration
of tritium containment ‘4n the seCondary salt. Design and construction of
engineering—scale tests of several parts of the chemical processing con-

cept were under way when the program was discontinued in 1976. The nature

 

 
 

40

of the technical progress théffiwés fiéde;}in:éonjunction with the less-

, stringent.requirements of the low-power-density denatured system, suggest
that such a system could eventually be sucqeéefuliy.defieloped._chqver,
‘_Vsubstantial time .and effort would be required to develop the MSR into a

licensable, commercially acceptable system.

_ REFERENCES ...

1. R. C. Robertson, Ed Conceptual Destgn Study of'a Stngle-FZutd "
- Mblten~SaZt Breeder Reaator, ORNL-4541 (June 1971).' .

2.7 P. N. Haubenreich and J. R. Engel '"Experience with the Mblten-Salt
o 'Reactor Experiment," Nuel. Appl.:Tech. 8(2), 118" (February 1970).

| . 73;f Ebasco Services Inc., 1000 Milte) Molten-Salt Breeder Reactor Con-
o ceptual Deszgn Study, Final Report, Task 1 (February 1972) . |

4; L. E. McNeese and Staff, PTOgram Plan fbr Development of'MbZten-Sth
’ Breeder Reactors, ORNL-5018 (December 1974). RN

5. Molten-Salt Reactor Program, Sbmmannual PTOQTQSS Rbport fbr Perzod
Ending August 31, 1971, ORNL-4728, pp. 178-83.

6. L. E. McNeese, "Fuel Processing," Chap. 11, pp. 33163 of The De-
velopment Status of Molten-Salt Breeder Reactors, ORNL-4812 (August
1972). _

7. G. T. Mays et al., Distribution and Behavior of Tritium in the Cool-
ant-Salt Technology Fuctltty, 0RNL/TM~5759 (April 1977)

8. W. R. Grimes, "Molten-Salt Reactor Chemistry,” Nucl Appl Tech 8
137 (1970) | _

9, L. M, Ferris et al.,r"Equilibrium Distribution of Actinide and Lan-
_ thanide Elements between Molten Fluoride Salts and Liquid Bismuth
~ -Solutions," J. Inorg. Nuel. Chem. 32, 2019 (1970)

w 10. ”F J. Smith and L. M. Ferris, Mblten—sult Redctor Program, Semi~
' annual Progress Report for Period Ending Fébruary 28, 1969, ORNL-
: 4396, p. 285. - .. _ . , e !

11, L. E. McNeese, Engineering Development Studies for Molten-Salt Breeder
... Reactor Processing, No. 5, ORNL/TM-3140 (October 1971).

12, H. F. Bauman et al., Molten-Salt Reactor Concepts with Reduced Po-
tential for Proliferation of Special Nuclear Mhtertals, ORAU/LEA (M)
77-13 (February 1977).

 
13.

14.

15.

16.

17.

18.

 

41

L. Brewer, Science 161, 115 (1968).

D. M. Moulton et al., Molten-Salt Reactor Program, Semiannual Pro-
ress Report for Period Ending August 31, 1969, ORNL-4449, p. 151.

MSR Staff, The Development Status of Molten-Salt Breeder Reactors,
ORNL~-4812 (August 1972).

An Evaluation of the Molten-Salt Breeder Reactor, prepared for the
Federal Council on Science and Technology R&D Goals Study by the U.S.
Atomic Energy Commission, Division of Reactor Development and Tech-
nology, WASH-1222 (September 1972).

H. E. McCoy, Jr., Status of Materials Development for Molten-Salt
Reactors, ORNL/TM-5920 (January 1978). .

J. R. Keiser, Status of Tellurium—Hastelloy N Studies in Molten
Fluoride Salts, ORNL/TM-6002 (October 1977).

 
 

"

 
had ]

37.
38.
39.
40.

70.
71.
72.
73.
74,
75.

76.
77.

78.

79,

80.

 

43

 ORNL/TM-6413
‘Dist. Category_UC-76

Internal Distribution

T. D. Anderson ‘ 41. R. S. Lowrie

Seymour Baron 42. R. E. MacPherson

D. E. Bartine 43, H. E. McCoy

H. F. Bauman 44, L. E. McNeese

H. W. Bertini 45. H. Postma

E. S. Bettis 46-50. W. A. Rhoades

H. I. Bowers 51. P. S. Rohwer

J. C. Cleveland 52. M. W. Rosenthal

T. E. Cole 53. Dunlap Scott

S. Cantor 54, M. R. Sheldon

J. F. Dearing 55. R. L. Shoup

J. R. Engel 56. M. J. Skinner

D. E. Ferguson 57. 1. Spiewak

M. H. Fontana 58. H. E. Trammell

A. J. Frankel 59. D. B. Trauger

W. R. Grimes 60. J. R. Weir

R. H. Guymon 61. J. E. Vath

W. O. Harms 62. R. G. Wymer

J. F. Harvey 63-64. Central Research Library
H. W. Hoffman 65. Document Reference Section
J. D. Jenkins 66-68, Laboratory Records .
P. R. Kasten 69. Laboratory Records (RC)

Milton Levenson

External Distribution

Director, Office of Fuel Cycle Evaluation, Department of Energy,
Washington, D.C. 20545

E. G. Delaney, Office of Fuel Cycle Evaluation Department of
Energy, Washington, D.C. - 20545

C. Sege, Office of Fuel Cycle Evaluation, Department of Energy,
Washington, D.C. 20545

S. Strauch, Office of Fuel Cycle Evaluation, Department of Energy,
Washington, D.C. 20545

K. A. Trickett, Office of the Director, Division of Nuclear Power
Development, Department of Energy, Washington, D.C. 20545
Research and Technical Support Division, DOE-ORO

Director, Reactor Division, DOE~ORO

H. W. Behrman, DOE-ORO

W. R. Harris, Rand Corporation, 1700 Main Street, Santa Monica,
CA 90406

S. Jaye, S. M. Stoller Corp., Suite 815, Colorado Building,
Boulder, CO '

W. Lipinskki, Argonne National Laboratory, 9700 South Cass Ave.,
Argonne, IL 60439

 
 

81.

44

'R. Omberg, Hanford Engineering Development Laboratory, P.O. Box

1970, Richland, WASH 99352°

82.
83-193,

E. Straker, Science Applications, 8400 Westpark Drive, McLean,
VA 22101 | _ :

For distribution as shown in TID-4500 under category UC-76,
Molten—-Salt Reactor Technology

"

"