NEW DEVELOPMENTS IN MATERIALSFOR M O LTE N-SALT R EACTO RS

I. _8. ryCoy, R. L. BEATTY, W. H. COOK, R. E. GEHLBACHc. _R. KENNEDY, J. W. KOGER, A. p. LITI\,iAN, C. E. SESSIONSand J. R. WEIR Metats and Cerarnics Diaision,Oak RMge Natioral Laboratory Oak Ridge, Tennessee JT8J0

Received August 4, 1969Revised October 10, 1969

KEYIYORDS: rodiotion effects,mollen-sqtt reoctors, breede'rreqctors, PoW er reoctors,reoclor core, fused sorts,coo.fonfs, ch ro m i u m olloys',molybdenum olloys, nickel'ii-loys, corros ion, brittleness,thermol neutrons, sodium fl;;:rides, sodium borides, mixing,grophite, fost neutrons, sfo-!!l,rr,,,expon-s i on, poiosity,H o stel I oy,- embri ttl e'me nl, m i x_tures, surfoces

operatins etcperience with the Molten-satt Re-actor Experiment (A[SRE) has d,emonstrated theetccellent compatibility of the gTaphite -Hastelloy -N-fluoride sqlt sys tem at 6s0"c. seueral im-prouements in mq,terials are needed for amolten-salt breeder rea,ctor with a basic plant tifeof 30 years ; specifi,cally: Hastehoy -N with im_proaed resistance to embri.tttement by thermalneuttrons; graphite u)ith better dimenilo,nt sta-bility in a fast nqttron ftuu; yaphite that is sealedto obtain a surface permeabi.lity of {10-s cmz/sec;ard a secondary coolant that is inexpensiue andhas a melting point of -400"c. A brief descriptionis giuen of the materials worh in progTess tosatisfy each of these requi.rements.

INTRODUCTION

our present concept of a molten-salt breederreactorr utilizes graphite as moderator and re-flector, Hastelloy -N for the containment vesseland other metallic parts of the system, and aIiquid fluoride salt containing LiF, B;Fz, u-Fo, andThF4 as the fertile-fissite medium. The fertile-fissile salt wilt leave the reactor vessel at atemperature of -?00"c and energr wiII be trans-ferred to a coolant salt which in turn is used toproduce supercritical steam.

E>rperience with the Molten-satt Reactor Ex_periment (MSRE ) has demonstrated the basiccompatibitity of the graphite-Hastelloy -Na -fluo-ride salt (LiF-BeF2 -zrF4 -UFn) system at 6b0"C.Howev€r r a breeder reactor will impose morestrdngent material requirements; tta*ely: the de-sign life of the basic plant of a breeder is 30years at a ma:rimum operating temperature of

700"c; the power density witl be higher in abreeder and will require the core graphite tosustain higher damaging neutron flux and fluenceland neutron economy is of utmost importance inthe breeder and the retention of fission products(particularly tt*u) by the core graphite must beminimiz ed. Each of these factors requires aspecific improvement in the behavior of materials.

Experience has shown that the mechanicalproperties of Hastelloy-N deteriorate as a resultof thermal-neutron exposure and a method mustbe found of improving the mechanical propertiesof this material to ensure the desired 30-yearplant life.

similarlyr graphite is damaged by irradiation.Although the core graphite can be replac€d, theallowable fast neutron fluence for the graphite hasan important influence on the economics ofmolten-salt breeder reactors. Thus , a programhas been undertaken to learn more about irradi-ation damage in graphite and to develop graphiteswith improved resistance to damage.

A big factor in neutron economy is reducing thequantity of ttsxe that resides in the core. Thisgas can be removed by continuously spargrng thesystem with helium bubbles, but the transfer bythis method probably witl not be rapid enough toprevent excessive quantities of t35xe from beingabsorbed by the graphite. This can be preventedby reducing the surface diffusivity to <10-t

"^r/sec, and we feel that this is best accomplished bycarbon impregnation by internal decomposition ofa hydrocarbon.

a Hastelloy-N is the trade name of ucc for a nickel-base alloy containing L6vo Mo, Tls cr, }vo Fe, 0.05!s c.This alloy was originarty developed at ORNL specifi-cally for use in mclten-salt systems. It has beln ap-proved by the ASME for use in pressure vessels undercode cases 131b and 194b.

APPLICATIONS & TECHNOLOGY VOL. B FEBRUARY 1970t56 NUCLEAR

McCoy et al. MATERIALS

A new Secondary coolant is also needed thatwill allow us greater latitude in operating temper-ature. Sodium fluoroborate has reasonable -phys-ical properties for this application, a n d the

"o*patibility of Hastelloy-N with this salt is being

evaluated.Our work in each of these areas will be de-

scribed in some detail.

EXPERIENCE WITH THE MSRE

Other papers in this series have elaborated on

the information gained from the MSRE regardingoperating experience, physics, chemistry, and

fission-product behavior. Additionally, valuableinformation has been gained about the materialsinvolved.2 -a

bTrade name of Union Carbideneedle-coke graPhite used in the

There are surveillance facilities exposed_to the Fie. 1. Graphite and Hastelloy-Nsurveillanceassemtlysalt in the core of the reactor and outside the "

""*bned from the core of the MSRE after

reactor vessel, where the environment is nitrogen 72 4oo MWh of operation' Exposed to flowing

pl:us -27oOz. Hastelloy-N tensile rods and sam- saltfor 15 300 h at 650oC'

ples of the gSade CGB graphiteb used in the coreof the MSRE are exposed in the core facility.The components are assembled so that portions construction of the reactor. Hastelloy-N is nickelcan be removed in ahot cell, new samples addecl, Uasea and contains about l6Vo Mor T%CrrandSVoand the assembly returned to the reaclo^rt^ffm; Fe. Under normal operating conditions, the fuelples were removed after 1100,4400, and9000_h ot salt cannot oxidize (form fluorides) any of thesefull-power (-8 MW) operation at 650'C- As shown

"i"*""ts except Cr. Since Cr is present in veryin Fig. 1, the physical condition of the qfaphite

"-rfi concentrations in the al'oy, the corrosion isand metal samples was excellent; identification timitea by the diffusion of Cr to the metal surface.numbers and machining marks were clearly visi- Corrosion can be reduced even further by con-ble. The peak fast fluence received UV lne qTl_ trolling the oxidation state of the salt, thusite has ueen 4.8 x 1020 n/cm' ()50 kev) ?9_]l" ;;;c the rate of the corrosion reaction at thedimensional changes axe (:0.1Vo. Pieces- of gSaph- *"Jl-fl".ft interface. The oxidation state of theite from the MSRE have been sectioned- and most salt in the MSRE is controlled by the addition ofof the fission products were found to be located on iervttium metal.the surface and within 10 mils below tl" :1"t3::: Hastelloy-N samples were removed from theHowever, a few of the fission products..nave surveillancl facitity outside the reactor vesselgaseous precursors and penetrated the q?phl_le to after 4400 and g000 h of full-power operation.greater depths. The microstructure ol'.th" l"?. This environment is oxidizing, and an oxide filmtelloy-N near the surface was modified to-a depth -Z *if" thick was formed on the surface after theof -1 mil, but a similar modification was,:::Tll longer e4posure. There was no evidence of ni-samples orposed to static nonfissioninc :il:.111.i1 t"iiing, rna the mechanicat properties of theseequivalent time. The near-surface moctificatiot

"*oll"" were not affected adversely b5r the pres-has not been positively identified, but its presenc-e .n""-ot the thin oxide film.is likely of no consequence. The

"""Y ,::it] Thus, extrlerience with the MSRE has proved inchanges in the amounts of chromium and iron rn service thJ excetlent compatibility of the Hastel-the fuel salt also indicate very low

"_o,1I:_"]_:" ioy-N-graphite-fluoride salt system.rates and support our metallographic observa-ttoH;"

observed low corrosion rate of Hastelloy-N

.T.*i#sx?'m:"":l """"T;ffi"";]JI""T}#H: il'.'x[.99$'$'3lofid%1FR^Bi^T.*i.J^T,fdl'

Since the MSRE was constructed, Hastelloy-N,as well as most other iron- and nickel-base

corporation for the alloys, was found to be subject to a type of high-

MSRE. temperature irradiation damage that reduces the

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O 157

McCoy et al. MATERIALS

stress rupture life and the fraction strain.s -1oThis effect is characterized in Figs" 2 and 3 for atest temperature of Gb0"c. The rupture 1ives forirradiated and unirradiated materials differ mostat high stress levels and are about the same forbelow 20 000 psi. The property change of mostconcern in reactor design and operation is the areduction in fracture strain. The postirradiation 3 cofracture strain is shown in Fig. 3 as a function of 3strain rate; the scatter band is based upon test 3results for three different heats of metal. The H

to

20

to

otoo

-0. lVo/h, with rapidly increasing fracture strainas the strain rate is increas€d, and slowly in- Fig' 2'

creasing fracture strain as the strain rate isdecreased" Thus, under normal operating con-ditions for a reactor where the stress levels (andthe strain rates) are low, the rupture life will notbe aff,ected significanily (Fig. z), but the fracturestrain will be only 2 to 4Vo (Fig. 3)" Howev€r,transient conditions that would impose higherstresses or'require that the material absorbthermally induced strains could cause failure ofthe material. Therefore, a material is desiredthat has improved properties in the irradiatedcondition and a program with this as its goal hasbeen embarked upon.c

The changes in high-temperature properties ofiron- and nickel-base alloys during irradiation inthermal reactors have been shown rather con-clusively to be related to the thermal fluence and d 9

more specifically related to the quantity of helium =

Iproduced in the metal from the thermal toB(n,o)tl,i E

transmutation.ll-14 The mechanical properties 6 7

a.re only affected under test conditions that pro- Educe intergranular fracturing of the material. B

6

under these conditions both cr eep and tensile f scurves for irradiated and unirradiated materials Lare identical up to some strain where the irradi-ated material fractures and the unirradiated ma-terial continues to deform. Thus , the main in-fluence of irradiation is to enhance intergranularfracture.

70

60

50

plot includes results from both tensile and creeptests. In tensile tests the strain rate is acontrolled parameter and the test results areplotted directly" rn creep tests the stress iscontrolled and the strain measured as a functionof time. The minimum strain rate was used inconstructing Fig. 3. The data are characteri zed,by a curve with a minumum at a strain rate of

cAlthough the fast-neutron flux wiII be quite high in thecore of our proprrsed breeder, the neutrons reachingthe Hastelloy-N vessel wiII be reduced in energy by thegraphite present. Thus, the fast fluence seen bv thevessel during 30 years will be < 1 x L02rn/. r,

"rid *e

do not feel that fast-neutron displacement damage is ofconcern. Experiments will be run to confirm this point.

Fig. 3. Fracture strain of Hastelloy-N at 6b0rc afterirradiation to a thermal fluence of - b x 10zon/ crn?-

tol lozRUPTURE TIME (h)

Creep-rupture properties650oC after irradiation to a-5x 1020n/cmz.

to3 qoa

of Hastelloy-N atthermal fluence of

14

,13

t2

q,

ro

{o- { {oo to|STRAIN RATE (% h)

o 27.3o 22.2t22.2o 21.?

,oz {o3

4

3

2

I

orFl,O-2

I Lll+ l,'.

I : :::?I!TYPI(

1".

)AL

I

I

l.

rlrri['

l"llTill-'I

\-o

\# i\{ o D

I

o\

( \|ilt | |ilil

PRETEST ANNEAL-{ h

SOL]D POINTS-I RRAIOPEN POINTS-IRRAD

lilll lllilill

o,,j ,)"Jr III]T-TATED < { 50"CATED 500-650"c

l lillill

HEAT NO.o 5065a 5067o 5095

PRETEST ANNEAL-{ h AT 'I 177"CSOLID POINTS- IRRADIATED < I5O"COPEN POINTS-IRRADIATED 5OO-650"C

CREEP

TESTSTEN SI LE

TESTS

r58 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. B FEBRUARY 1970

A logical solution to this problem would be toremove the boron from the alloy. However, boronis present as an impurity in most refractoriesused for melting, and the lowest boron concen-trations obtainable by commercial melting prac -tice are in the range of 1to 5 ppm. The lowhelium levels that have caused the creep-ruptureproperties to deteriorate in Hastelloy-N make thisapproach very unattractive. For example, in-reactor tube burst tests at 760'C showed that therupture life was reduced by an order of magnitudeand that the fracture strain was only a few tenthsof a per cent when the computed helium levelswere in the parts -per -billion range.to Thus, W€

have concluded that the properties of Hastelloy-Ncannot be improved solely by reducing the boronlevel.

Another very important observation has beenthat the properties are altered by irradiation atelevated temperatures only when the test temper -ature is high enough for grain boundary deforma-tion to occur (above about half the absolutemelting temperature for many materials). Thus,the role of helium must be to alter the propertiesof the grain boundaries so that they fracture moreeasily.

The size of boron lies intermediate between thesizes of small atoms such as carbon that occupyinterstitial lattice positions and the larger metalatoms, such as nickel and iron, that occupy thenormal lattice positions. For this reason boronconcentrates in the grain boundary regions wherethe atomic di sorder provides holes large enough

to accommodate the boron atoms. Thus, thetransmuted helium will be generated near thegrain boundaries wher e it will have its mostdevastating effects. We reasoned that the additionof an element that formed stable borides wouldresult in the boron being concentrated in discreteprecipitates rather than being distributed uni -formly along the grain boundaries. The trans -muted helium would likely remain associated withthe precipitate and be less detrimental. Addition-ally, certain precipitate morphologies and alloyingelements are beneficial in improving the resis-tance of a}Ioys to intergf anular fracture.

Following this reasoning, small additions of Ti,Hf, and Zr have been made to Hastelloy-N and thepostirradiation properties were improved mark-edly.14r15 The titanium -modified alloy was chosenfor development as a structur al material for a

molten-salt breeder experiment (MSBE ). A fur -ther modification made in the composition was

reducing the molybdenum content from 16 to LZVo.

This change was prompted by the observation thatthe additional molybdenum was used in forminglarge c arbide particles that made it difficult tocontrol the grain Size. The vacuum-melting prac-

McCoy et al. MATERIALS

tice was adopted to reduce the concentrations of

other residual elements thought to be deleterious.The str ess -ruptur e properties at 650'C of

several heats of the titanium-modified alloy (0.|Vo

Ti) are summarized in Fig. 4. The properties inthe absence of irradiation are improved overthose of standard Hastelloy-N and the rupture lifeof the modified alloy is not reduced more than

-lTVo by irradiation at 650'C to a thermal fluenceof 5 x 1020 n/crn'. The postirradiation fracturestrains of the titanium-modified alloy are alsoimproved over those of the standard alloy (Fig. 5).

The modified alloy has a very well-defined duc-tility minimum as a function of strain rate, blt theminimum strain is -SVo compared with 0.5V0 forthe standard alloy.

Electron microscopy has shown that thetitanium-modified alloy forms very fine-grainboundary and matrix precipitates of the MC typewhen annealed at 650'C. These precipitates areonly a few tenths of a micron in size and those inthe grain boundaries are Spaced at -2-U intervals.They have a faee-centered cubic crystal structurewith a lattice parameter of -4.24 ^4, and are likelycomplexes involving Mo, Cr, Ti, C , Nr and B.

These complex compounds also form precipitatesin the matrix. This microstructure should lead totrapping of some of the helium as proposedea-rlier and should also inhibit fracture along thegrain boundaries. Howev€f r further studies haveshown that the precipitates which form during longexposure at ?60'C are relatively coarse MozC

carbides and that the postirradiation properties of

I tO IOO

RUPTURE LIFE, h

Fig. 4. Creep-rupture properties of several heats ofmodified Hastelloy-N at 650Pc, samples irra-diated to a thermal fluence of -5 X 10'" n/ crn-.

CT'coo9,;al^lEFU'

HEAT NOS.

o - 66-548 (O.5% Ti)

o -21545 (O.57" Ti)

PRETEST ANNEAL.I h AT II77"CIRRAOIATED AT 5OO-650"C

rr,rt I ll>STD. IRRADIATEDJ I

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 159

dz8aEirTLJElF6(J

ELLF

3

McCoy et al. MATERIALS

HEAT IRRADIATED TESTTEMPERATURE "C TEMPERATURE "Co 6252 650 650

o 6252 760 760o 59{ I 650 650. 594 { 760 760A ?1545 5OO- 650 6 50A 21545 5OO - 6 50 7 60o 66-548 5OO- 650 650r 66-548 5OO- 650 760

ANNEAL { h AT 1177OC BEFORE TESTING

{8.'l 4e a{6.45 nrqa

{o-l 4ooSTRAIN RATE (%/hl

Fig. 5, variation of fracture strain with rate for sev-eral Hastelloy-N type alloys. samples irra-diated to a fluence of - b x 1do n/ cmz prior tote sting.

material irradiated at 760'C are very poor. Sincesome designs require the vessel of the MSBR tooperate at -700"C, this difference in precipitatemorphology and subsequent deterioration of prop-erties was of concern.

Further work has shown that the desired Mc -type carbide can be stabiltzed at higher servicetemperatures. This carbide is favored by in-creasing amounts of Ti, Hf, and Nb and decreasingconcentrations of Si. Several alloys have been

prepared that retain good postirradiation prop-erties after irradiation at 260"C. Several of thealloy additions that appear satisfactory are l.ZVoTi, lVoHt, lVo Hf plus l% Ti, 0.iTo Ti pi,rc ZVo Nb,and 2Vo Hf. The results of tests on these alloysgtve encouragement that a commercial alloy canbe developed that has properties at least as goodas those shown in Figs. 4 and b.

IRRADIATION DAi,IAGE IN GRAPHITE

Neutron irradiation alter s the physic aI pr op -erties of gr^aphite, but the dimensional changesthat occurlorl? ar e of maj or conc ern. Thesedimensional changes are illustrated in Fig. 6where the data of Henson et al.tt a"e presented forisotropic graphite. with increasing fluence thegraphite first contracts and then begins to expandat a very high rate. several potential problemsarise as a result of these dimensional changes.First, the initial contraction will change thevolume occupied by fuel salt and change thenuclear characteristics of the reactor. Thesedimensional changes seem small enough for mostisotropic graphites that the nuclear effects may beaccommodated by design. A second problem isstress generation due to flux gradients across apiece g{ graphite. Graphite creeps under irra-diationtn and this creep is large enough to reducethe stress intensities to quite aeeeptable values.The third and most serious problem is that therapid growth rate represents a rapid decrease indensity with potential crack and void formation.At some fluence this will cause the mechanicalproperties to deteriorate and the permeability tosalt and fission products to increase. We feel thatproperties will be aceeptable, at least until thematerial returns to its original volume, and havedefined this fluence as the lifetime. A fourth

o

Fig. 6.

{o zo 30 40 50 ( xro2t )

NEUTRoN FLUENCE (n /cmzl ( 6 > 50 keV)

Volume change in isotropic graphite Dounreayfast reactor irradiations.

85lrJ(9z5o9EFt!=3-so

-to

/'50-600" c

\ /

/

c-440" c\ >z /+o

160 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

problem is that the dimensional changes aredependent on temperature and the curve in Fig. 6

is shifted up and to the left for increasingtemperature. Thus, str esses develop in a parthaving a temperature gradient since segments ofthe part are seeking different dimensions. Again,this stress is relieved by the irradiation-inducedcreep in graphite geometries of interest. There-fore, we consider the onset of rapid growth to bethe primary problem and the initial dimensionalchanges of secondary importance.

Graphite temperatures between 550 and 750"Care anticipated in MSBR's, and operation with afast flux (>50 keV) as high as 1 x 101s n/@m'sec)is desired. Data in Fig. 6 indicate that this fluxwilt cause this particular graphite to expandrapidly after a fluence of -3 x LO22 n/em' i.s

reached (-1 year of operation). The flux can bereduced by decreasing the power density, but thisusually increases the fuel inventory and doublingtime. Hence, it is quite desirable that graphite beused with better resistance to irradiation damagethan the graphite shown in Fig. 6. The dataavailable on current reactor graphites irradiatedto high fluences were examined and the resultsdescribed a fairly consistent picture. The flu-ences required for graphite to reach its minimumvolume were strongly temperature dependent (de-creased with increasing temperature), but werenot appreciably different for any of the graphitesstudied to date. Although this observation is

McCoy et al. MATERIALS

discouragtng, current experiments show that bet-ter graphites already exist and that others canprobably be developed with only small changes inpresent materials and processing. Let us lookbriefly at a simple description of the origin of thedimensional changes and then return to our spe-cific observations.

Graphlte, after being well graphitized at tem-peratures above 2000"C, has a hexagonal close-packed crystal structure consisting of elose-packedlayefs, (basal planes) of carbon atoms with verystrong covalent bonds within the basal planes (adirection) and very weak Van der Waals' forcesbetween atoms in adjacent basal planes (c direc-tion). This anistropy in atomic density and bondstrength is reflected by very anisotropic prop-erties.

The changes that take place in a single crystalof graphite during irradiation are shown schemat-icalty in Fig. 'l . A neutron having an energ-y above

- 0.18 eV can displace a. carbon atom from a

close-packed basal plane with a reasonable prob-ability of creating a vacancy in a basal plane andan interstitial carbon atom between the basalplanes. Repetition of this process and diffusion atelevated temperatures can result in the formationof defect clustef s, specifically partial planes ofatoms between the basal planes and vacancyclusters within the original planes. This leads toan expansion perpendicular to the basal planes (lc

direction) and a contraction within the layer

PERFECT SINGLE CRYSTAL

DEFECTPLANES

BASALPLANES

:\\---AVvo

ALLo

(a)

POLY CRYS TAL

OBSERVED

oat-t-"

PRED ICTED

Fig. 7 , Graphite dimensional changes due to irradiation.

AVvo

(e)

DE NSI F ICATION

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 t6l

McCoy et aI. MATERIALS

planes (a direction) as shown in Fig. 7b. Thisnew configuration leads to a slight increase involume (Fig. 7c).

P olycrystalline graphites are not initially oftheoretical density. The voids present in thematerial are a result of shrinkage of the binderduring graphitization and from separation or frac -ture of layer planes during cooling from thegraphitization temperature. Initially, during irra-diation the porosity within the material tends to befilled as the crystallites expand in the c direction.This produc es a densification illustrated by thebottom curve in Fig. 7d. As the porosity fills, theshrinkage saturates and the dimensional behaviorbegins to be dominated by the volume expansiondue to the growth of the crystallites in thel cdirection. Thus, a minimum volume is obtainedas shown in Fig. 7 e. During the subsequentexpansion, the material either remains internallycontiguous, in which case the volume change rateof the polycrystalline material should be similarto the small rate of expansion exhibited by thecrystallites themselves, or fractur es internallydue to the stresses generated between the crys-tallites of differing orientation (causing a higherrate of growth to occur ). Observations to dateindicate that most graphites increase in volume ata faster rate at high fluences than expected if thematerial r emained internally contiguous.

One further consideration helps to explain whythe unpredicted rapid growth takes place. A

schematic representation of several coke par -ticles and binder after graphitization is shown inFig. 8.'o Each coke particle consists of severalcrystals with a very high degree of preferredorientation. Although the coke particles are ar -ranged randomly in a large piece of graphite,there are still interfaces between particles ofwidely different orientations. As each particlechanges dimensions, these interfaces must bestrong and able to shear large amounts withoutfracturing. The observation that graphites under -go large dimensional changes at high fluencesindicates that these interfaces or boundaries arefracturing. As indicated by the sketch in Fig. 8,these boundaries are made up largely of thegraphitiz ed binder materials. Thus, the prop -erties of these boundaries are influenc ed largetyby the nature of the binder material and its inter-action with the coke particles.

we are making graphites with known filler andbinder materials, but our work in this area hasnot progressed very far. This work also includesa study of the properties of several commercialgraphites that may be potentially useful for MSBRapplications and others that should grve somebasic information about irradiation damage ingraphite. Graphite irradiations have been done at705 * 10"c in the High Flux Isotope Reactor(HFIR) where the peak flux (>b0 keV) is 1 x1015 n/@m'sec). Thus, samples can be irradiatedto fluences of 1 x 1022 n/emz in ,-,4 months.

GRAPHITE

Pitch coke crystalfites

Micropores

Petroleum cokecrystallites

Fig. 8. Proposed aruangement of crystallites in graphitized stock.

Petroleumcoke particle

162 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

A summary of the results obtained to date isshown in Fig. 9. Several materials show a

si gnific ant deviati on fr om the ' 'typic aI' ' behaviorillustrated in Fig. 6. The POCOd graphites showexcellent resistance to irradiation with very smalldimensional changes out to fluences of 2.5 x10" n/crn'. Thus, these data support the suppo-sition that the typical behavior of graphite canbe improved markedly, but test results have notbeen extended to fluences high enough to deter -mine the exact magnitude of this improvement.

SEALING GRAPHITE

Entry of fuel salt into graphite can be pre-vented by keeping the entrance diameter of theaccessible porosity smaller than 1 ;tr. Althoughthis does require some extra care during pro-cessing it can be aceomplished routinely on largeshapes. In fact, the grade CGB graphite obtained5 years ago for the MSRE satisfies this require-ment." Howev€f , the graphite structure must bemuch more restrictive to prevent gaseous fissionproducts, particularly t3uxe, from diffusing

- into

the graphite. We presently propose to strip ttsxe

from the fuel salt by purglng with helium. Heliumbubbles will be injected and later removed in agas -Iiquid s e p ar at o r. The efficiency of thispur gtng depends very heavily on the siz e ofbubbles that can be injected and circulated and themass transfer of 13uxe from the salt to the heliumbubbles. Both of these factors are presentlyuncertain, and we must anticipate that largequantities of ltuxe will be available to the graphitesurfaces and that excessive (>0.5V0) retention ofttsxe wiII result if this gas can enter the graphiteSurface at a high rate. Present calculations show

that the acc essibility of tsxe to the graphitesurfaces will be impeded by a laminar salt filmand that the graphite offers an additional resis-tance to gas flow only when its diffusivity to lsxe

isThe best grades of commercial graphites pres-

ently available have b u Ik diffusivities in therange of 10-t to 10-4 cm'/s€c, and it is unreason-able to expect that techniques can be developed formaking massive shapes with such a restrictivestructure. The techniques used for reducing theporosity of graphite involve multiple impreg-nations of the material with liquid hydrocarbonsand then firing to graphitize this material. As thebulk diffusivity decreas€s, it becomes progres -sively more difficult for the gases released by thedecomposing impregnants to diffuse out of thematerial and the times required to reach the

dpOCO Graphite, Inc., Garland, Texas.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

McCoy et aI. MATERIALS

o5lo,|5202530FLUENCE ln /cm?) (f >5o kev) ( x to2l )

Fig. 9. Volume changes in graphite iruadiated at705oC.

graphitizing t e m p er atu r e become excessive.Thus, it is more reasonable to reduce the surfacediffusivity by a postfabrication surface-sealingprocess involving gaseous impregnation. Sincethe pyrolytic carbon, that would be deposited, ffidthe graphite substrate will change dimensionsdifferently under irradiatioh, it is imperative thatthe pyrolytic carbon be linked with the substratestructure and not deposited as a surface layer thatcan be sheared easily.

The task of sealing the graphite is illustratedby the photomicrograph in Fig. 10. The pro-cessing parameters must be adjusted so that thevoids are filted internally in preference to closingthe voids near the surface and forming a coating.This can be accomplished by using a flowingstream of hydrocarbon at low partial pressure and

temperature appropriate to maintain very lowdeposition kinetics, but this requires long pro-cessing times. We have used a different method

+>l o1fo

c

oo

FEBRUARY 1970 r63

McCoy et aI. MATERIALS

Fig. 10. Photomicrograph of the edge of graphite showing a pore that has been partially coated and then sealed overwith pyrocarbon.

to accomplish penetration which involves pulsingthe sample environment between a rich hydro-carbon environment and vacuum. The vacuumcycle removes the reaction products (primarityhydrogen) and allows more hydrocarbon gas toenter the void. specificalty, w€ have used !,3butadiene at 20 psig, deposition temperatures of800 and 1000"c, and cycle times of ,-., 1 min forthevacuum and a fraction of a second for the hydro-carbon. Butadiene was chosen because it is a gasat room temperature and because of its highcarbon yield per molecule. The temperaturerange is restricted to 800 and 1000"C becausehigher temperatures result in a surface coatingnot penetrating the pore structure and lowertemperatures yield intolerably low depositionrates. The lengths of the vacuum and pressureperiods are very important because they not onlyinfluence the processing rate, but also the depthof penetration of the impregnant. The time re-quired for the process will be important indetermining the cost.

Two commercial graphites have been used inthis work-u'ia., AXF made by POCO and ATJ-SG

made by ucc. These materials had widely dif -f erent accessible pore spectra; nearly all thepores in the AxF material were <0.8 tt indiameter while the ATJ-SG grade had pores in allsize ranges up to 17 trt. Thus, the sealing charac-teristics of the two materials were widely dif-ferent. The results of some parameter studiesare shown in Fig. 11 where the vacuum-hydro-carbon cycle times were varied at Bb0"C. Theinitial slopes are proportional to the surface areabeing coated and the slope is much steeper for theAxF graphite than for the ATJ-SG material. Thesharp break in the curves for the A>(F graphiteindicates that the pores have been fitled or closedoff and that th e sur fac e ar ea bei ng c oat ed i sreduced. The sharpness of this break attests tothe uniform pore size of the AxF graphite. Thehorizontal portions of the curves represent, essen-tially, surface coating, and the data suggest thatsome finite amount of surfac e c oating i s nec es -sary to attain the MSBR permeability specifi -cation of <10-8 cm'/sec for ttuxe at ?00"C. (Attuxe permeability at ?00'C of 4 x 10-t

"^'/sec isapproximately equal to a helium permeability at

APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970164 NUCLEAR

McCoy et aI. MATERIALS

-9otdzaIFI9ld

=JFoF

o.ro

o.o8

o.o6

o.o4

o.o2

oo.to

o.o8

o.o6

o.o4

o.o2

oo

Fig. 11.

16 20 24 28

TOTAL PROCESSING TIME (h)

Impregnation rate of graphite using 1,3 butadiene at 850oC.

25"C of 1 x 10-8 emz / sec. ) Helium permeabilitymeasurements were made and the goal was ahelium permeability of <1 x 10-t

"^'/sec (denoted

on these curves by a 33v" mark). The ATJ-SGgraphite was not sealed to the desired level underthe conditions shown in Fig. 11 and the slopechanges very gradualty due to the wide variationin the pore sizes.

The data indicate that the proc essing timecould be reduced by shortening the length of thevacuum cycle. Another interesting feature of theprocess for the AXF graphite was that the finaltotal weight of carbon deposited was increased byshortening the vacuum pulse. This indicates thatthe depth of penetration of carbon into the ma-terial was increased. Thus, shortening the vac-uum pulse accelerated the process and improvedthe product, both very desirable characteristics.

These studies are not yet extensive enough tooptimize the deposition conditions but are suffi-cient to make us optimistic about being able toreduce the diffusivity of graphite to the desiredlevel. The remaining question of prime impor-

tance is the integfity of the seal after exposure tohigh neutron fluences.

CORROSION IN FLUORIDE SALT SYSTEMS

Two decades of corrosion testin g22-3o and ex-perience with the MSRET'4 have demonstrated theexcellent compatibitity of Hastetloy-N and graphitewith fluoride salts containing Li F, BeF2, ThF4,

and UFa. The fertile-fissile salt for an MSBR willcontain these same fluorides, So only proof -testing wilt be required for the primary reactorcircuit. Howev€f r a coolant salt is needed with alower melting temperature than the LiF-BeFz saltpresently used in the MSRE; a sodium fluoro-borate salt (NaBF4-8 mole7o NaF) has been chosenas a potential coolant salt for further study. Thissalt is inexpensive (<$ 0.50/1b) and has a lowmelting point of 385"C. A significant characteris-tic of this salt is that is has an appreciableequilibrium overpressure of BFs gas (e.g., 180

mm at 600'C ).

VACUUM (sec) HYDROCARBON (sec)

a6oV2e6Ol" 30 1tz

.3Olo 15 1/z

v f.O x lO-8 c^2 /tec (HELIUM )

v . lO-lo c^2/sec (HELIUM )

MATERIAL: POCO AX F GRAPH ITE

s x io-a l'n m 'Hg ('A lR

2.5 xlO-2 nr m

MATERIAL: ATJ-SG GRAPH ITE

zsito-2mthHs (AlR)

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 165

McCoy et aI. MATERIALS

Much of our present corrosion work is con-cerned with the compatibility of Hastelloy-N withsodium fluoroborate. Some earlier thermal con-vection loop studies involving a relatively impuresalt of composition NaBFn- mole7o NaF-6 moLe7oKBF4 showed that a croloy-gM loop plugged after1440 h at a maJ(imum temperature of 60?"c and atemperature difference of l4soC, and that a Has-telloy-N loop was partially plugged after g?6b h ofoperalion under the same temperatur e condi -tions.sl The plug in the croloy was comprisedprimarily of pure iron crystals and the partialplug in the Hastelloy -N loop was made up of acompact mass of green single crystals ofNarcrFs. The salt charge from the Hastelloy-Ncontained large amounts of cr, Fe, Ni, and Mo, allmajor alloying elements in Hastelloy-N.

In more recent tests a purer fluoroborate saltof composition NaBF+-8 mole%o NaF has beenused.32r33 A thermal convection loop is being usedfrom which we can remove salt samples forchemical analysis and metal samples for weighingwithout interrupting operation of the loop. Theweight changes for the hottest and coldest samplesare shown in Fig. L2 for the two loops presenily inoperation. The loops are constructed of identicalmaterials, but the removable samples in one loop(NcL-13) are standard Hastelloy-N and those inthe other loop (NC L- 14) are a modified Hastelloy-N containing 0.5V0 Ti and only O.lVo Fe (standardHastelloy-N contains 4Vo Fe). As shown in Fig. 12,the weight changes of the modified Hasteltoy-Nare small€r, and this is later shown to be dueprimarily to the lower iron content of the modifiedmaterial. The rate of weight change was steadyexcept for a small perturbation after 1b00 h ofoperation and a large variation after 4200 h ofoperation. These times corresponded to timeswhen moist air inadvertently came in contact withthe salt. The changes in chemistry shown in Fig.13 also reflect the admission of air at these timessinc e the oxygen and water levels in the saltincreased. The iron and chronium concentrationshave continued to increase at a rate proportionalto (time)'/', indicating that the process is con-trolled by diffusion in the metal. The nickel andmolybdenum concentrations in the salt have re-mained very low except for times when moist airwas inadvertently contacted with the salt. Theseresults show qualitatively that the corrosion ratesincreased when the onFgen and water levels in-creased. Capsule tests in which sodium fluoro-borate containing 1400 ppm Oz and 400 ppm HrOwas contacted with Hastelloy-N for 6800 h at 607"Cexhibited very low corrosion rates (<0.1 mtl/year).Future work will be directed toward defining theo4ygen and water levels that result in acceptableeorrosion rates.

COLDEST SPECI

VIAr-/-'^/ ',/

MENS

l'

{::-:

r-

-a-a-O- ----1/

-c'

-rvl!

^-F!\___\\-^:\n-rTt-a'r

o Tl'

nvt ttr_Jt JrE\./ilvtLt\J f-607 "C \

\raf\lll l[,, Mnnttrttrn L]n QTtrt | ^v \

I l-l-l-\., | \N-NCL-{4 I

I STANDARD HASTELLOY ]

N -NCL-|3 | \o

N

5oc'rgfJ -2(9bzI(J,-4IIlrj3-6

-8

-{o

-1212345

TIME (h)6 (xt03)

Fig. L2, comparison of the weight changes of Hastelloy-N specimens inserted in NaBFa-NaF (92-grnoreld thermal convection loops. (Assuminguniform corrosion, a weight change of. 20 ^dcm' in 8750 h is equivalent to a corrosion rateof ".,1 mil/year.)

The information obtained on the changes in saltcomposition and the weight change of sampleslocated at various points (and temperatures)around the loop is sufficient to attempt a massbalance for the system. The weight of metal lostmust equal the weight of metal deposited plus theweight of metal in the salt-i.e.,

AWto""=Ardeposit.d* AW"ut, (1)

A weight change vs temperature profile is con-structed based on the removable samples and theassumption is made that each segment of the loopwall follows this same curve. This procedureresults in mass balanc€s, Eq. (t ), that closewithin L\Vo.

Diffusion theory can be used for further anal-ysis. As mentioned earlier, chemical analyses(Fig. 13) indicate that the iron and chromiumconcentrations in the salt are increasing, so it isassumed that the salt selectively removes theseelements from the alloy. The modified Hastelloy-N tested here is relatively free of iron, so the

,t

166 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 19?O

n"f/t-/ \

o

A

-----I

I

I

ol

T'l ./ i i)r i i

x \o \o

\_ra -:J \o

oo

^ 3OOOEC\I6 zoooFG,Fz tooolrJo=ooo

400

300

E

3 25o

zIk 2ooEFzUJ2 tsoo()

200

r50

roo

McCoy et al. MATERIALS

metal that is not diffusion controlled. Two obser-vations argue against the latter possibility. First,microprone analyses have not shown any transferof nickel or molybdenum to the colder surfaces of

the loop except during a brief period after 4200 h

of operation in which we knew that large amounts

of impurities were present. A second and more

convincing argument is based on the relativebehavior of standard and modif ied Hastelloy-Nduring the first 4000 h of operation. Rewriting of

Eq. tZl in terms of a reaction rate constant, K,

instead of the diffusion coefficient yields

350

ffi4ooo 5ooo 6000 Tooo 80oo emoTIME (h)

Fig. 13. Variation of impurities with time in NaBF4-

NaF (92-8 molefle), thermal convection loop

NCL-14.

weight loss of this sample should be due primarilyto the removal of chromium. The quantity of

material removed by diffusion under conditions

where the surface concentration of the diffusing

element is zero is given bY:

LM = Qo'lTi ,

where

A, M = material removed , g/ em'

Co = bulk concentration of diffusing species,g/cmt

D = diffusivity, cm' f sec

t = timer sec.

using the data of Grimes et a1.34 for diffusion of

chromium in Hastelloy -N, this analysis showed

that the quantity of chromium removed by dif-fusion cannot aecount for the total weight lost.

This discrepancy can be accounted for by

short-circuit diffusiott mechanisms entrancing the

rate of chromium removal at these low tempera-

tures or by impurities (likely HF formed by water

ingestion) lnrt lead to some general attack of the

(3)

(2)

AM = Qorm

This equation predicts the material transported by

diffusion to be too row for the modified alloy when

K equal D , but let K take on a value so that the

p""di"ted and observed weight losses for a given

time of operation agf ee with Co equal to 7Vo Cr'Now consider the standard alloy in which co

corresponds to |Vo Cr + 4Vo Fe. The same Kchosen for the modified alloy predicts the ob-

served weight change for the standard alloy'Thus, the difference in the weight losses between

the modified and standard alloy appears due

principalty to the iron content. Had much general

corrosion occurr€d, this adjustment in Co should

not have worked. In fact, this analytical proce-dure was entirely unsatisfactory for the short

time period after 42OOh when the water and oxygen

levels were high and nickel and molybdenum were

being ""*ou"d (Figs. 12 and 13). A further

possibl" role of impurities is to provide the

oxidizing potential necessary to keep the surface

concentr"tiott of iron and chromium at zel^o'

Thus, even though the process remains diffusion

controlled, the rate can be increased by im-puriti es.

Although some of the curves in Figs - 12 and 13

have quite large slop€Sr the corrosion rates are

not very high. using the rather rough method of

converting ift" weight losses to corrosion depths

indicates that the average rate during 8000 h of

operation has been 0.? mil/year. The rate has

deereased to -0.3 mil/year for long time periods

in which operation was not disturbed. In scaled-

up MSBR systems, & cold trapping technique willtit<ety be used to remove some of the corrosionproducts so that their solubilities are not ex-

ceeded. we presently feel that the sodium fluoro-

borate salt wilt provide a satisf actory and

economical secondary coolant for molten-salt re-actor s.

SUTiIIARY

Experience with the MSRE has proven the basic

compatibility of the graphite-Hasteltoy-N-fluoride

Fe

v Cr

-/{ ,/r a 7 a

//

/

I I \ L,/ t\ -

NUCLEAR APPLICATIONS & TECHNOLOGY VOL' 8 FEBRUARY 1970 167

McCoy et al. MATERIALS

salt system at elevated temperatures. Howev€r,a molten-salt breeder reactor will impose morestringent operating conditions, and some improve-ments in the graphite and Hastelloy-N for thissystem are needed. The mechanical properties ofHastelloy-N deteriorate under thermal-neutronirradiatiotr, but the addition of titanium in combi -nation with strong carbide formers such as nio-bium and hafnium makes the alloy more resistantto this type of irradiation damage. Graphiteundergoes dimensional changes due to exposure tofast neutrons, and the possible loss of structuralintegrity due to these dimensional changes pres-ently limits the lifetime of the core graphite.Although the core graphite can be replaced asoften as necessary, these replacements influencethe economics of the reactor, and a program tofind a better graphite has been initiated. Studiesto date indicate that graphites can be developedthat have better resistance to irradiation damagethan conventional nuclear graphites. The graphiteused in the core will be sealed with pyrocarbon toreduce the amount of t*xe that is absorbed.Techniques have been developed for this sealing,and studies are in progress to determine whetherthe low permeability is retained after irradiation.Corrosion studies indicate that the corrosion rateof Hastelloy-N in sodium fluoroborate is accep-table as long as the salt does not contain largeamounts of impuriti€s, such as HF and HrO.

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NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 169