WASH-1222.txt

 

WASH 222

      

 

 

e ‘?AN EVA LU ATION
OF THE
MOLTEN SALT BREEDER REACTOR

 

 

Prepared for the Federal Council on Sc:ence and Technology R&D Goa!s Study
(o By the U.S. Atomic Energy Commission , =
Dmsnon of Reactor Development and Techno!ogy j__

 

DISTRIBUTION OF THIS DOCUMENT lS UNTJWTED g

 

 
 

 

‘owned nghts

 

lEGAl NOTICE

 

' 'Tlus report was preparcd as an account. of work sponsored by the Umted '

States Government.’ Nelther the Umted States nor the United States Atomic

Energy Commission, nor any of their employees, nor any of their 1

contractors, subcontractors, or their emp!oyees, makes any warranty,
express or implied, -or assumes any legal liability or responsibility for the —

accuracy, completeness or usefulness of any information, apparatus, product -
or process disclosed, or repmsents that its use would not mfnnge pnvately ;

 

 
 

 

' prepared for the Fel

WASH-1222 -
uc-80

AN EVALUATION

'THE MOLTEN SALT BREEDER REACTOR

 

 

 

 

[ NOWICE—

:{ This report was prepared.as an account of work

| sponsored by the United States Government, Neither

-{ the United States nor the United States Atomic Energy

‘| Commission, nor any of their employees, nor any of ‘
I their contractors, subcontractors, or their employees, | -

;| makes any warranty, express or implied, or assumes any

| 1egal fiability or responsibility for the accuracy, com- | |
_i| pleteness or usefulness of any information, apparatus, o
[} product or process disclosed, or represents that its use o
.| would not infringe privately owned rights, - F
[ = -

. Tt S
“-_ September 1972

 

e;&i,Cofinciifon;sc;gnce and Technology -
o " R&D Goals Study 1 fon S
e e e e g ' gssion -
o By the U.S. Atomic Energy Commissi o
e Z'Divisizn ‘of Reactor Development and ".I'echno_lé e .

| ] inten { Documents, U.8. Ggvernm
For sale by the Superintendent o: g ;
~ Weshington, D.C., 20402 %, :

ent Printing Office

PISTRIBUTION OF THIS DOCUMENT 1S unumrr%y\ |

 
 

This report was prepared as input to the Office
of Science and Technology's Energy Research and
Development Study conducted through the Federal
Council for Science and Technology. The contents
represent the views of the panel members and not
necessarily those of the Office of Science and

+ Technology. '

 
 

1.

1I.
'111.
v

V.

_TABLE OF CONTENTS

INTRODUCTION ....".;'."...;;;.....-.............,..."....‘.-'-....'

SUMMARY -ooaoobbfitmelldoini)iOiic;ooeooo-ooooicobnaiooootoiio-

RESOURCE UTILIZATION ......CQI'C.Q.O...0.0.0...O.....t'.......o

HISTORICAL DEVELOPMENT OF MOLTEN SALT REACTORQ resnsssssnssanes

MOLTEN SALT BREEDER REACTOR CONCEPT DESCRIPTION .ivesiescascees

VI STATUS OF MSBR TECIiNOLOrY ...C.‘.....'...‘;‘...V.-.".'....V....l‘.

A.

B.

. 7 F'.

:L*-Hélj

,9111"

o

G,

INDUSTRIAL PARTICIPATION IN THE HSBR PROGRAM cccsvsesassaccncne

mNCIJUSIONS .l..O.'.................lfii......l.li‘."‘.‘......l.

MSRE = The Reference Point fof'Current Technolopv'........

Continuous Fuel Processing = The Kev to Breedinz ..........;

1, Chemical Process Development sevsvesecssesseinasaann

2, Fuel Processing Structural Mhterials cssosvensiscsnes

‘Molten Salc Reactor Design - Materials Requiremen:s_......

1, Fuel and Coolant SaICB A
2. Reactor Fuel Containment Materials .o.eeeevenneenen.
3. Graphite:..;.;i....;....-.;.....-.....;.......;.....
4, Other Structufal-fleteriele ..,........;.....;..;;...
Tritium ~ A beblefi_of}Control.;.......,...;..;.5..;,.....
Reactor Equipment and Svstems .Deve]_.opment teeterescacseens
1. Components ;.;..;}..;....;..;..............,........

2. Svstems 000..0'.000000..!.Q.t.'.cOtOCQO-ioytoo-oeo'-

'.Maintenance - A Difficult Problem for the'MSBR cessessases

'Safetv - Differentmlssues-for the MSBR,....;.............

Codes, Standards, and High Temperature Design Methods ;...

REFERENCES ..............Q....'..O.....I....Ql.i."'...lll...ll'

APPENDIX A .o._Q_coocoo’._c.qéeooiool.ooooioooeooQo.o...c0|e.oeo_ool.o

10
12
14

19

19

22
22

, 25

26

27

30
33
34
35
38

38
41

43
45
47
48
o

53

A-1

 

 
 

LIST OF TABLES AND FIGURES

Tables

I Selected Conceptual Design Data for s Large MSBR.,

-

II Important Dates and Statistics for the MSRE

~ III . Comparison of Selected Parameters for the MSRE and ~ :

1000 MW(e) MSBR ° o

FiguresiA

‘1. Single-Fluid, Two-Region Molten Salt Breeder Reactor

2. Ylawéheét"foriProcessf;g irsingiefrluid HSBR; o

coa s T L, -

17

20

.21

18

24

 
 

AN EVALUATION OF- THE' Mox:rm SALT BREEDER REACTOR

 

o _Ii Imonucnon :

: The Division of Reactor Development and Technologv, USAEC, wss sssigned

| . the responsibility of assessing the status of the technologv of the ’(

 

-i-'Mblteanalt Breeder Reactor-(MSBR) as.part of the Pederal Council of ;
d-Science and Technology Research snd Development Goale Studv.- In
"conducting this review, the attractive features and problem areas
1associated vith the concept hsve been exemined- but more importantly,

'the assessment has been directed to provide a viev of the technologv
snd engineering development efforts and the essociated government and

»induetrial commitments vhich‘vould be required to develop the HSBR

into a safe. reliable and economic power source for central otation

o appl_icstion._» -
'1;The MSBR conccpt, currently under study et the Osk Ridge National )
'storetory (ORNL) .is based on use of a circulating fluid fuel |
:resctor conpled with on-line continuous fuel processing. As presently
: envisioned, it wonld operate as a thermal spectrum resctor svstem
'-iutilizing a thorinmruraninm fuel cycle._ Thus, the concept vould offer
”:the potential for broadened utilizetion of the nation 8 nstural
{resources through operation of s breeder system emploving another |

""K_fertile material (thorium instesd of uranium)

 

Ihe long-term objective of eny neW'reactor concept end the incentive for

the government to support its development are to’ help provide a self-

 

L msTRIBUnbN OF THIS DOCUMENT IS UNL\M!TED_

 
 

Lol

sustaining, competitive industrial capability for producing economical
-power in a reliable and safe manner. A basic ‘part of achievement of |
this objective is to gain public acceptance'of a nev.form'of power
production.- Success in such an endeavor isfreQuireddto permitthe:fi
utilities and others to consider the concept as a viable option for B
generating electrical power in the future and to consider making the |
heavy, long-term commitments of resources in funds, facilities and
personnel needed to provide the transition from the early experimental

facilities and demonstration plants to full scale commercial reactor

power plant systems.

Consistent with the policy established for all power reactor development
programs, the MSBR would require the successful accomplishment of three

basic research and development phases'

. An initial research and development phase in which the basic
.technical aspects of the MSBR concept are confirmed involving-’
exploratory development, laboratory experiment, and conceptual
engineering. 7 - |
. A second phase in which the engineering and manufacturing
capabilities are developed. This includes the conduct of
. in-depth engineering and prooftesting of.first-offafkind:dfi
~components, equipment and systems. . These would then be

incorporated'into experimental installations and supporting

 

 
 

-3-
,test facilities_to_assnretadequatennderstanding*of design__"
and performance characteristics, as well as to gain overall
- experience associated with major operational, economic and
environmental parameters.: As these research and development'
to be resolved and decision_points.reached that would,pernit
i.development to proceed with necessary confidence.. When'the
,technology is sufficiently deve10ped and confidence in the
w}system was attained'_the next stage wonld,be the,constrncf _:,,
'f_;;tion of large demonstration plants. |
- . Al third phase in which the utilities make 1arge scale commitments
- to electric generating plants by. developing the capability to
.manage the design, construction, test and operation of these

power plants in a safe, reliable, economic, and environmentally

- acceptable manner. -

| Significant;experiencewith”the'hight Water'Reactor;(LWR),'theiflieh
Temperature.cas—cooled Reactor (HTGR) and the Liquid Metal-cooled Fast
'Breeder Reactor (LMFBR) has been gained ‘over’ the past two' decades
pertaining to’ the efforts that are’ required to develop and advance S
nuclear reactors to the point of pnblic and commercial acceptance;'f”'
_fThis experience has clearly demonstrated that the phases of develop— _

' ‘_ment and demonstration should be similar regardless of the energv id ';'

- concept being explored‘ that the logical progression through each of

 
 

-4-

the phases is essentidlctefi&'that’cempleting the worfi"thfough'the
three phases 18 an extremely difficult, time consuming end costly
-undertaking, requiring the highest level of technical management,
professional competence and organiZational skills. This,has_again
been demonstrated by the tecent'experience in the expanding LWR
design, construction and néefismg activities wh‘ich‘--'emphasfize'_'eiearly
the need for even strongef'technology.and engifieetingefforte than
‘were initially provided, although ‘the'ae were satisfactory in many
cases for thefifirst'experifiEhts‘ahd demonsttation'plants. The LMFBR
program, which is relatively well advanced in ite'developfieht,’tracks.
closely this'fiWRexeerience'ahd hes'furthef”teinforced”thiéheed as it
applies teFthe”techfielogy, defelopmentend engifieefing eppiicetiog“

areas.

It should also be kept in mind that the large backleg,of;commitmente
end the shortage of qualified engineering and technical managenent |
personnel and prooftestrfacilitieskin the'government,inmindustrv'aner
in the utilities make it even more necessarf»that all the reacter
systems be thoroughly designed and tested_before additienaliaignificant
cammitment.to,“andiconstguction_of,'commercial power plants are ..
mi'tia;ed.. i et o

. With regard to the MSBR, preliminary reactor designs were evaluated in |

WASH-1097 ("The Use of Thorium in Nuclear Power Reactors") based upon

 
e _75_

'the information supplied by ORNL Two reactor design concepts were

' _,1fconsidered --'a two fluid reactor in which the fissile and fertile i

- salta were separated by graphite and a single fluid concept in which

: the fissile and fertile salts were completely mixed This evaluation
, identified problem areas requiring resolution through conduct of anj_
intensive research and development program.% Since the publication of
'IHTWASH-1097 all efforts related to the two fluid system have been__

, fdiscontinued because of mechanical design problems and the development

-fof processes which would, 1f- developed into engineering systems,

. permit the - on-line reprocessing of fuel from single fluid reactors. -

ij At present,-the MSBR concept is essentially in the initial research

V'Vand development phase, with emphasis on the development of basic MSBR
technology. The technology program is centered at ORNL'where L
-”essentiallyall research and deve10pment on molten salt reactors has )

been performed to date. The program is currently funded at a level |

of $5 million per year.r Expenditures to date on molten salt reactor
technology both for military and civilian power applications have

;amounted to approximately $150 million of which approximately $70 million_

has been in support of central station power plants.‘ These efforts date

3 back to the l9&0's.fi1" S

 

=, - a o ST zef L, e B -
# e - - L z s :

lfi!{In considering the MSBR for central station power plant application, it

'5"ffis noted that this concept has several unique and desirable features.,rl

';«lf 1at the ‘same time, it is characterized by both complex technological and

 
 

-6-

practical engineering problems which are specific to fluid fueled
reactors and for which solutions have not been developed. Thus,r;
this concept introduced major concerns that are different in kind and
magnitude from those coumonly associated with solid fuel breeder
reéétsfs.“ The'development'of satisfactory experimental units and

' further consideration ofrthis concept for use as a commercial power.
plant uill require‘resolution of these as well as other problems which

are common to all reactor concepts.

As part of the AEC" s Systems Analysis Task Force (AEC report |
_WASH-1098) and the "Cost-Benefit Analysis of the v.s. Breeder |
H'Reactor Program" (AEC reports WASH-llZG and WASH-1184), studies
were conducted on the cost and benefit of developing another |
breeder system, parallel" to the LMFBR. The consistent conclu-
sion reached in these studies is that sufficient information is |
available to indicate that the projected benefits from the LMFBR
program can support a parallel breeder program.' However, these
results are highly sensitive to the assumptions on plant capital
costs'uith the recognition;evenyamong concepts in uhich.amplef -
experience exists, that capital costs and especiallygsnall estimated
differences in costs are highly speculative for plants to be built '
i15 or 20 years from now.- Therefore, it is questionable whether
analyses based upon such costs should constitute a major basis for
making decisions relative to the desirability of a parallel breeder -

effort.

 
 

-7-

.Enperienee in reactor derelopnentprogrems in this eountry and sbroad
' has demonstrated that different organizatfons, in evaluating the projected
, : _ . ) ,
eoets of—introdueing a reeétorJdevelopmentfbroéram end‘Earrving it forward
_.to the point of - large scale commercial utilization, ‘would arrive at
'different estimates of - the-methods, seOpe ‘of develonment and engineering
efforts,=and=the‘eosts'and'time‘required'torbring'that program ‘to a-stage
of successful large scale application and public acceptance.
Based uponfthe'AEC's'exfierienEEfwith'6ther”e6nn1ex'reaetor_deveionment_'
programs, it is-estimated that e'total governnent investnentfunifo""
about 2 billion dollars in undiscounted direct eosts* could be required
to bring the molten salt breeder or any parellel breeder to fruition as

a vidble, commercial power reactor.r A magnitude of funding up to this

'level could be needed to esteblish the necessary technologv and

I engineering bases obtain the required industrial capebility, and

 

advanee through a series of test facilities, reactor experiments, and
demonstration plante to a commercial HSBR safe and euitable to serve

as a major energy option for eentral station power generation in the '

utility environment._ffi

 

'*msa-nsa - Updated (1970) Cost-Benefit Analvsis of the v.s. o
Breeder Reactor Program, January 1972. : k

 
 

-8-

II, - SUMMARY

| The MSER concept is a thermal spectrum, fluid fuel reactor which

~ operates on rheAthoriunrnrsnium:fuelwcycle and when coupled with

on-1line fuel processing has the potential for breeding at a

meaningful level. ;The’nergededifferencesTinstheVconcept:es;comparedf'
,to“solidlfueled,reactors,;nske;the MSBER a distinctive alternate.
Although the4concept“hasTettrgctiverfeetures,.there:are:e.nunber,oftu

- difficult development problems that must be resolved; many of these

are unique to the MSER uhilegothers_sre pertinent to . any complex

reactor‘system,fi__

The technical effort accomplished since the publication of WASH-1097
and wASR-lO98 has identified and further defined the problem areas, ’

‘however, this work has not advenced the program beyond the initial

phase of research and development. Although progress has been made
in several areas (e.g., reprocessing and improved graphite), new

problems not addressed in WASH-1097 have arisen vhich could affect :

- the precticality of designing and operating a MSBR. Exsmples of

major uncertainties relate to materiels of construction, merhods fori -

control of tritium, and the design of components and systems along

with their special handling, inspection and maintenance equipment.

Considerable research and development efforts are required in order

to obtain the data necessary to resolve the uncertainties.'F; |

 
-9-

| Asédming that prabticél #plfitibfiéttblfifieseAproblfims cfin'be found, a
furthe:assesfiment would havetO"Be;madg-as;tothe_hdvisability'of
procéedingaté'the next'atagéof;the-develépfient.ptbgram; *In advancing
quthé-fiext_phase,.itwfiuld-bé-necesgary toadeveloo:agrgatly_exfianded-“'
industrial and utility.particibatibnnénd'commithgnt"along fiith a
sdbstafitial'inérease infgovernmeht sfipp6rt; Sfich bro&déned 1hvoive_
ment would requiré_anevaiuafiiénof:the'MSBRfIfi féfmSVOf already '
existing cdmmitmentstgsothéfnuéiearpawef~and'h1gh-prioritv efiergv

 
 

-10-

III. RESOURCE UTILIZATION
It has long beenvrecognizedrthat the imporfianée-of nuclear fuels for
power production depends initially on the utilization of thefnaturally
occurring fissile U-235; but it is the mpfe abundafit fertile materials,
‘fi;238 and Th-232, which will be the majfir source of nuclear power
genérated in the future.  The basic physica characteristics of fiséile
plutonifim firoduced from U-238 offer the potential for high breeding
gains in fast reactors;vandthe.potential.tO'expand greatly the
utilization of uranium resources by making feasible the utilization éf‘
additional vast quantities of otherwise uneconomic low grade ore. In
a similar manner, the basic-physics characteristics of the thorium .
,cycle will permit full utiliz;tion of-the nation's thorium resources
while at the same time offering the potential for breeding in thermal

reactors.

The estimated thofium reserves fire sufficient to supply the world's
. electric énergy needs for many hundreds of years 1if the thorium is
uéed in a high gain breeder reactor. It is projected thafi if this
quantity of thorium vere used in a breeder reactor, approxifiatelv

1000 Q (1 q = 1018

Btu) would be realized from this fertile material.
It is estimated that the uranium reserves would also supply 1000 Q*

of energj if the uranium were used in LMFBRs. In contrast, only 20 Q

 

*Uranium recoverable at U308 firice up to $100/1b.

 
S -l11-
would be available 1f thorium were used as the fertile material in

~an advanced converter reactor because the reactor would be dependent

o

" upon U-235 availsbility for fissile inventory make-up. (Note: a

conservative estimate is that between 20 and 30 Q will be used for

| 31e¢tr1¢‘P°fier'gefieratiofibétweén-now'andthé‘year2100.):

 

 
 

 

-12-
.IV.» HISfORICALDEVELOPfiENT OF‘HDLTEN SALT:REACTORS:

The inveatiggtipn_qf:mplcen_sglt_rgactorsbegan 1n,the'1qtg_1940's:gs

part of thg U;S.Aircraft Nuciegg ?:opulsiofl(ANP)Program._ZSnbsedfiefitly,
the Afrcraft Reactor Expg:imepfi(ARE).wfis built at Oak Ridgg_and_;n 195§_v:
it was opérated succgasgu;ly‘fngniqe dgys:g; power ;gyg;a up.:q‘,;Q
2.5 MA(th) and fuel cutlet temperatures up to 1580°F. The ARE fuel was a
- mixture of NaF, ZrF&, and UFA. The.moder#to: was_BeOfand the piping and

. vessel were constructed of Inconel.

In 1956, ORNL began to study mplten salt reactors for application as
centr@l station converters and breeders. These studies cgncludéd that.
graphite moderated, thermal spectrum reactors operating'oh'g.thdfiumf
uranium cycle were most attractive for economic fiower productibn. -pased;
on the technology'at that timé;-it was thought that a two—fluid_:eactof -
inQthch the fertile and fissiie salts were kept sebarate was tequired

in order to have a breeder system. The single fluid reactor, while not

a breeder, appeared simpler in design and also seemed to have the

potential for low power costs.

.Oye: the next few years, ORNL continued to study both the twd,fluid'énd_
‘'single fluid concepts, and in 1960 theidesignrof the'singléfiuid /
‘8 MW(th) Moltén Sal; Reactor Experiment (MSRE) was begun. The MSRErwas |
completed in 1965 and operated sficcessfully during the peribd 1965 to |

1969. The MSRE experience is treated in more detail in a later section.

 
-13-

Concurrent with the construction of - the MSRE ORNL performed reaearch
. and development ‘on means for processing molten salt fuele.l In 1967

- new discoveries were made whicb suggested that a aingle fluid reector_

could be combined with continuous on—line fuel processing to become a .-

- breeder system. Because of the mechenical design problems of the two -

fluid concept and the laboratory—scale development of procesees which;
) would permit on-line reprocessing, it was determined thet a ehift in -
'emphasis to the single fluid breeder concept should be made~ this

P .

system is being studied at the present. iff”"

 
 

~14-

V. MOLTEN SALT BREEDER REACTOR CONCEPT DESCRIPTION

The breeding reactions of the thorium cycle are:

232 233 B 233 B8 233,
e e —( ——————
Th + Th 57 5in. Pa 55 ra U

Because of the nuhber of neutrons produced per neutron absorbed and the
small fast fission bonus associated with U-233 and Th-232 in the |
thermal spectrum. a breeding ratio only slightly greater than unitv is
achieveble. In order to realize breeding with the thorium cvcle it is
necessary to remove the bred Pa-233 and the various nucleat poisons
produced by the fission process from the high flux region as quickly as
pdesible, The Molten Salt Breeder Reectof concept permits rapid removal
of Pa-233 and the nuclear poisons (e.g. Xe-135 and the rare earth
elements). The reactor is a fluid fueled;syseem containing UFA and
'l‘hF4 dissolyed in LiF -'Berg The molten fuel salt flows through a
graphite moderator where the nuclear reactions take place. A side
strean is cpntinuously processed to remove the FPa and rare earth

elements, thereby permitting the achievement of a caICulated breeding

ratio of about 1.06.
The MSBR is attractive because of the following:

1. Use of a fluid fuel and on—-site processing would eliminate the

problems of solid fuel fabrication and the handling, and

 
2,

3.

15-

_ shipping and reprocessing of spent fuel elements which are = °

associated with all other reactor types under active

consideration.

MSBR operation on the thorium-uranium fuel cycle would help = -
- conserve uranium and~;herium_resources_by utilizing thorium-

“reserves with high efficlency. -

The MSER is projeeted tb'have-attractive ffiel'eycle coets.,

_The major uncertainty in the fuel cycle eost is associated ‘

| with the continuous fuel processing plant which has not been

developed. ;

The safety issues associated with the MSBR are generallv )

- different from those of solid fuel reactors. Thus. here T

might be safety advantages for the MSBR when considering-'
major accidents. An aceuretefassesement of MSBR gsafety 1is

not possible today because ef the early state of deveIOpment.

.-Like other advanced reactor systems such as the LHFBR and
HTCR the MSBR.would employ modern steam technologv for power
e7ageneration with high thermal efficiencies. This would reduce

'the amount of waste heat to be discharged to the environment.’

 

 

 

 
-16-
Sélected conceptual design date for a large MSBR, based primarily on : . 7?

design studlies performed at ORNL, are given in Table I.

There are, however, éroblém areas associated with the.HSBR'which nust
be_ovetcome'beforg“the'poténtial of the-cohcept could:bé-atgaified.
These includg development of continuous fuel processing, reactor and
processing structural métérials, tr;tiumlcontrol méthpds,'reaCtor;
equipmefit and systems, maintefiancé techniques, safétv,téihnolozv, and
MSBR codes and standards. Each of these problem areas will now be
evaluated 1h_som§_defiii:?fi§1fi§ as.h feferefice,fibififrthe tégfifloldgv
 which was demonstrated fiy thé"Hbiiéh_Sglt'Réactot Expérifient‘(HSRE)
during its desigfi,'congtruction and operation at Oak Ridgé"énJ.Ehe
conceptual 3esizn paramétgrs pfesented in fable I and in Afipéndix A.

A conceptual flowsheet for this system is shown in Figure 1.

 

 
s v+ et

 

-1

. Table T &

' fSelected Conceptua1 Design Data for a Large MSBR

Net Electrical Powet, Mw(e)

“Steam Syatem  f:

Fuei Sait@' AR

T*,;iflbderator_}

 ‘3f;Breeding Ratio

i’iCore Temperatures,

';Reactor Thermal Power, MW(th)

:?fCompounded Doubling Time, Years

 

Primary.Pipifig.hnd'Vesse1fiM£§éfiai‘{ffjf5‘

o Specific Fissile Fuel Iuventory, KgIMW(e)

 

722 L, 16z BeF

1000
- 2240

3500 psia, 1000‘

442 netrefficiencv ‘
7

2’

127 ThF ._o 3% UF,

_;fflastelloy N
| f;ASealed Unclad Graphite

1.06

1.5

1050 inlet, 1300 outlet

 
    
   

 SINGLE-RLUID, TWO-REGION MOLTEN SALT BREEDER REACTOR

PRIMARY SECONDARY
SALT PUMP

SALT PUMP. oBi,-NaF
COOLANT SALT

 

 

  
   

11

 

 

 

T

 

 

 

 

 

 

 

 

 

 

   

 

=)

-8‘[_

    

 

 

 

 

 

 

 

 

 

 

 

 

   

4"1Aqnpr __d) i

 

     

TLiF -BeF, - ThF, - UF, |
FUEL SALT o

 

 

    

CHEMICAL. |
| PROCESSING |
PLANT

SALT _ femiEYUNENY) o mii ¥ 3 |

-: . . S GRA PH 'TE . ‘ ‘ . N R e ‘ 3

| - i - MODERATOR 1 o Y
N 8 REACTOR 1 b
3 » 1k HEAT 1 | :
1 . - EXCHANGER {RIHKL | - = t_z
. | "
o — ! % - | |

 

  

 

 

IDCOOF.

 

 

 

 

 

 

 

 

 

=

 

TURBO- |
GENERATOR

Figure 1

 
- -19-

 

VI. STATUS,OFLMSBR{TECHNOLOGY

MSRE - The Reference Point for Current Technolqu

'The Molten Salt Reactor Experiment (MSRE) was begun in 1960 at

~ ORNL as part of the Civilian Nuclear Power Program. The purpose'

of the experiment was to demonstrate the basic feasibility of

- molten salt power reactors. A11 objectives of the experiment

vere achieved'duringfits'succesanl'operation”from Jnnefl955'to |

December 1969. ‘These included the distinction of becoming the

first reactor in the world to operateqaolely'onlU4233.;’Some'of'
the more significant dates and statistics pertinent to the MSRE

are given in Table II.

In spite of the success of the 'MSRE, there are many areas of ‘molten

salt technology which must be expanded and developed in order to.'

iproceed from this small non-breeding experiment to a safe, reliable,

and economic 1000 MW(e) MSBR with a 30—year life. To illustrate

this point, some of the most important differences in basic design

'and performance characteristics between the MSRE and a conceptual

1000 MW(e) MSBR are given in Table III.V Scale-up would logicallv

be accomplished through development of reactor plants of increasing

. size. Examination of Table III provides an appreciation of the
o ,fscale~up requirements in going from the MSRE to a large‘M%BR. ‘Some

'??problems associated with progressing from a small experiment to a

commercial high performance power plant are not adequately

 

 
 

-20-

.. .Table II.
Important Dates and Statistics for the MSRE -

Datss: | B
Desigfl initiated ," f'_. o__:_o ’_.,',_. "_07:7,0‘7_0 .® ,:O O ‘., .Jiu]_'(y 1960 .

235 Fuel . o o« <+« o o . June 1, 1965

Critical with
Opetatian:at.fulljpcqer.- B,MW(fih)_; o o o o+ May 23, 1966

Complete 6-month UM o ¢ « o ¢ o o o o o ,fMarchjZO,,IQGB;rufi

End Operatfon with 23U fuel . . + + . . . o March 26, 1968
Critical with 2330 fuel . . .+ + + « « . . October 2, 1968 .

233

Operation at full pdwer with U fuel . . ._Jgnuafy,?B,;lQGQ N

Reactor ofiération terminated « « + ¢ . o . o December 12, 1969

Statistics:
rflours .critical .7. e ®» ¢ @ ® 8 o s & o s e+ 17'655

Fuel loob time circulating salt (Hrs). « o0 21.788
| 235, .. e o

/ - Equiv. full‘powér hours qith U fuel . . ; .9.005

233y fuel . . . 4,167

.Equiv. full power hdhfs_wiéh

 

 

 
 

,gii} i.

General |
'-~.1Thermal Power, MW(th)
Electric Power, HW(e)
7Plan;.1ifetime, yeers‘”gfg;g

Fuel Processing Scheme
Breedifiélkacio;?L¥L:;, |
BEEEEQ;

Fuel Salt S
Moderator ,

'*Reactor Vessel Material_j;V -

Power Density, m/n:er o
) Ex‘,t Temperltute, .F et e

- Temperature Rise Across Core.-°F 40 T
8 20

Reactor Vessel Height. Ft.

V,Reactor Vessel Diameter, thf, o

 Vessel Design ‘Pressure, psiafi
- Peak Thermal NeutronnFlnx,

,e'Neutrons/cmzdeec'

Ot.her Cong:onents and Svst:ems Data =t

*f'g unsealed ‘graphite

 

Nnmber of Primary Circuits ,w;fi;fi,,,rgrrifffw |
‘Fuel Salt Pump Flow, gpm :Fie.:f;*fl**rf*-*““ o

‘Fuel Salt Pump Bead .

. ,Intermediate Heat Exchanger
| Capacity, MW(th)

‘ff:Secondary Coolant Salt

"fSecondary Salt Pump Head, ft.
- Number - of Steam Generators

 

,7,,

';ffNumber of Secondary Circuits ,fg
78 300
‘Steam Generator Capacity, MW (th) 0 o

3:Secondary Salt Pump Flow, gnm;

B -21;
' Table 111

Comparison of - Selected Parameters of the MSRE and

1000 MW(e) MSBR 1’

__MSRE MSBR 17

8 | 2250

o 1000

' Off-line, batch
s & “processimg: o

On-line, continuous
processing

L | : ,;1,06,fl:,,‘.,_,

(No Th preSent) ST

'7 | 7

LiF—Ber--ZrF{'UF4

~Unclad,

Lif-BeF,~ThF,~UF,

Unclad,
_ sealed graphite

Standard Hastelloy-N Mbdified Hastellov-n

2.7 . - 22
iB86T T st
250

s 22
8.3-x 10M%

A g 150- e

LiF—Ber NaF-NaBF,
171,»,;;};;;.;@e,,;4i,: B

20,000

121

'_l/'Based on information from "Conceptual Design Studv of a Single Fluid Wolten
~ Salt Breeder Reaetor," 0RNL—4541 June 1971. , ,

 

 
 

=22~

represented by the cdmpéfiéon'breéented'1n'the-Table.f'Thefefdie;i

it is useful to examine ad&ifiona1>fapets’of MSBR technology in

more

detail. .

- Continuous Fuel Processi;g The Kgy to Breeding>””'

In order to achieve nuclear breeding in the single fluid MSBR it |

is necessary to have an qn—line continuous fuel processing'svstem;

This

would accomplish the foilowing:

Isolate protactinium-233 from the reactor enviromment so it

- can}decay"into the»fissiie fuel isotope uranidm-233'beforé

‘being transmuted into other isotopes by neutron irradiatien.

, Remove‘undesirable neutron poisons from the:fuei salt and

thus improve the neutron economy and breedinngerforfiance'
of the system.
Control the fuel chemistry and remove excess uranium—233

which is to be exported‘from the breeder system.

Chemical Process Development
The Oak Ridge National Laboratory has propoéed_a-fuel .
processing scheme to accomplish breeding in the MSBR, and

the flowéheet pfocgsses involve:

a. Fluorination of the fuel salt to remove uranium gs"Ufs.‘

 
 

 

| vhich would require injecting helium bubbles into the fuel

-23-

b. Reductiverextraction'of protsctinium by contacting the

salt with a mixture of 1lithfum and bismuth.

c.,Metal-trsnsfer nrocessing to preferentially remove_the
rare earth;fiséion”product‘poisons vhich would otherwise
hinder breeding performance..
The fuei“p£5&éésifig systen shown’ianig.VZ 1s in an esrlv stage
of development at present and this tvpe of svstem has not been

demonstrated on an operating reactor. By comparison, the MSRE

'required only off~11ne batch fluorination to recover uranium

from fuel sslt.

At this time, the basic chemistry involved in the HSBR

i ‘processing scheme has been demonstrated in leborstorv scele
-experiments.- Current efforts at Oak Ridge are being directed

. toward development of subsystems incorporeting many of the

required processing steps. Ultimately a complete breeder

Vprocessing experiment would be required to demonstrate the

system with sll the chemicel conditions and operstionsl

requirements which would be encountered with any HSBR

: Not shown on the flowsheet is a separate processing svstemrfllfi

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SALT UFq PROCESSED SALT
PURIFICATION] REOUCTION | e
' My . \
: EXTRACTOR
I
'ur. '
:- SALT CONTAINING RARE EARTMS I
‘ ‘ t
' oo - mn g e . - w— - o e v ma e mwe e e A A e s e - . -
t | — |
I ! -
' ‘
t
‘ EXTRACTOR
‘ UF, :
REACTOR COLLECTION e
1 EXTRACTOR ..
i ) .
|
|- _ ‘
- , : Uy ' .
| ' I ' ' : - ‘ - .
L —on{ HYDROFLUOR }—tm n.uomunon—-.L PA DECAY ]
| : | g } 4 o
: HF F, . -
il , ¥, —a PA DECAY | ey L
: 27| FLuoriNaTOR ™ e | Lo o
| ' |
Ul 'SALT .
o - 10 o
b — —_o| mEDUCTANT | WASTE -
| ADDITION '
Li

FLOWSHEET FOR PROCESSING A SINGLE-FLUID MSBR BY FLUORINATION-REDUCTIVE EXTRACTION AND THE METAL-TRANSFER PROCESS.. -

Figure 2

 

 

 

 

 

 

 

 

  

 

  
    

-__---__r-J

b

!
-4

e e, e ———— —

L_........_...__.—_——-f

o Bl;Li
"’ (0.8 MOLE FRAC.
Li)

 

1 Bi~Li

L~ - +DIVALENT .RARE

- EARTHS
b = — Bi-Li,

Bi-Li
EARTHS

o e e e v

{0.08 MOLE FRAC. |
L) )

— & + TRIVALENT RARE

-‘—--H-A_—_-------—————‘-'—-—--—-‘4—_-—-—-.—4-”-‘

 

-vz—

 
 

; - } ":2._5 =

' -p'oalt, allowing:them'to.circulate,in the reactor'sjstem until

' 'sfthey collect fission product xenon, and then removiug the

';bobbles-and ;enop,from,thekreactorusyotem._ Xenon is a highly

_ undesirable neutron poison which will hamper breeding perforn-

*?hsé'bv,cgpturing;neutféns-fihich;fibuid'otherwisebreed'new

;_ifoel.; Thio7concept“for-xenoo~stripping was ‘demonstrated: in

L ;principle by the MSRE, although more efficient and controllable

-stripping systems will he desirable for the MSBR. VThefxenon

: opoisoniogiip.thelMSRE;wasfredoced by;a;faetor‘ofnsix_byixenon

stripping; the goal for the MSBR is'a factor of ten reductien.

Fuel,Processing,structoral,Mhterialg“]w_

_ Aside from the chemical processes themselves, there are also

_developoeot_requiremeot§;osoociotedjwithacontaiooent materials

for the'fueluprocessing"aystems.;-Infparticoler;~liqoid-bismuth

;,presents difficult compatibility problems with most. structural

, metals, and present efforts are ooncentrated on using molybdenum

Land graphite for containing bismuth. Unfortunatelv, both-

_;molybdenum and graphite are difficult to use. for such: engineering_

r;;applications. Thus. it will be necessarv to develop improved

techniques for fabrication and joining before their use is

possible in the reprocessing system. w o | -

 
c.

-286-~

A second materials problem of the current fuel’proeesstng gvstém \"f.
is the éont&inment for'the fiuorinatioh3stéb {h which uranium is
volatilized from the fuel'aalt.‘*The fluorine and fluoride salt
mixture is corrosive'to”most'struétural'materials,‘inclhding
graphite, and present ORNL flowsheets show a "frozen wall" |
fluortnator which operates with 'a protective layer of frozen
fuel halt"covetifié1§ Héételloy—N-veasel wgll.'“This component
would require qéfisiflerablegfigineerin#'fievelqfiménf'béfére it is

truly practical for use in on-lime full processing systems.

Molten Salt Reactor Design - Materials Requirements

In concept, the molten salt reactor core is a comparatively
uncomplicated type of heat source. The MSRE_re&ctorcore;7for
example, consisted of a prismatic'atfdcturé of unclad graphite
moderator through which fuel salt flowed to be heated by the
self-gustaining chain reaction which took place as long as the
Balt was in the graphite. The entire reactor internals and fuel
salt were contained in vessels and hiping'made of Haétéllbv-u, a
high strength nickel base glloy which was developed under the
Aircraft'Nucleér'Propulsion‘Program. Over thé fouréfearrlifetime
of the MSRE, the reactor structural materials performed satis-
factorily for the purposes of the experiments although operation

>f the MSRE revealed possible problems with long term use of

 
 

,;27-

"Hastellov-N in contact with fuel aalts containing fiasion

| producta. -

" The MSBR application 15'fibré'aemandipg*inxfiany'fespec:a than the
'MSRE, and additional development work would be required in
" geveral areas of materials technology before suitable materials

—- -

- could become available. |

'fi,l._,Fuel and Coolant Salta

7.

:The HSRE fuel aalt was a mixture of LiF-BeF-ZrF, -UF

4Ty 1
L_proportions of 65. 0-29 1-5.0-0.9 mole %, respectively.

Zirconium fluoride was included as protection against dbz
,_precipitation ahould inadvertent oxide contamination of the
. aystem occur.: HSRE operation indicated that control of

oxides was not a major problem and thus it is not conaidered

necessary to include zirconium in future molten salt reactor
fuels. It ahould also be. noted that the HSRE fuel contained
. no. thorium,whereas the propoaed HSBR fuele would include
~ thorium as the feftile material for breeding. - With the
~:fipossihle exception of incompatibilities with Hastelloy—fl,

,the MSRE fuel salt performed aatisfactorily throughout the

life of the reactor. i:}f,ra~-

--t!The MSBR fuel aalt, as currently proposed by ORNL would be a

,;fdudxture of 7LiF—BeF2-ThF£—UF4 in proportions of 71 7-16-12-0 3

 
 

-28-

.-molé Z; fespéhtively.- Tfiis salt has‘amglting pfiint‘of about
930°F and a vapor pressure of less then Oéi mm Hg at the mean
_ope;afiing_pempqrature of 11509?., Italéq has about 3.3 times
the density and 10 times the viséqsity.ofyater.lts thermal
_cpndfic;ivit}”and;vplumetrichéat capacifiy are comparable to
water.

The high melting tember#tute’is'an;dbviousiimit&fion for a
gystem using thié galt, and theMSBR'is'limité&to‘high
temperafureJOperation}”’In>addit£6n,’thé‘lithium component
must be enriched in Li-7.in order to iilbd"nfitlear breeding,

': 91nce?natura11y occurring lithium cantéihs’about 7.5% L1-6.

 Li-6 1s findesitab{g"in the MSBR because df'itéitendency to

" ‘capture meutrons, thus penalizing breeding:perfqrmance{

'The'chemicalfand‘physical characteristics of the proposed
'MSBR ffiel mixture have beefi'ahd‘até"béifigminveStigated; and

‘ theyarerea;ohably weli known for unirfadiafiéd salts. The
'major unknowns are asscciated with the féactor'fuel hffet it
" has been;irrédiht&d;‘lfbr'example,'not:énough'is%known about
the behavior of fission products. The ability'tb predict
fission producthghafiior ié 1mportgnt to plant safety¥

" operation, and maintgfiance. " While "‘t.:t‘ie. MSRE provided im_zch_

- useful information, there is still a need for more information,

Kiié

|
i
L BRI 3 TSI Estiot 121 3 1

 
ur'wrtzg_ .

'1fihfticnlarlY'witfi”ieéard’to?thE?fate”of tbe”so-called "noble
':metal“ fission products such as molybdenum, niobium and
’*'others vhich are generated in substantial quantities and

jwhose behavior'in.thejsystem is notnwell'understood.

o

'fb'A‘more complete understanding of the physical/chemical

'flicharacteristics of the irradiated fuel salt is 3180 needed.

As an illustration of this point, anomalous power pulses were

fobserved during early operation of the MBRE with Uh233 fuel
itwhich vere attributed to unusual behavior of helium gas

'}djbubbles as they circulated through the reactor. This
fbehavior is believed to have been due to some physical and/or
brchemical characteristics of the fuel salt which were never

fully understood.fi Out-of-reactor work on molten fuel salt

1whfission productrchemistrv-isficurrently under wav.t Eventuallv,

VI_lthe behavior of the fuel salt would need to be confirmed in an

fPPgratifigr?fifiFFQF' - ”,,f

;;The coolant salt in the secondary svstcm of the vsnn was of

7LiF-3423eF While this coolant

'°Pmolar composition eez
fperformed satisfactorily (no detectable corrosion or reaction :

;;could be observed inethe oecondarv svstem), the salt has a

aa7high melting temperature (850°F) and is relativelv expensive.

"inf’~Thus, it may not be the appropriate choice for power reactors

 
 

-30-

for two reasons: - (1) larger vo;umea;of coolant salt will be

iuaea to generate steam in the MSBR, and (2) salt,températures
_in the steam generator should be low enough, if poasible, to

‘utilize conventional steam system technology with feedwater

temperatures up to gbout 550°F. The operation of MSRE was
less affected by the coolant salt melting temperature since

it dumped_the,BVMW(tfi) of heat via an air-cooled radiator.

The high melting temperaturés of potential coolant salts

remain a problem. The current choice is a eutectic mixture

-of sodium,flufiride_and sodiu@ fluoroborate vith a molar |

composition of 8% NaF-%2% NaBF4; this saIt melts at 725°F.
It is comparatively'1nexpensive and has satisfactorv heat

transfer properties.

However, the effects of heat exchafiger'ieaké'between the

coclant énd fuel saits, and between thé'coolant salt and

steam systems, must be shown to be tolerable. The

fluoroborate salt is currently being studied with respect

" to both its chemistry and comp&tibility with Héstellov—N.

-

- Reactor Fuel Containment Materials

i

A prerequisite to succesg for the MSBER would be the abilitv

to assure reliable and safe containment and handling of molten

K;;;

 
. -constru(:ted.

=31~

_ fuel salts et all times during the life of the reactor. It

uould be necessary, therefore, to “develop suitahle contain~

" ment materials for MSBR application before plants could be

\A'serious.questionconcerning'cofipatabilityof_Hastelloy-N with
the constitueete,of!itrediated~fue1;salt~was raised by the post-
operation'examifietionédfetheeMSRE:ifi.1971._;Although-the MSRE

-materials:fier£0tméd;setisfacterily_for'that,svetem_during its

operation, subseQuent'examinetion of .metal which was exposed to

MSRE fuel salt revealed that the alloy had experienced inter-

' gr&hfilet'etteck"tojaeptfiswdf;ebdut'0.00T inch. The attack was
" not obvious unitil metal specimens vere tensile tested at which
" time cracks Opened up as the metal was strained. ‘Further
”iexamination revea;edjthat’several fission products, including
T teildrifim;-he&jfienettEtédthe’metalito depthscemparable to
’ 7“thdse“of'thé’cieek§:5Aé'théffifeseht°timé, it is thought thet
" ‘the 1nter§tahu1aé'at535k‘wae due to the‘presencetef tellurium.
_Subsequent laboratory tests have verified that tellurium can
?ptoduce, under certain conditions, interyranular cracking in

"’”‘Hastelloy N.

*--Althoughf:fie-ltmited}pefietrationof eracks;p;esented no'problems |

.fer;the;MSRE,;cdncetnfflbw;exists,withsrespeetetp thé chemical

 

 
-32-

compatability offiHastgllpyffi_andMSBR_fugl salts whenaubjected
to the more stringentHSBRreqfiirements_of higher power dénsity
and 30-year life. If the observed intergranular gttack was
indeed due to fiséion,firoduct attack of the Hgsgg;loy-n,athen

- this material may not. be suitable fof'either the’#ifiing or the
éessels which Quuld be éxposed to much higher fission proddct

" concentrations for‘longer-peridds of time. Efforts are ufider'
wvay to under;tandfandexpléinthe cracking>prdb1€m, and Eo—
'detetmine whether altéfna;e reactor containment m§té;1a1s

should be actively considered.

In addition to the intergranular corrosion problefi, the stan&afd-
_ Hastelloy-N used in the MSRE 1s not suitable for use in the MSBER
";becauserita mechanical properties detgriotate-;o'anvunacceptable
- level when subjected to the higher neutron doses vhich would

occur in the highé:_power density,'}dnge:—life MSBRL:'The problem

is thought to be due mainly to impurities in the metal which are

transmuted to helfum when exposed to thermal neutrons. The helium

is believed to cause a deterioration of medhanicél'profiérties by

~1its presence at grain boundaries within the alloy. It would be

necessary to develop a modified Hastelloy-N with improved 1rradia§'_

tion resistance for the MSBR, and some pxogress is being made in
that direction. It appears at this time that small additions of

| cettainfelements§‘such:as/titénium,'improVe’the irradiation,

 
 

'-335

-

' performance of Hastelloy=N substantially. Development work on.

" modified alloys with improved‘irradiatiOn‘resiatance?ia

3,

- currently under way.

Graphite

*j:Additional developmental effort on two problems’ 1is required to

roduce grAphites auitable for HSBR application. The first is

o associated with irradiation damage to graphite structures which

B results from fast neutrons. Under high neutron doses, “of the

"porder of 10

22 neutrons/cmz, moat graphites tend to become

J:{dimenaionallv unstable and gross awelling of the material occurs.

':ffBased on testa of amall graphite samples at ORNL, the best .

_about 3 x 10

'commercially available graphites at this time may be ueable to

22 neutronslcmz, before the core graphite would have

”to be replaced. This corresponds to roughly a four-year rraphite

| lifetime for the oanr reference design. While this might be

,acceptable, there are still uncertainties aboutathe fabrication

and performance of large praphite pieces, and additional work

--"'would be required before a four-year life could be aasured at

o *the higher MSBR power denaities how beinp considered.fl In any

”=fu;ewent, there would be an obvious economic incentive to develooe

: 5:ef-; longer lived graphites for HSBR application_since a four-year

B f;;;flife for graphite ia eatimated to represent a fuel cvcle cost

 

 
 

-34-

‘penalty of about 0.2 mills/kw-hr relative to a ovotem with

thirty year grefihige_lifef. |

The second major problem_associated with praphites for MSBR

_application is the development of a sealing teehnioue_whieh'

will keep xenon, an,undeeireblerneutron,poison, from diffusing

- 1into the core graphite where it can capture neutrons to the

4,

? detrinent of,breeding'pe;formence. _while graphite_sealing

. may not be_neeessafy to achieve nuelear-bfeeding_in_the,MSBR,

the use of sealed graphite would certeinlvennaneebreeding
performance. The economic incentives or penalties of graphite
sealing cannot be assessed until a suitable sealing process is

deVeloped.

.Sealing methods whieh have been 1nvest1gated to date inelude‘

_pyrolytic carbon coating and catbon 1mpregnation. Thus far,

however, no sealed graphite that has been tested remained

sufficiently 1mpermeeb1e to gas at MSBR design 1rradiation

doses and researeh and deveIOpment in this area is continuing.

Other Structural Materials

" In additfon to the structural matefiais"reqnifenenteifor the

. reactor and fuel processing systems proper, there are other

components end-systems:whiehheve special materials require-

ments., Such components as the primary heat exchangera and

 
;:”?351
..fateam-generators}must.functionzwhile-1nfcontact with‘twou.
- different working'fluids.l,_”_:r |
_;-At the present:time, Hastelloy-N is considered to be the most
| promising material for use in all salt containment systems,__
including the secondery piping and components. Research to date
3 indicates that sodium fluoroborate and Hastelloy-N are compat-
ible as long as: the water content of the fluoroborete is kept

, low, otherwise, accelerated corrosion can occur.' Additionalr

rtesting;would be needed and is_underway. a _'

:Hastelloyéfi5ha§'not'heeniadequstely-evaluated‘for service under
a range of steam conditions and whether it will be a auitable

material for use in steam generators is still not known.

D. Tritium - A Problem of Control
Because of the lithium present 4in fluoride fuel salts, the present MSBR
concept has the inherent problemfof generating tritium, a radioactive
isotope-of hydrogen. Tritiumfigiproducedpby”the'followine reactions:
" % @a) u
@ 7Li,(n,an) 3.
'lDue primarily to these interactions, tritium.would be. produced st a rate

'_of about 2400 curies/day‘in s;lOflOVMWeVMSBR. This compares with about

 
- 3_()" |

| 40 to 59 curiesfdav’for ltpfit:watéf;"fiafi—conlefl, and"faqf‘h?eeder
reactors, in which tritium i#rproduced bfimaxilv as 5 16W'§{é1d fission
produét. Tritium productinn in heavv water reactors oF.connarable size
is geuetallv in the range 3500 to 5800 curies/dav dfiéTtfl-figufrnh inter-

actions with the deuteriun nreqent 1n heavv water.

1

Tn‘Furfher éompound t}e prnhlem tritiun diffuqes readilv>throunh :
Haqtellov-N 1t elevnted temperatureq; Aq a reqult. 1t mav be difficu]t
tn.prevent tritiun from diffuqina throuph the pipinvy;na co%@nnonta of
the YSBR system (such aqrheat exchanners) and gvenfiuallv_reachinqrtho
steam system whererit m{gbtabe:dificharned tp»the‘eqvifpnrent_aq tritfintpd

water.

The problem of tritium control in the “SBR {g being studied in detail at
NMPNL. The followinp are beins considered as potential methods for

tritium control:

1.  Exchanging the tritium for anv,hfidronen present in the secondarv -

conlant, therebv retaining the tritfum {n the secondarv coalant.

2. Using coatings on metal surfaces in order to inhibit tritium

diffusion.

 
 

[ J
Mt !

;37l--

"3l-'70perating the reactor with the salt more oxidizino, therebv SR

caueinp the Formatton oF tritium fluoride whtch could be

removed in the-off—nas svstems.

&, B qunr ‘a different qecondarv coolant e.g., qodium or helium,'ahd-

'“proceqqtny th{q coolant to remove tritium.'

—45;<:-Usingwanother;int?rmedlaté loob between the fluorohorate and

C ‘steam: to "gettel‘" tl‘itim-' mELE

6. Using duplex tubing 1n either the heat exchanrer or qteam

generator with a purge pas between the walls.

Of theee notential solutions, the use of an additional intermediqte loop
'hetween the ‘secondary and eteam qystems 1s" conQ1dered the most effective

'f'method technicnllv. but 1t would also be expensive dup to -the additinnal

equipment required and ‘the losq of thermal eFF{ciencv. T

From an econom1c viewnoint, the mOQt deqirable qolutinn is: ‘one which

does not - sianiflcantlv complicate the svetem, euch as’ etchanpe of
.lrzétritium for hvdrooen preqent‘in the eecondarv coolant.5 Thiq techninue j
'ifis beino 1nvestigated aq part of the flRNL effortq on trttiun cheniqtrv

'lifThe tritium retentinn nroblem may be eased by the lnw nermeabilitv nf

g

 

 
 

 

~-38-

oxide coatings which occur on steanm genera;or‘mhterials in contact with

steam, and this is also being investigated at ORNL, -

E. Reaétdr Equipment énd Systems Development

While the MSBREwould utilize some existing engineering technology from

~ other reactor types, there-fite spéciiic components and systems for which
additional development work is required. Such work would have to take
into acéount the induced activity that those components would accumulate
in the MSBR system; i.e., special handling‘and maintenance equipment would
algo need to be developed. The previous discussion has already dealt'f

with a number of these, such as fuel processing components and systems,

but additional discussion is appropriate.

1. Components
As indicated in the Table III, a number of components must be

scaled up substantially from the MSRE sizes before a large MSBR
is possible. The development of these larger components along
with their specialihandling and maintenance equipment is proB-
ably one of the most difficult afid costly phases of MSBR
development, However,nreliable, safe, and,maintaihable
compbnénts would need to be developed in order fqt afiy,reactor

system to be a success,

The MSBR pumps would likely be similar in basic design to those

for the‘MSRE, nanely, vertical shaft, dverhung impeller pumps.

 
-39-

- Substantial exneriencehee;been-gained-over_theyearsin the
design;ffabricationandloperation'of smallerrsalt pumps, but
“the size ‘would have" to be 1ncreased substantially for MSER
.’application. The - deve10pment and ‘proof testing of such units
' ?’along with theirfhendling*and”ma;ntainence-equipmentjand test

facilities are expected to be costly‘and time consuming.

The intermediate heat exchangers for'tfie'HSBR'nust‘bcrform.with
a minimum of salt 1nventory 1n order to'improve the breeding

performance by lowering the fuel 1nventory.. Special surfaces to

| ?“enhance'heat tranéfer'would=he1p-achieVe’this,*andimore studies

1;wou1d be in order. Based on previous experience with ‘other reactori
-systems;'itvis believedfthatvthese-units'would,require a diffi-

“* cult -development and proof testing effort. -

_ The steam generator for'MsBR applications is probably:the most
difficult large component to develop_since it represents an
*'item for which there has been almost no experience to date. It

“i8 beiieved~that afdiffiéfilt'dévelopfiefit’afiA»proaffiesting pro-

| f?*gram would be needed to provide reliable and maintainable units.

"%ifAs discussed previously, the high melting temperatures of

“_:“candidate secondary coolants, such as sodium fluoroborate,

.:f‘present problems of matching with conventional steam system |

-wfiitechnology. ~At this time, central station power plante utilize

 
 

-40-

feedwater'temperatufesonly_ufi-to about SSO“F;',Thetefore,
coupling a conventional feeéwater systeg;to,a_necondary'
coolant whiéh freezes at 725°F presents oSvious;problems in

. design,and‘conttol.;;It_might yevneQQSSary to provide modifi-
éations_to-convegtional-steam-éystem designs to help tésolve
the problems, Because of these factors, a study ;élated to
the design of steam generators has been initiated at Foster

i Uheeler.Corpdration.”~

Control rods andid:ives for the HSBR.vouldlfilso need to be
. develoned,__The,HSRElcontfolquds were air cooled and operated
inside Hastelloy-N ;himbles which protruded‘down_into;the fuel
salt. The MSBfi,fiould,requite‘more efficient cooling due to the
vhigher-power densigieS;involved.r Presumably rods and drives
- would be needed'whidh peruit fhé rods to fiofitact.and be cooled

by the fuel salt.

- The salt valves for large MSER's represent another develfipment |
‘problem, although the freeze valve comcept whiéhawaa employed
successfully in the MSRE could likely be scaled up in size and
~utilized for many MSER applications. Mechanical throttling
_.ydlvgé,would}also be needed for the MSER salt s&stems,_even
- though no ;hrottling(valvé gasuégd vith the MSRE. Mechahical
shutoff;valfiesnfor.aaltfsgstems,_ifmrequired;,woulfi have to be

developed;»:

 

 
 

41

';'”OtherJCOmponents:whichfwouldreqnire considerable-engineerinp

' development and testing include: the heiintUbblegeneratore and
- ‘gas strippers'oh1Ch-arefnroposed”for use in removing the fission

';product xenon from the fuel ealt. "Research ‘and denelopment in

this area is currently‘under‘way ‘as part of the technology

program at ORNL, e

stte

- +The integration of all required components into a complete ‘MSBR

/”Vcentral station power plant would involve a number of systems for

- which development work is still required. Itrahouldfbe noted
 that some components,_sndh as pnmps and control rod drives would 1
frequirertheir,owneindividual systems for functions-such as

" cooling and lubrication.’ °

' *“Given the required components and materials of construction, the
”basic reactor primary and secondary flow systems can ‘be designed.
, However, the primary flow eystem'would require supporting systems
-fiiéfifor continuous fuel processing, on-line fuel analysis and control
i of salt chemistry, reactor control and safety, handling of radio- |
;ifl':active gases, fuel,draining from every possible holdup area in .

¥A1%T?Components and equipment, flftthEGt control, and temperature —

’f;fcontrol;dnringflnon—nnclearmoperations." |

 
 

-42-

fThe éontifiuous,fuel'prdcessing»systems.PrODosgd,tojdate are | _ §‘#j
quite complicated and:include.grnumber bf subsystems,all of
which would have to 6perate satisfactorily,V1thin the_constrainfs
- of economics, séfety,»and relidbility}g The effecfs‘of off-desisn
. conditions on thesélsystemsiwofild_have«to be4undér§fiood so that
control woulé be possible to prevent inadvergént:contaminagion

of the prim#ry system by undesirable materials, |

The[ffiel drain system 1s important to both~operation:and safety
. since it would be used to cofitainthemoltenjfuel_whenever a
need arises to drain the primary systenm or;any-camponent or

- instrument for maintainence or inspection. Thus, additional
systems would be required, each with its own system for
maintaining énd controlling temperat;res. The fuel salt drain-
tank would have to be equipped with an auxiliary cooling system
capable of rejecting about 18 MW(th) of heat should the need

-arige to drain the salt immediately following nuclear operation.

The_sécondary'coolantsystem would aiflo réquire subs%stems for
draining and controlling of salt-chemistry.and'température. In
. addition, the secondary loop might require systems to control
vtritiuh-and»torhahdle-the consequences of,s;eam generator or

heat exchanger leaks;.

 
f43"‘
efliThe steam system for the MSBR might require a departure from
conventional designs due to the unique problems associated with
using a coolant having a high melting temperature. Precautions _
fwould have to be taken against freezing the secondary salt as it
travels through the steam generator, suitable methods for system
fstartup and control would:necd‘to be incorporated._-ORNL has
-Aproposed the use*bf:; éfipefeiitibai-siean'éistem which'operstes'
at 3500 psia and provides 700°F feedwater by mixing of supercritical
steam and high pressure feedwster. This system would introduce
Wmajor new development requirements becauae it differs from |
conventional steam.cycles._i |
F. " Maintemsnce - A Difficult Problem for_the MSER
Unlike solid fueled resctorsrin which the primary-system contsins
isctivstion products and only those fission products which msy leak from
defective fuel pins, the MSBR would hsve the bulk of the fission products

"dispersed throughout the reactor system.r Because of this dispersal of

'i'fradioactivitys remote techniques VO“ld be required f°r many maintenance

 

functions if the reactor were to—have an acceptable plant svailsbility in

the utility environment.p"'

B The HSRE’was designed for remote msintenance of highly radioactive.éw
rcomponents, however, no najor‘naintenance problems (removal or repair of

ilarge components) were encountered after nuclear operation was initiated

 

 

 
 

-44-
Thus, the degree to which the MSRE experience on maintenance is

| applicsble to large commercial breeder reactors is open to question.

Ashssheen evident.in plsnt layoutwork_on nuclenriggcilitiesto date._‘
_this requirement for remote maintenance will significantly affect the
ultimate design and performance of the plant system. IhemHSBR would
require remote techniques snd tools for inspection, weldiug and cutting
of pipes, mechanical sssembly and dissssembly of components and systens,
- andrremoving, transportiug endqhendling.large;component items after they
become’highlyvradioactive,;lhe_removsl and replacement of'core
internels,‘such as grsphite,'might pose difficult msintenance.problems
because of the high radiation levels involved snd the contamination

protection which would be required whenever the primary system is opened.

_mAnother potential problem is-the sfterheat.generation'hy"fissionproducts
which deposit in components such as the primary heat exchangers. ‘ |
Auxiliery cooling might be required to prevent damage when the fuel sslt is
drained from the primary system,gendla requirement”for;such_cooling m?uld

further complicste inspection and maintenance operations,

In some cases, the inspection and maintenance problems of the MSER could
be'solvedlusing present technology snd particularly erperiemeeigsined from
fuel reprocessing plants. However, additional technology development .

| would,be_requiredrin other areas, such as remote cutting, alignment,

 
-45-
cleaningtand‘welding'of*netalimembers.: Dependingvto'somefdegree on the
fparticular plant arrangement, other special tools and equipment would

l also have to be designed and developed to accomplish inspection and.

TAmaintenance operations."~" e

Infthe_finalvanalysis,»the‘develonment'of adequate insoection-and-4
-_maintenance techniques and_procedures,and hardwaretor»theMSBR?
hingeaon_the.success_of_otnertacets of»the proéram, such as'naterials
and component develdnnent;fandion_thefreduirement that'adequateficare be
"taken”during plant'desién"tolassure thatcall sfstems and“componenta
which would require maintenance over the life of the plant are indeed

7 maintainable within the constraints of utility operation._-'

Safety=- Different Issues fot ‘the MSBR

| The MSBR concept has certain characteriatics which might provide
'_advantages relating to- safety, particularly with resoect to poetulated
major types of . accidenta cutrently considered in licensing activities.
- Since the fuel would be in a molten form, consideration of the core

'meltdown accident is not applicable to the fiSBR. Also,—in the event

--1of a fuel spill secondary criticality is not a problem since this is

 

a8 thermal reactor system requiring moderator for nuclear criticality. o

tLOther safety features include the fact that the primary system*would

-T,Operate at low preasure with fuel salt that is’ more than IOOO'F below

 
 

-46-

its boiling éoint, that fission product iodine and strontium form

stable compounds in the fluoride salts, and tfiat the salts do not react
rapidly with air or water. Because of the éontifiuous_fuel_prpcessing,-
-the need for excess reactiviéy would be decreased and some of the fission
products would be continuously removed from the primary system. A prompt.
fiegativé temperature coefficient of reéctivity;is_also a characteristic |

of the fuel salt.

Safety disadvantages, on the other hand, include the very high radio-
active contamination which would be ptgsent,throughout_:hg primary .
system, fuel processing plant, and all agxiliéry;primafysystems_such

as the fuel drain and off-gas systems. Thus, cofitainment.of*these'q
é&stems would have to be assfired. Also, removal of>decay‘heat from fuel
storage systems would have to be provided by always_ready'and:re;iable_
‘_éooliug-systems, particularly for the fuel drain tank and the Pa+233 decay
tank in the reprocessing plant where megawatt éuantities of'decay.heat,r
must be removed. The tritium problem, already discussed, would have to

be controlled to assure safety.

Bgsed_on the present state of MSBR technology, it is not possible. to
provide a complete assessment of:MSBR'safety;:elgtive to other reactors.
It can be stated, however, that the saféty issues for the MSEBR aré
generally different ffomflghose for solid fuel reactors, and that more
detailed design work must,bé;donerbefore the aaf?ty éd#antageg and -

diséflvantagestof the MSBR could be fully evaluated.

 

 
a7-.
H. Codes, Standsrds wandHigh.Tempersture Design Methods |
Codes and standards for MSBR equipment snd systems must be developed in
'conjunction with other research snd development before large MSBR 8 can
be built. In particular, the materials of construction which are
'currently being deve10ped and tested would have to be certified foruse

infinuclear power‘plant:spplicstions,

rfThe need for high temperature.design technology is a problem for the HSBR
_-as well as for other high temperature systems._ The AEC currently hss
| under way a program in support of the LMFBR which is providing msterials
'_'dats and strnctural analysis methods for design of systems employing
;various steel alloys at temperatures up to 1200°F. This progrsm would
Vgneed to be broadened to include MSBR structursl materials such a5
'Hsstelloy—N snd to include temperatures as. high as l&OO‘F to provide
i'the design technology applicable to high-tempereture, long-term |
persting conditions which would be expeeted for MSBR vessels, components,

‘and core fitrHCtureS+_ - W;es

 
 

-48-

. VII. INDUSTRIAL PAR"ricIPA'rIoN "IN THE MSER PROGRAM

 

- Privately funded conceptual design studies and evaluations of HSBR _

technology were performed in 1970 by the Molten Selt Breeder Reactor
Associates (HSBRA), a study group headed by the engineering firm of
' Black & Veatch and including five midwest utilities. The HSBRA con-
cluded that the economic potential of the MSER is attrsctive:reletire
to 1ight water reactors, but they recognized a number of problems‘which
must be resolved in order to realize this potentiel. Sincethat‘time

the HSBRA has been relatively inactive. .

A second privately funded organization, the Molten Salt Group, ie headed
Tby Ebasco Services Incorporated and includes five other industrial firms
and fifteen utilities. In 1971 the Group completed an evsluation of the
HSBR concept and technology and concluded that existing technology is
sufficient to justify construction ‘of an MSBR demonstretion plent
althOngh the performance characteristics-could;not'beunredictedynith

: confidence. Additional support for further studies hes“recentlfi been

committed by the members of this group.

In addition to these studies, manufacturers of graphite and Hastelloy-N

" have been cooperating with ORNL to develop improved materials.

There has been little other industrial participstion in the HSBR

Progrsn aside from ORNL snbcontractors. At the present time, there are

 
-49-
o ORNL eubcontraota in effect., Ebeaco SerVioes,'Ino.. utilizing the
rfindustrial firms who ero participante in the Hblten Salt Group is
performing a design and evalnation etudy. Foster Wheeler Corporation is
,ourrently performing design studies on steam generatore for HSBR

vapplication.

A number of factors can.beidentified thch tend to. limit further

{ndustrial involvement at this time, namely:

1. The existing major industrial and utility commitmehte,to_the

‘LWR, HIGR, and LMFBR.

2. The lack of incentive for industrial investment in supplying
fuel cycle services, such as those required for solid fuel

reactors.

+ 3. The overwhelming'manufacturing and opereting experience with
eolid fuel reactors in contrest with the very limited involve—

'_1ment with fluid‘fueled¥reeetors.

e Attlthe less advanoed atate of‘HSBR technology and the 1a°k °f

",demonstrated solutions to the major technioel problems f_

",'associated*with the HSBR concept. -

 
 

-50- .

It'dhOfild be noted that these factors are also relevant considerations
 in establishing the level of governmental support for the MSBR program
vhich in turn, to some extent, affects the interest of the manufacturing

and utility industries.

 
~ -51-

l;:vtii;' CONCLUSIONS -
The Holten Salt Breeder Reactor, Af successfully developed and marketed
could provide a useful supplement to tbe currently developing uranium-
plutonium reactor economy. This concept offers the potential for°
. Breeding in a thermal spectrum reactor; o
. Efficient use of thorium as a fertile material
!: Elimination of fuel fabrication and spent fuel ehipping,

. High thermal efficiencies.

Notwithstanding these attractive features, this assessment has

reconfirmed the existence of major technological and engineering

problems affecting feasibility of the concept as a reliable and ‘

Je

economic breeder for the utility industrv. The principal concerns

include uncertainties with materials, with methods of controlling

tritium, and with the design of components and systems along with

their special handling inspection snd maintenance equipment._ Many

of these problems are compounded by the ‘use of a fluid fuel in which

. fission products and delayed neutrons are distributed throughout the v

primary reactor snd reprocessing systems.

The resolution of the problems of the MSBR will require the conduct 7'

: of an intensive research and development program._ Included among ef,s:.

the major efforts that would have to be a°°°mplished are-'"r"'

-f:;. Proof testing °f an 1nteflrated reprocessing system,'T:J -

T e Development'of,a_suitable,containment_material,

 

 
 

. Development of a Satisfactory.nethod for the control and
) retention of tritium--fivfi | ) o
,A:Attainment of a thorough understandinp of the behavior of fission
”'products in a molten salt system, o ‘ |
. Development of long life moderator.graphite,vsuitable for.
breeder application' |
. Conceptual definition of the engineering features of themany
components and system3°
. 'Development of adequate methods nnd equipment for remote

inspection, handling,}and{naintennncerofrtherplant.;“ -

The major problems ascociated with the MSBR are rather difficult in nature
and many are unique to thia concept.fi Continuing support of the research

~ and development effort will be required to obtain satiafactorv solutions
to the problems. When significant evidence is available that demonstrates.
'realistic solutions are practical a further assessment could then be made
as to the advisability of advanciny into the detailed design and | |
engineering phase of the development process including that of industrial
involvement. Proceeding with this next step would also be contingent |
upon obtaining a firm demonstration of interest and commitment to the
concept hyhthepower industry and the utilities and reasonahle assurances
that large scalegovernment andindustrialresources can{he,madeavailahle-l
on a continuing basis to this programin-light of,other;commitnents to \
'the commercial nuclear pouer;program end higher priorityenergy' |

deuelopment efforts.

 
1.

2,

3.

&,

- .'..53-

" IX. REFERENCES |

u. s.-aecaie znefgyficommissaon; "Tfie*1967’Shpp1emen£’:d?tfié-1962‘

Report to the Preaident on Ctvilien Nuclear Power" USAEC Report,'

February 1967. \

u. S. ‘Atomic Energy Commission; "The Use of Thorium in Nuclear |

,_Power Reactors" USAEC Reporr WASH—1097 1969. ;fl
R S. Atomic Energy Commissiongl"Potentlal Nuclear Power Crowth
‘Pat:erns," USAEC Report VASH-1098, Decenber 1970.

U, S. Atonic Energy Commission, "Cost-Benefit Annlyeis of the

| jif*U. S. Breeder Reactor Program. | USAEC’ ReP°rtHWASH‘1126’ 1969'

5.

7.

QU. S. ‘Atomic Energy Commission;r"Updated (1970) Cost-Benefit Analvsis

of the U. S. Breeder Reactor Program." USAEC Report WASH-llS&

'_fiJanuary 1972.°

Edison Electric Institute.:"Report on the EEI Reector Aoaesement Panel "o

EEI Publication No. 70-30 w1970.1 |

Annual Hearinge on Reactor Development Program, U. S. Atomic Energy :

gl ssion FY 1972 Authorizing'LegisIation, Hearings ‘before the ‘Joint

: Committee on Atomic Energy. Congress of the United Srates p. 820-830

";f3u. S. Government Prinring Offlce

8. ,..,"i,f

';Nuclear A lications & Teéhnolo Volume 8 Februarv 1970.:,__'

Rdbertson,'R. D. (ed), “Conceptual Design Studv of a Single Flnid

"fi?erolten Salt Breeder Reactor," ORNL-4541, June 1971. ';';fi:;"".

 
 

10.

12.

13.

14.

-54-

Rosenthal, M. W., et al.;:"Ahvances‘in the Development of Molten-Salt
Breederfkéactors,"_AICUNF. 49lP-048, Fourth United Nations
International Conference on the Peaceful Uses of Atomic Energy,

Geneva, September 6-16, 1971.

Trinko, J. R. (ed;), "Molten-Salt Reactqf Techpology,"_Technical

Report of the Mblfen-Salt Group,.Part I, Decembgr 1971.7 |
Trinko, J. R. (ed), "Evaluation of a 1000 Mie Molten-Salt Breeder
Reactor," Techhical Repbrt of the Molten Salt Grdup;_Pa:;_II,_

November 197;.

Ebasco Services Inc., "IOOO_MWe HbltenJSalt Breeder Reactor Conceptual 5

Design Study,“ Final Report Task I, Prepared under GRNL subcontract
3560, February 1972. |

"Project for Investigation of Molten Salt Breeder Reactor," Final

Report, Phase I Study for Molten Salt Breeder Reactor Associates,

September 1970.

,,,Car&well, D. W. and Haubenreich, P. N., "Indexed Absttacts of

~ Selected References on Hblten-Salt_Réactor Technology,' ORNL-TM-3595,

16.

17.

December 1971.

Kasten, P. R., Bettis, E. S. and Robertson, R. C., “Design Studies

of 1000 HW(e) Mblten—Salt Breeder Reactors," 0RNL-3996 Angust 1966._

belten Salt Reactor Program Semiannual Reports beginning in o

_Fébruary 1&62. .

\E,j

 
 

 

A-1
Appendlx A

| Summary of principal data for MSBR power station

 

 

Average thermal-neutron flux

Peak thermal-neutron flux .
Maximum graphite damage flux (>50 keV)
Damage flux at maximum damage '

. region (approx) ,

Graphite temperature at maxmmm neutron -

flux region -

Graphltc tempezature at maximum graphltc
‘damage region :

~ Estimated useful life of graplute

Total weight of graphite in reactor . *

Maximum flow velocity of salt in core

Total fuel salt in reactor vessel

Total fuel-salt volume in pnmary system

Fxssxle-fuel inventory in reactor pnmary
system and fuel processing plant

Thorium inventory

Breeding ratio

Yield

Doubling tlme compounded oonunuously,
at 80% power factor ,

anary heat exchangers (fox each of 4 nmts) '

. Thermal capacity, each .
Tube-side condltlons ( fuel salt)
“Tube OD 7 B
- Tube iength_ (approx) '
~ Number of tubes ,
Inlet-outlet conditions
“Mass flow rate o
- Total heat transfer surface Ce Tl
Shell-s:de condmons (coolant salt)
“ShellID ,
Inlet-outlet temperatures
Mass flow rate

Overall heat transfer coefficient (approx)

2.6 X 10'* neutrons cm” 2 gec
8.3 X 10'* neutrons cm ™2 sec
- 3.5 X 10" necutrons cm™2 sec
3.3'% 10'% neutrons cm™2 sec

1284°F
- 1307T°F

4 years
669,000 Ib
. 8.5 fps

1074 ft3
1720 103

3316 1b

: 150 000 lb
.- 1,06
3.2 %/year
; 22 years

CS563MW(D

Hin.
2221t
5896
- 1300-1050°F
-23.45 X 108 Io/hr -
13' 000 ft2 L

850-1150°F
17.6 X 10% Ib/hr

850 Btu hr™? ft™2 (°F)~t

-1
-1
-1
-1

-Engmeermg units® . lnternatlonal system units?
-~ General 7 K
" Thermal capacity of reactor 2250 MW(t) 2250 MW(t)
Gross electrical generation 1035 MW(e) 1035 MW(e)
" Net electrical output - © 1000 MW(e) . 1000 MW(e)
- Net overall thermal efficiency 44.4% 44.4%
_Net plant heat rate 7690 Btu/kWhx . 2252 J/kW-sec
Structures - . : o
~ Reactor cell, diameter X height TT72X 421t 220X 12.8m
Confinement bufldmg, diameter X he:ght . 134 X 189 ft 40.8 X 57.6 m
Reactor L L '
Vessel ID 22,21t 6.77m
Vessel height at center (approx) 20 ft 6.1m
Vessel wall thickness 2 in. 5.08 cm
Vessel head thickness Jin. 7.62 cm .
~ Vessel design pressure (abs) 75 psi 5.2 X 105 N/m?
Core height . : ‘ 131t 39 m
Number of core elements _ R 1412 1412
Radial thickness of reflector , " 30in. 0.762 m
Volume fraction of salt in central core zone 013 ‘0.13 -
Volume fraction of salt in outer core zone 0.37 0.37 :
Average overall core power density - ‘ 22,2 kW/liter 22.2 kW/liter .
Peak power density in core - . 70.4 kW/liter 70.4 xW/liter

" 2,6 X 10'* neutrons cm -2 sec

8.3 X 10'* neutrons cm ™2 scc
3.5 % 1014 ncutrons cm ™2 sec
3.3 10M neutrom cm 2 see

969°K
982°K

4 years

' 304000kg .- .

2.6 mlscc"

" 30.4 m’:

48.7md
1504kg

68 lOOkg

1.06 = .
3.2 Flyear

. 556.3W(t) oL

"0953cm |

68m.

5896

978-839°K - '
2955 kgfsec: - o

1208 m?

1.73m-

727-894°K
2218 kg/sec
4320 Wm™2 ¢°K)"!

-

-1
‘ L
it |
=1

 

 

 

 
 

A-2

-_Appéhdix A (continued)

 

 

Apparent overall heat transfer coeffic:ent
range -

490-530 Btu ! £t72 CF)?

* Engineering units? " International system _units®?
Primary pumps (for each of 4 units)
Pump capacity, nominal 16,000 gpm 1.01 m3/sec
Rated head 150 ft 457 m
Speed 890 rpm 93.2 radians/sec -
Specific speed 2625 rpm(gpm)®” "'I(ft)° 75 5.321 radtanslscc(mslsec)o sI(m)" 75
Impeller input power 2350 hp i 1752 kW
Design temperature 1300°F 978°K
Secondary pumps (for each of 4 units) ’
Pump capacity, nominal ' 20,000 gpm 1.262 m3/sec
Rated head 300 ft 914m )
Speed, principal 1190 rpm 124.6 radians/sec
Specific speed 2330 rpm(gpm)° sI(ft)" 78 4.73 radians/sec(m®/sec)®*$ J(m)®* 75
Impeller input power 3100 hp 2310 kw .
Design temperature 1300°F 978°K
Fuelsalt drain tank (1 unit) :
Outside diameter o 14 ft 427m
"Overall height 22 ft 6.71 m
Storage capacity 2500 ft3 70.8 m3? w
Design pressure 55 psi 3.79 x 108 Nlm
Number of coolant U-tubes 1500 1500 '
Size of tubes, OD a in. - 1.91 cm
Number of separate coolant circuits 40 40 . .
Coolant fluid 7LiF-BeF, .TLiF-BeF,
Under normal steady-statc condmons ' S : '
Maximum heat load ' 18 MW(t) 18 MW(1t)
Coolant circulation rate 830 gpm 0.0524 m?/sec
Coolant temperatures, infout 900~1050°F 755-839°K
Maximum tank wall temperature ~1260°F ~955°K
Maximum transient heat load 53 MW(t) .53 MW(t)
Fuel-salt storage tank (1 unit) o
Storage capacity 2500 ft? - 70.8 m?
- Heat-removal capacity 1 MW(t) I MWE) -
Coolant fluid } Boiling water Boiling water
Coolant-salt storage tanks (4 units) |
Total volume of coolant salt in systems 8400 i3 - 2379 m
Storage capacity of each tank 2100 fe3 59,5 m3
Heat-removal capacity, first tank in series 400 kW 400 kW
Steam generators (for each of 16 'units')’
Thermal capacity 120.7 MW(1) 120.7 MW(t) -
Tube-side condmons (steam at 3600 -3800 '
ps) '
Tube OD 1 in. 1.27 cm
Tube-sheet-to-tube-sheet length (approx) 76.4 ft 23.3m - -
Number of tubes ' 393 e 393 e e
Inlet-outiet temperatures 700-1000°F 644-811°K ..
Mass flow rate 633,000 1b/hr 79.76 kg/sec . :
Total heat transfer surface 3929 ft2 - 365m? . - o -
Shell-side conditions (coolant salt) - i -
Shell 1D : ‘1.5ft , 0.457m L
Inlet-outlet temperatures 1150-850° F 894-727°K
Mass flow rate 3.82 x 10% lblhr

481.3 kgfsec

2780-300S Wm™2 ("K)V

 
 

- -,.,iFueI—prooessing system (Chefl"‘llf-'iil 'l'l‘eafi'“‘mt o
 Plant)

o A-3

~_Appendix A (continued)

 

Engineering units? - International system units®

 

Stea'rnt.eheaters‘(foréachoffluhifs)- L T _ L
Thermal capacity . 36.6 MW(D) ©36.6 MW(D) - -

- Tube-side conditions (steam at 55_0 ps:) : T T
“TubeOD - . o Hin - 1.9 cm
- Tube length C .. 3031t o 924 m

"Number of tubes = = - o 400 - : o ST 400 - -

Inlet-outlet temperatures =~~~ 650-1000°F - . . 616-811°K :

Mass flow rate : . 641,000 ib/hr a 80.77 kg/scc

Total heat transfer surface ' 2381 £ - 2212m?
Shell-side conditions (coolant salt) S -

ShellID ST “20.26n . | 0.54 m

Inlet-outlet temperatures o e 1150-850°F 894-727K .

- Mass flow rate ' o 1.16 )(_106 Ib/hr - 146.2kgfsec. - . -
Overall heat transfer coefficient -~ - - 298Btubet £tV °P) 7! 1690 W m™2 (°K)™?

rTu_rbi_n'e-ge'nexjatogblant (sec"fGeanfal" above) I 7 o ' :

Number of turbinegenerator units =~~~ ° S S : 1

Turbine throttle conditions -~~~ .~ 3500 psia, 1000°F . 24.1X 10%N/m?, 811°K
Turbine throttle mass flow rate - 7.15X%10% Ib/hr 900.9 kgfsec -

- Reheat steam to IP turbine_ . - [ 540psia, 1000°F . 3.72x 108 Nlm’ 8II°K -
Condensing pressure (abs) . . . - LSinHg - , 5,078 N/m? .
Boiler feed pump work - - : : 19,700hp - . o ' 14,690 kW - o

.~ (steam-turbine-driven), each of2umts S : L
Booster feed pump work (motor-driven), - 6200 hp - ' - 4620 kW
~ each of 2 units ‘ ' : . ' ’

Fuel-salt inventory, pnmary system
Reactor - , : ' - , B e s
Core zonel - - _ 290 13 o g2m? .
. Core zone Il o ' - o3g2md o - 10.8m3 .
Plenums, inlets, outlets - S - 218 13 : , o 62m
- 2-in. annulus o espd 3.8 m3
‘Reflectors - - o0, S 49f 1.4 m?
'anaryhcatexchangcxs e s ,«
" Tubes LR s nenE o269 76m® .

- Inlets, outlets 2718 _ o . 0.8m3
Pumpbowls Come e e S 1ssE T L o 52md
Pnpmz,mcludmgdmmlme L L 1es e ‘ 4.1m?
Off-gas bypass loop. .~ . - T . 10 0 Cen e e 7 0.3m®

: Tank heels and m:scellaneous o e S C03m?

‘Total enriched salt in primary system -~ 172066 ... 48 Tmd -

 

 

 

_Inventory ofbancn salt (LlfBng'-ThFfl 480 ft3 . l36m

inplant © Db s o eI e
Processingrate .~ T L gm0 63X 10" 3Isct.
“Cycle time for salt mventory SRR - 10days - -~ . o 10 days

Heat generatxonmsalt to prooessmg plant 56 kWIft3 o 7 S _1980 kWImf T

Deflgn pmpemesof fucl salt L 7 o . - :
~Components ~ :°: % .ot L 7L1F-Bng-ThF4-UF4 T ',7LIF-Bng-ThF4-UF4 :
 Composition  Tono L sl T T, 7-16-12-0 3 mo!e % n 7-16-124) 3 mole%
- - 'Molecular weight (approx) . o BT T T 64 ; R

" Melting temperature(approx) S : 930°_F . -:' o T ST 772"1( , S
7 Vapor pressure at 1150°F (894.3°K) -~ -~ ' <0.1lmmHg ... <133N/m*

- Density:® p (g/em®) = 3.752 - 6.68 X 104 P T T T e ' L

€O (0/ft3) = 235.0 - oozsrm"l-") I -

At 1300°F (978°K) * S 204.9 m/fe o 32839 kg/m?

At 1175°F (908°K) 207.8 Ib/ft3 33304 kg/m?
At 1050°F (839°K) S 210.7 b/ 3376.9 kg/m®

 
 

A-4

Appendix A (cont fnued) ' .

 

Viscosity:2 u (centipoises) = 0,109
exp {4090/T (°K)]; u (Ib ft=1 hr™Y)

=10.2637 exp [7362/T (°R)]
At 1300°F (978°K)
At 1175°F (908°K)
At 1050°F (839°K)

~ Heat capacity® (specific heat, ) -

Thermal conductivity &
At 1300°F (978°K)
At 1175°F (908°K)
At 1050°F (839°K)

Design properties of coolant salt
Components
Composition
Molecular weight (approx)
Melting temperature (approx)
Vapor pressure:$ log P (mm Hg)
=9.024 - 5920/T (°K)

At 850°F (727°K)
" At 1150°F (894°K)

Density:€ p (gfcm®) = 2.252 - 7. ll X 107%
(°C); p (Ib/ft3) = 141.4 — 0.0247¢ (°F)

At 1150°F (894°K)
At 1000°F (811°K)
At 850"!—‘ {(127°K)

Viscosity:4 u(centnpowes) 0.0877
exp [2240/T (°K)];u (Ib, ft™! hr

= 0.2121 exp [4032/T (°R)}
At 1150°F (894°K)

At 1000°F (811°K)

At 850°F (T27°K)

. Heat capacityh (specific heat, ¢)

Thermal conductivity (k)¢ p
At 1150°F (894°K)

At 1000°F (811°K)

At 850°F (7127°K)

Design properties of graphitef
Density, at 70°F (294.3°K)
Bending strength
Modulus of elasticity coefficient
Poisson’s ratio
Thermal expansion coefficient

Therma! conductivity at 1200° F,

unitradiated (approx)
Electrical resistivity

Specific heat

At 600°F (588.8°K)
At 1200°F (922.0°K)

. Helium permeability at STP thh sealcd

surfaces

Engineering units?

17.31bhe! ft~1
238 b he7l fr!
3451 he! £

0.324 Btu b1 CF)™ £+ 4%

0.69 Btuhr™! CF)~! fit™!
0.71 Btuhr™! C°F)"! 171
0.69Btuhr™! CF)™! 171

‘NaBF 4-NaF
92-8 mole %
104

725°F

8 mm Hg
252 mm Hg

113.0 Ib/ft3
116.7 b/fe3
120.4 b/fe3

26 ft" et
341 ft7! he?

4.6 ft™! et

0.360 BtuIb~! (°F)"‘ t 2%

0.23 Btu nr'l R a7t
0.23Btuhe™! P! £t

0. 25 Btu hr™! ( F)" ft! -

115 b/fe?

4000-6000 psi

1.7 X 10° psi

0.27

2.3 107S°F

18 Btu he™! (°F)71 !

8.9 X 107-9.9 X 10*-2cm

0:33Btulb P!

042Btulb”l R
~ 1% 1078 em?/sec

_ International system units?

0.007 N sec m™2

0.010 N secm™2
0.015 N secm™?

1.357J g! (°K)" t 4%

1. 19Wm'1 ("K)"
1.23Wm™ CK)™!
1LI9Wm™ CK)

NaBF4-NaF .
92-8 mole %
104

A58°K

1066 N/m? S
33,580 N/m?* -

1811.1 kg/m?
1870.4 kg/m?
1929.7 kg/m3

0.0011 N sec m=2
0.0014 N sec m*?
0.0019 N secm™3

1507 T kg™ (°K)"! £ 2%

0.398 Wm™ (°K)™!
0.398 Wm™! (°K)"!
0.450 W _m" R

1843 kg/m® |
28 x 109-41'x 10 Nlm’
11.7 x 10° Nlm

027
13x10'°l°x
C312wm (°K)?

89X 1074-9.9 X 10~ 2-cm

1380 J kg™ (°K)™}
1760 J kg™ (°K)™?
1x 107® cm?/sec

 
 

A-5

Appendix A (con_tinued)_

 

 

Thermal conductivity
At 80°F (300°K)
At 1300°F (978°K)
Specific heat
At 80°F (300°K)
At 1300°F (978°K)
‘Thermal expansion
At 80°F (300°K)
At 1300°F (978°K)
Modulus of elasticity coefficient
At 80°F (300°K)

At 1300°F (978°K)
Tensile strength (approx)
At 80°F (300°K) ~

~ At 1300°F (978°K)

Maximum allowable design stress -

At B0°F (300°K)
At 1300°F (978°K)

Melting terhperature

C6.0Btuhet CHL !
'12.6 Btuhr™! (°F)71 ¢!

0.098 Btu Ib~! (°F)~1
0.136 Btulb™! (°H)?

S TX 107°8°F
9.5 X 1079/°F

31 x 10° psi
25 X 10° psi

115,000 psi
75,000 psi

25,000 psi
3500 psi
2500°F -

Engineering units® International system units?
- Design properties of Hastelloy Nk
 Density : _
At 80°F (300°K) - 557 1b/ft3 8927 kg/m>
At 1300°F (978°K) - 541 Ibfft3 8671 kg/m3

104Wm-! (°K)!

218Wml °K)!?

- 4107 kg™ )Y

569 T kg™ (°K) !

32X 1075/°K
5.3X% 10"_‘I°|<

214 X 10° N/m?
172 X 10° N/fm?

793 X 10% N/m?
517 X 10° N/m?

172 X 10° N/m?
24 X 10° N/m?

- 1644°K

 

"Enghsh engineering units as used in MSR literature.
Meter-kilogram-second system. Table closely follows International System (SI). See Appendlx C for conversion factors from

engineering to SI units.