ORNL-TM-2136.txt

 

  

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OAK RIDGE NATIONAI. I.ABORA'I'ORY;"
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for the - | |

v. S ATOMIC ENERGY COMMISSION

ORNI. TM- 2136

 

 

 

 

_ GRAPHITE BEHAVIOR AND ITS EFFECTS ON MSBR PERFORMANCE

 

 

 

 

o HDT!BE This documenf contains - information of a prellmmnry nature
" . and was prepared primarily for internal use at the Oak Ridge National
Loboratory. It is subject to revision or correcflon ond fherefore does .

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" contractor of the Commission, or employee ‘of such confractor, to the extent that such employee

‘or contractor of the Commission, or employee of such contractor prepores, disseminates, or

 

LEGAL NOTICE

This report was prepared as an account of Government sponsored work. Neither the United States, -

nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, exprossed or implied, with respect to the accuracy,
completeness, or usefulness of the information contained in this report, or that the use of

" any mformahon, cpparetus, method, or Pprocess disclosed in this report may not infringe

privately owned rights; or : ' .

B. Assumes any liabilities with respect to |he use of, or for demages nsuhmg from the use of
any information, apparatus, method, or process disclosed in this repors.

As used in the above, ''person acting on behalf of the Commission’” includes any employee or .

provides access to, any information pursuant to his employmenf or contract mth the Commtss:on,
or I'us employment wnth such contractor,

 

 

 

 

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ORNL-TM-2136

1

Contract No. W-TkO5-eng-26

MOLTEN SALT REACTOR PROGRAM

GRAPHITE BEHAVIOR AND ITS EFFECTS ON MSBR PERFORMANCE

 

 

 

P. R. Kasten
E. S. Bettis S. S. Kirslis
W. _ Hq Cook H. E. McCOy
W. P. Eatherly A. M. Perry
‘ D. K. Holmes ‘ R. C. Robertson
» Ro Jo Kedl Do Scott
. - C. R. Kennedy -R. A. Strehlow
"
LEGAL NOTICE
; This report was prepared as an account of Government sponsored work, Neither the United
| States, nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefulness of the information contained in this report, or that the use
of any information, apparatus, method, or process disclosed in this report may not infringe
yrivately owned rights; or

B. Assumes any Habilities with reapect to the u.éa of or for damages resulting from the

* use of any imformation, nppantus method, or process disclosed in this report.
: As used in the sbove, *‘person acting on behalf of the Commisaion’ includes any em-
)‘ ployee or contractor of the Commission, or employee of such contractor, to the extent that
. such employee or contractor of the Commission, or employee of such coniractor prepares,
disseminates, or provides access to, any information pursuant to his employment or contract
with the Commission, or his employment with such contractor.
FEBRUARY 1969
- OAK RIDGE NATIONAL LABORATORY
OCak Ridge, Tennesgee |
‘,fi operated by
- UNION CARBIDE CORPORATION

| for the
-~/ U.S. ATOMIC ENERGY COMMISSION

 

 
 

 

 

 
 

 

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y CONTENTS
: ABSTRACT =-=-=nsmmomemcomoomrommamn o e e em e men O 1
1. INTRODUCTION ----- - mememeeemeeeecceesceeescceenesacea———- 2
2. SUMMARY AND CONCLUSIONS --cnecomommmmmmmmnommmnmnmmnmnemmmnmnmee 3
3. GRAPHTTE BEHAVIOR -me-eeeesan- femmmmmcecememmemmmmeme—e—aememn O
3.1 Irradistion Behavior of Graphite --------;---- --------------- 10
3.2 Stresses (Generated in Graphite During Irradiation ---------- 17
3.3 Penetration of Graphite by Gases and Sgltg -wececvecceew —e—— 32
3.3.1 Penetration by Gases -weeccccccccccana. cmsmme———— wm—— 32
3.3.2 Penetration by 58ltg ~ewececcccccccccncocccccccocaaaa 34
3.3.3 Pore Volume Sealing Technique ------cnmnmmmcmmmmnmme 36
3.3.4 Surface Coatings and Seals ~ew-w—ceeee ;--;-. .......... 39
: 3.4 Near-Term Industriasl Production Capability -eccececccceece-- b1
€ 4. FISSION PRODUCT BEHAVIOR IN MOLTEN-SALT REACTOR SYSTEMS =cece—me- 42
‘ 4.1 In-Pile Capsule PeELE =-mmmmmommcmmanann ———mem————- e 43
4.2 Exposure Tests in the MSRE Core --=-e=- ——memscssaaaaa wommmmee 43
4.3 Tests in the MSRE Pump BOWl -=---ceccceccmacccccanaaax r—————- b7
4.4 Chemicel State of Noble-Metal Fission Products memamemenannn 50
L.5 Results from ORR Loop Experiments ~e-eeceeccccae-- ;;-; ....... 50
4.6 Evaluation of ResultE =mecesccceea cecmsmmsmn—————— ;--------- 51
5, NOBLE-GAS BEHAVIOR IN THE MSBR ---nc-m-mcocemnmmmmnmmmmnmnemsmnee 53
6. INFLUENCE OF GRAPHITE BEHAVIOR ON MSER PERFORMANCE -
. AND DESIGN ---n-- R —— s cmmmmme- 61
’ 6.1 Effect of Core Power Deneity on MSER Performence ----- wmome=s 61
= 6.2 Effect of Graphite Dimensional-Chenges on MSER
\iJ ‘ PerfOormance -=ecemccmcaccacecaw o o 0 o o 62
6.3 Mechenical Design Factors &nd Cost Considerations =eeecececea- 65

 
 

 

 

iv

6.4 The Influence on MSER Performance of ‘Noble-Metal

Deposition on Graphite _----,------_-----.--_---; -------- - Th

" 6.5 Conclusions e e o 0 om0 0 0 -----;--; 79
7. PROGRAM TO DEVELOP IMPROVED GRAPHITES FOR MSER'S e —- 80
- 7.1 FundamefitalPhySical Studies mimismmmmamm el o ma————m—m—— 82
T.2° Fundafientai Chemical Studies ----------------;------;---;;_- 83

' 7.3 Febrication Studles e ———————————— '-~---;---4--;- 8k
Tk Enéineering Properties i m—mn oo me e e 85
7.5 TIrradistion Program e e eim e m 86

. 7.6 Conclusions ~eececccaces ;-,--------f-----_--_,-f;----. ....... 86

APPENDIX Graphite Exposure MEasurements and Their '
: Relationships to Exposures in an MSBR ««ec-acccccavcccas --

 
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GRAPHITE BEHAVIOR AND ITS EFFECTS ON MSBR PERFORMANCE

P. R. Kasten
E. S. Bettis S. S. Kirslis
W. P. Eatherly A. M. Perry
D. K. Holmes ' R. C. Robertson
R. J. Kedl - D. Scott
C. R. Kennedy R. A, Strehlow
ABSTRACT

Graphite behavior under Molten-Salt Breeder Reactor
(MSBR) conditions is reviewed and its influence on MSEBR
performance estimated. Based on the irradiastion behavior
of small-sized graphite specimens, a permissible reactor
exposure for MSBR graphite is sbout 3 x 1022 neutrons/cm2
(E > 50 kev). The stresses generated in the graphite due
to differential growth and thermal gradiente are relleved
by radistion-induced creep, such that the maximum stress
during reactor exposure is less than 1000 psi for reactor
designs having & peak core power density of sbout 100
kw/liter and reactor exposures less than about 2- 1/2 years.
The corresponding power costs for single-fluid MSBR's
would be ebout 3.1 mills/kwhr(e) based on a capital charge
rate of 12% per year and an 80% load factor. Experimentel
data on graphite behavior also indicate that graphites -
with improved dimensionsl stability under irrediaetion can
be developed, which would lead to improved reactor per-
formance.

The deposition of fission producte on graphite does
not appear to be large (10 to 35% of the "noble-metal"
fission products based on MSRE experience); taking into
account graphite replacement every two years, fission .
product deposition reduces the MSER breeding ratio by
gbout 0.002. Also, it appears that xenon poisoning cen
be kept at a 0. 5% fraction poisoning level by using pyro-
lytic carbon as & pore impregnant which seals the surface
of MSBR graphite and/or by efficient gas stripping of the
fuel salt fluid by injecticn and removal of helium gas
bubbles.

It 1s concluded that good MSBER performance can be
obtained by using graphite having combined properties
presently demonstrated by small-size samples, and that
development of MSBR graphite having such properties is
feasible.
 

 

1. INTRODUCTION

Recent experimental results"concernihg the physical behavior of
graphite during reector irradiations have indicated that sighificant
dimensional changes can take place at exposures of interest in Molten-
Salt Breeder (MSBR)'sysfems. These results indicate the need to efialuate
graphite behavior under MSBR conditions, to estimate what constitutes a
rermissible reactor exposure for the graphite, to determine the influence

of core power density and graphite replacement costs on MSER performance,-

and to initiate an experimentsl program for the purpose of developing
improved graphite. Also, in assessiné overall reactor performance, a
nurber of other interrelated problems are involved. For example, the
deposition of fieSion produets on graphite has an adverse effect on reac-
tor performance,-and this deposition 5ehevior in an MSBR environment
needs to be determined. Thus, the purpose”of this study is to summarize
" and:evaluate presently available information concerning graphite behavior
and properties as they relate to MSBR operation. Furfiher, investigations
are proposed which may lead to development of improved graphites. Topics
specificelly treated in this report inelude the behavior of graphite |
under reactor radiation eonditions; the evaluation of irraaiation data;
the stresses generated in graphite under MSBR conditions; the‘pefletration
of graphite by gases and salts; the sealihg of'graphite peres; the depo-
sition of fission products on graphite; the effects of gas stripping and
of graphite permesbility on 135%e neutron poisoning; the influence of
graphite dimensional changes on MSER fuel cycle performance, mechanical
design, and power costs; the effect on MSBR fuel cycle performance of
£ission product depritien on graphite; .and a proposed program for devel-
oping improved graphites which includes;physical, mechanical, chemical,
fabrication, and irradiation studies. |

As mentioned sbove, the effect of graphite behavior on reactor per-
formance influences reactor design. Until recently, the term MSBR vas
applied to a two-fluid concept, in which fuel salt containing fissile
material was kept separate from fertile-containing fluid by means of
graphite plufibing. Such a'concept is giveh in reference 1,which presents

 

1MSR Progrem Semiann. Progr. Rept. Aug. 31, 1967, ORNL-4191
(Dec. 1967).

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design information on a lOOOeMw(e) plant employing four reactor modules,
each module generating the equivalent of 250 .Mw(e). The core of each
reactor uses grephite fuel cells in the form of reentrant tubes brazed
to metal pipes. The pipes are welded 1nto.fue1-aupply.and discharge-
pPlenums in the bottom of the reactor vessel. . The fertile salt fills the
interstices between fuel cells as well as a blanket region around the
core. Such & reactor is termed a two-fluid MSER.

Also considered here is a single-fluid.MSBR, in vhich the fissile
and fertile salts are mixed together in caerrier salt but which is other-
wise similar to the two-fluild MSER. Such a concept does not require
graphite to serve as fuel plumbing, which is deslrable fromrthe viewpoint
of reactor operation. ,H0wever,_in order to.oPerate,asazbreeder, a fuel
processing scheme is required that can rapidly and economicallyrretain
233pg outside the core reglon. Recent chemical developments indicate?
the feasibility-of such a process. Thus, both the two-fluid and single-
fluid MSBR's are referred to in the.folloning sections. However, no
differentiation is made to items which apply equally well to both reactor
concepts. |

2. SUMMARY AND CONCLUSIONS

When graphite is exposed to fast neutron doses, it tends to contract
initially, with the rate of contraction decreasing with exposure until a
minimum volume is attained; further exposure tende to cause volume expan-
sion, with the rate of expansion increasing rapidly at neutron doses above
about 3 x 1022 neutrons/cm2 (E > 50° kev) in graphite tested to date. This

'behavior.is due to atemic displacements which take place when graphiteris
exposed to fast neutrons, andtisldependent upon'the*sourCe-and fabrication
. history of the material and also the exposure temperature. Irradiation‘
‘results for different grades of graphite have shown that gross volume

changes are & function of crystallite arrangement as well as size of the

~individual crystallites. The initial decrease in graphite volume with

reactor exposure 1s_attributed ‘to the closing of voids which vere gener-,
ated in the graphite during fsbrication. These voids (as microcracks)

 

3MSR Program Semisnn. Progr. Rept. Feb. 29, 1968, ORNL-425L.

 
 

 

afford accommodation of the internal shearing strains without causing
gross volume growth which would otherwise teke place due to the differ-
ential growth rates of coke particles. Once the original microcracks
are closed, however, this accommodation no‘lOnger exists, and macroscopic
growth occurs with increasing exposure. o

.The rapid volume expansiofi of graphite observed &t very highfreactor
exposures indicates that for these conditions the internal straining is
not acccmmoaated by particle deformation, but by craeking. Exeminations
show that this cracking generally takes place in the interparticle, or
binder region. Thus, it appears that the binder region has little capacity
to accommodate or control particle strain and thus fractures because of
buildup of mechanical stresses. This indicates that graphites with im-
proved radiation resistance might be obtained by developing graphites
having little or no binder content, and there are experimental results
which appear to encourage such development. Experimental data also indi-
cate that improved radiation resistance is associated with isotropic
‘graphites made up of large crystallites. Consequenfily, a research and
development program aimed at producing improved graphite would emphasize
development of graphite having large erystallite sizes and little or no
binder content. Such a program would involveephysieal, chemical,.fiechan—
ical, fabrication, and irradiation studies, end could be expected to
develop graphites with permissible fast neutron exposures of 5 to
10 x 1022 neutrons/en® (E > 50 kev).

_ Volume changes in graphite during irradiation can influence reactor
,performance'characteristics and thus affect MSBR design 3pecif1cations.,
Consistent with the desire to maintain low permeablility of the graphite
,to‘gases, obtain high nuclear performance during MSBR operation, and to
simplify core designlfeatures, the maximum permissible graphite exposure
- was limited to that which cauees the graphite to expand back to its original
vvolume. . On this basis, and considering results obtained to date with
present-day graphites, the permissible exposure under MSER conditions is
estimated to be about 3 x 1022 nvt (E > 50 kev) at an_effecfiive tempera~
ture of 700°C._ More specifically, at a peak core pover density of 100
kw/liter under MSBR operating temperatures, return of the graphite to its
original volume corresponds to about 2. > years of reactor operation at
90% load factor.
 

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Neutron-flux_gradients in the MSER will lead to differential volume
changes in graphite components, end if the graphite is restrained from
free growth, étresses areAgenerated; The magnitude of the stress depends
on the fast neutron fiux distribution end also on the radiatifin;in&uced
creep of the graphite. Based on a single-fluld MSER design 1n'which the
péak fower dens1ty-1s 100‘kw/11ter and where the grephite shape is repre-
sented by an annular graphite cylindef'having an external radius of 5 cm
and an internal radius of.1.5 cm, the maximm calculated stress ih the
grephite during a.2.5-year reactor exposure was 1ess than TOO pel due to
spatialiy symnetric neutron flux variations, and less than 240 psi due to
asymmetric flux variations (fiux variations around the tube periphery).
Since-MSBR graphite;is estimated to have a tensile strength of ebout 5000
psi, the abdve stresses due to changes in graphite dimensions do not
appear to be excessive. For, the above conditions, the net change (decrease)
in the length of the graphi@b cylinder is estimated to be about 1.6, an
amount which does not appear to introdfice significant core design 4iffi-
culties. | B D
Graphite for en MSBR should have low penetration by both gas and salt,
in order that performance characteristics of the system remain high., If
neutron poisoning due to 1%5Xe is fio be limited to 0.5% fraction poisons
fiy diffusional resistance of the graphite alone, a material is needed in
which the xenon diffusion coefficient is sbout 10™® £t2/hr. The most
promising of several epproaches for producing such a graphite is that of
sealing the surface pores with pyrb1ytic'éaern or graphite. Experi-
mental results indicate that grafihite gealed in this manner has e dif-
fusion coefficient'of_about 1078 ftzyhr (associated with the surface geal),

- and that this seal can be maintained éven-though some thermal cycling

OCCUrs. Alternatively; neutron poisonihg.could bermaintaifled‘at low
levels by efficient stripping of fission gases from the fuel salt with
hélium,iand if this is éccomplished,jan increase in graphite permedbility
during reactor exposure may'belpermissible. fDue to the nonwetting

~ characteristics of molten fluoride salts, penetration of graphite by

salts does not appear to be a prdblem._ _
Fission products other than gases also have access to the graphite.
Retention by the graphite of fission products could significantly reduce
 

.the nuclear performence of MSER systems. However, tests conducted in the
Molten-Salt Reactor Experiment (MSRE) have demonstrated that only & small
fraction of the total fission producté'genefated accumulate on the graph-
ite. The primary intersction between MSRE graphite and fissioning fuel
salt is the partial deposition (sbout 10-35%) of fission products that
form relatively unstable fluorides. Of the "noble-metal" fission products
which deposited, over 99% of the_associatéd activity was within 5 mils of
the graphite surface. In no case was there permeation of fuel salt into
the graphite or chemical demage to the graphite. Test results can be
interpreted such that the percentage of the noble metals deposited on
graphite depends on the ratio of graphite surface to metal surface in the
fuel system, with deposition on graphite decreasing with decreasing ratio
of graphite-to-metal surface. Finally, the MSRE results indicate that
significant fractions of the noble-metal fission products appear in the
gaé phase in the fuel pump bowl. If these fission products can be re-
moved from MSER's by gas stripping, such a process would provide a con-
venient means for their removal. '

Based on the results obtained in the MSRE .and taking into account
the higher metal/graphite surface area in an MSBR relative to the MSRE,
it is estimated that deposition of fission products on the graphite in
an MSER would reduce the breeding ratio by ebout 0.002 on the average if
graphite were repléced every two years, and sbout 0.004 if réflaced every
fdur years. Thfis, although complete retention of the noble-metal fission
products on core graphite would lesd to a significant reduction in MSER
breeding ratio, the deposition behavior inferred from MSRE results corre-
sponds to only a small reduction in MSER performance.

Graphite dimensional changes due to exposure in an MSBR can alter
the relative volume fractions of moderator, fuel salt, and fertile salt
in the reactor. Such changes influence the design of a two-fluid MSER
moré than a single-fluid reactor, since in the latter the fertile and
fissile materials are mixed together and their ratio does not change
vhen the graphite volume changes. By constructing a two-fluid reactor
- such that the fissile and fertile materials are confined to channels
within the graphite assemblies and the spaces between graphite assemblles
are filled with helium, changes in graphite volume fraction lead largely

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to reletive.volume change in the helium space. Such volume changes have

~only a small_effect on fuel cycle performance and on power distribution.

In a. single-fluid MSER, graphite dimensional changes would have little
effect on nuclear performance since the fissile and fertile salt volumes
are equally affected. Also, the ability to independently adjust fissile
and fertile material concentrations in both two-fluid and single-fluid
MSBR's permits adjustment in reactor performance as changes in graphite
volume occfir. Thus, 1itt1e change in nuclear'performance is expected
because of radiation damage to graphite, g0 long as the graphite volume
does not increase much beyond its initial value and the graphite diffusion
coefficient to gases remaine low during reactor exposure (the latter con-
dition neglects the posaibility of removing xenon efficiently by gas
stripping). . |

A limit on the permissible exposure of the graphite can have a sig-
nificent influence on reactor power costs. If there were no exposure
limit, the average core powver density corresponding to the minimum cost
would be in excess of 80 kw/liter. However, if a limit exists, high
pover density can lead to high cost because of graphite replacement cost.
At the same time, decreaéing the core power density leads to an increase
in capital cost and fuel cycle cost. Thus, a 1limit on permissible graph-
ite exposure generally requires a compromise between various cost items,
with core power density chosen on the basis of power cost. The optimum
power density also varies fiith MSER concept gince only graphite requires
replacement in a single-fluild MSER, while both the reactor vessel and
graphite appear to require replacement in a two-fluid MSBR'because of the
complexity of constructing the latter core. Further, reactor power out-
age due aolely tografihitereplacefient'reqfiirements can be a significant
coet factor. However, if graphite were replaced at time intervals.no
less than two years, it appears feasible to do the replacement operation
during normal turbine maihtenance periods, such that nc effective down-
time is assigned'to'graphite replaCement._ A tfib-year time iaterval is
associated with an average power density in the power-producing 'core”

of about kO kw/liter. For the sbove "reference" conditions, the single-

fluid_MSBR has power costs about 0.35 mill/kwhr(e) lower than the two-
fluid MSBR. Doubling the permissible graphite exposure [fb a value of
6 x 102 nvt (E > 50 kev)/ would be more important to the two-fluid
 

 

 

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concept and would reduce power costs by about 0.15 mill/kwhr(e); the
corresponding change for the single-fluid MSER would decrease power costs
by ebout 0.07 mill/kwhr(e). If a two-week effective reactor downtime |
were assigned solely to graphite replacement operations,.the associated
power cost penalty would be about 0.05 mill/kwhr(e) for either'concept.

) Conclusions’éf these studies are: B
1. Satisfactory'MSBR perfofmance can be obtained using graphite having

. the combined”properties presently demonstrated'fiy-small-sized Samples,

| with single-fluid MSBR's appearing economically superior to two~-fluid

MSBR's.

2. ' The development of MSBR graphite having desired properties is feasible.
(It appears that at least two vendors could produce a material satis-
factory for initial MSBR use, based on present industrial éapahility
for graphite production.) ”

3. The radiation behavior of small-sized graphite specimens indicates
a permissible reactor exposure in excess of 2 years for a peak MSBR
power density of 100 kw/liter, based on & zero net volumetric growth
for graphite exposed to the pesgk pofier density. The maximum stress
generated in the graphite under these conditions due to dimensional
changes and thermal effects is estimated to be a factor of 5 less |
than the expected tensile strength of MSBR graphite. |

4, The deposition of fission products on/in graphite does ndt appear
to influence nuclear performance significantly. Deposition of
noble-metal fission products appears to reduce the breeding ratio
about 0.002 every 2 years of graphite exposure. Also, it appears
feasible that xenon concentrations can be kept at a 0.5% fraction

poison level by‘suiface sealing~of the graphite with pyrolytic
carboh; further, gas stripping provides a means of keeping‘xenoh

poisoning at a low level,

5 The désigp and operation of MSBR's appear sufficiently flexible
that a high nuclear performance can be maintained even though

graphite undergoes dimensional changes during reactor operation.

 
 

 

 

 

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3. GRAPHITE BEHAVIOR
" H. E. McCoy

Althoughvthe dimensional instability of graphite‘under neutron irradi-
ation_has been known for some time, volume changes associated with very
high reactor exposure appear to be greater then originally anticipated.
Until recently, grephite had been éxposed to:fast neutron doses of only
gbout 1 x 10%2 neutrons/cm®. ISotropic graphite was noted to contract,
vith the rate of contraction continuously decreasing.: It eppeared that
the contraction would cease and that the dimensions would begin to expand
slightly as defects were produced by irradiation. However, graphite has
now been irradiated to higher doses,, and a very repld rate of expansion
is noted after the initisal contraction.‘ A large and rapid physical expan-

- sion is undesirable from the viewpoint of reactor performance, also, if

tke penetration of xenon 1nto graphite vere to increase markedly as the
graphite density decreases, the nuclear performance would be adversely
affected. Based on present information,'a reasonable core design life
appears to be that which permits the graphite to return to its original
volume. : | _

The initial graphite contraction with exposure would lead to en

~ increase in the volume fraction of salt within the core regibn of the

reactor. Since the oontraction'wonld teke place slowly_with time, the
nuclear performance of'thessystem could remeinfrelatively constant by
adjusting the fuel concentration, and if the graphite nolume‘fractlon-
did not increase much above its initial value, Expension of the graphite
would lead’ to a decreaSe in the ealt volume in the core, and eventually
lead to a decrease in nuclear performance of the system. However, i
the core graphite vere replaced.before it expanded much beyond 1ts
original volume, the effect of moderetor dimensional changes on nuclear
performance would be small.

Graphite for MSBR use ehould have low penetration by both gas and

'_salt so0/that the nuclear performance will remain high Since salt nor-

mally does not wet graphite, there 1s little tendency for the salt to
penetrate. the graphite unless high pressures are applied or. wetting con-
ditions erise,_and these latter conditions would normally not exist.

 
 

 

 

 

10

Gaseous penetration is cofitrolled by the diffusion coefficient of the gas
in the graphite and by gas stripping with helium bubbles. The'mest sig-
nificant of the fission product. gases is 1%9%e. Even though xenon can
be removed by stripping the salt with helium‘bubbles, it is desirsble
that the graphite have and retain e very low permesbility so as: “to main-
tain xenon retention in the eore at a low leVel Ways for developing
such a graphite are listed below, with method three the preferred one
at present. | | |

- 1. Development of a monolithic graphite having the desired

characteristics.
2. Impregnation of the graphite with pitch,
3. Deposition of pyrolytic carbon within graphite by
decomposition of hydrocarbon gases.

k, Deposition of metal on the graphite surface., -

An important eonsideration is the ability of the MSBR graphite to
retain low values of the gaseous diffusion coefficient throughout the
reactor exposure period. |

As indicated above, the proposed use of graphite in molten-salt
breeder reactors poses some rather stringent requirements upon this mate-
rial. Tt must have excellent chemical purity in order to have the desired
nuclear properties. It should be impermeable to molten selts and have a
diffusion coefficient (to gaseous fission products) of about 108 £t%/hr.
Also, the graphite must have reasonsble dimensional stability to fast |
neutron doses in the range of 1022 to 102® neutrons/cm® (E > 50 kev).

In the next sections a critical eesessment 1s madé ‘of the status of

graphite development for molten—salt breeder reactors.
3.1 Irradiation Behavior of Graphite.

C. R. Kbnne&y

Graphite undergoes displacement damage under neutron 1rradiation,

. reeulting in anisotropic crystallite growth rates. The crystal expands
in the c-eXie-flirection and experiences an:a-axis contraction. 'Irradi—_
etion studiessieh'iSOtropic large-crystallite pyrographite have shown_

 

- 3p, T, Netlley and W. H. Martin, The Irradiation Behavior of Gra _phite,
TRG Report 1330(c) (1966) |

 
 

 

 

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that the overall growth rates correspond to & very small volumetric:
expansion. The volume expansion is attributed to minor adjustments in
lattice parameters to accommodate the vacancy and interstitial atom con-
centrations. However, the linear growth rates in highly orienteted pyro-
graphite are extremely large and represent the growth rates of individual

 crystallites of the filler coke particles in reactor-grade graphite.

Also, the irrsdiation behavior of graphite is dependent upon its fabri-
cation history.

A comparison of graphite irradiatibn behavior obteined at different
laboratories is made difficult by the various éxposure scales used by
the different experimanters.‘ Perry* has examined this problem and con-
cluded that an exposure scale based upon neutrons with energies greater

than 50 kev can be used to compare results obtained from widely different

reactors. This exposure scale will be used in our enalysis of the

| existing data.

Reutron irradietion causes various grades of graphite to undergo an
initial decrease in volume rather than the expansion observed in pyro-
graphite having an equivalent crystallite size. -Irradietion resultsS’®
are glven in Figs; 3.1 and 3.2 for an isotropic and an anisotroplc grade
(AGOT), respectively. The actual changes in linear dimensions are, of
courbe, different from grade to grade, and depend largely on the degree
of anisotropy present in the graphite, The initiel decrease in volume
is attributed to the closing of voids generated by thermal strains
during cooling in the fabrication process. The closing of the vold volume
is asccompanied by c-axis growth and a-exis shrinkage. The oriEntafion.of
the crack or void structure, due to.the thermal strain origin, allows the
c-axis growth to be sccommodated internally; the changes in crystallitg
difiénsions do not contribute to the overall changes 1n macroscoplc
dimensions until the cracks are closed.

 

~ %A, M. Perry, eppendix of this report.

SR, W. Henson, A. S. Perks, &nd J.H.VW. Simmons, Tattice Parameter
and Dimensional Cheanges. in Graphite Irradiated Between 300 and 1350°C,
AERE R 5kB9. .

®J. W. Helm, Long Term Radiation Effects on Graphite, Paper MI TT,
8th Biennial Conference on Carbon, , Butfalo, New York, June 1967T.
 

12

ORNL-OWG 67— 10501A

 

o

 

550-600° C

 

N / //

7 / 300-440°C —

 VOLUMETRIC CHANGE (%)
o

/

 

 

 

 

 

 

 

 

-0

 

0 10 20 30 40 50 (u102')
NEUTRON FLUENCE (neutrons/cm2) (£ >50 kev) '

Fig.. 3.1, Volume Changes in Isotropic Graphite During Irradiation in
the Dounreay Fast Reactor

ORNL—-OWG 67-10500RA

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20
15
775-825

10
&
: |
z / 925-975°C
<
© 5
ul
S
2
3
> .

0 /
=5 [ 1000-1075 \550-—700
S 575-625 .
400- 475
=10 2
0 ' ‘5 10 15 20 - 25 (x10°)

NEUTRON FLUENCE (neutrons /em?) (£ > 50 kev)

Fig. 3.2. Volume Change in Anisotropic (AGOT) Graphite During GETR
Irradiationsr

 
 

 

o

-’

r

"

 

13

The original graphite void volume also affords & degree of sccommo-
dation of the internal shearing strains that would otherwise be produced
by the differential growth rates of graphitized coke particles. However,
once the cracks are closed, this accommodatioh no longer exists, and the
macroscopic dimensional changes should then reflect the c-axis growth.

If the shear_etrains are accommodated as in isotropic pyrolytic

cafbon,7_iarge internal shear strains resulting from more then 160%

differential growth of the crystallites can be accommodated fiy plastic
aeformafion without intermal fracturing of the graphite and with very
small gross volumetric-expansions. However, as shown in Figs. 3.1 and
3.2, experimentel results show that, for samples tested, the graphite
generally contracts to & minimum voiume and then expands very rapidly.
The very tapid rate of volume expansion indicates that the expansion in
all directions is dominated by c-axis growth. This is difficult to
explain unless continuity in the direction of the a-axis has been lost,
since there are two a-axes in the crystal and only one c-axis. It, there-
fore, appears that continuity has been lost between the ad jacent grains
and -that overall the a-axie contraction cannot restrain the c-axls growth.
The above explanation for the changes teking plece inside the

graphite implies that the internal straining due to differential growth
is accommodated primarily by cracking and not by deformation. To date,
the highly exposed graphites have been subjected to casual, low-magnifi-
cation surface examinations.. These:reveal, es expected; that the gereral
reglion of faillure has been in the 1nterpart1c1e or binder region. Only
one isolated case has been found of a crack running across the layer
plenes of e particle. These results indicate that the binder region
has little capacity to accommodete the sheer strain and as & result it
fractures. ‘

~ If the graphite-volume decrease (during irredistion) ie & result of
closing the voidsgenerated;by'thermal etreine (introduced during fabri-
'cetion), the minimumAdécrease'in volume'and»the'exposure required to

 

7J. C. Bokros and R. J. Price, "Radiation-Induced Dimensional Changes
in Pyrolytic Carbons- Deposited in & Fluidized Bed," paper presented at 8th
Biennial Conference on Carbon, Buffalo, New Ybrk, June 1967 (proceedings
to be issued)
 

 

 

1k

achieve the minimum volumé should be temperature dependent; i.e., there
would be partial closure of the void volume simply by the thermal expansion
accompanying heating. Therefore, increasing the irradiation temperature
should decrease the irradiation growth required to close the cracks and
achieve the minimum volume. Thus, unless the irradiastion growth rates in
the c-axis and a-axes vary appreciably with temperature, the time to con-
tract and then to expand to a specified volume should decrease with in-
creasing temperature. This behavior has been observed as shown on Figs.
3.3, 3.4, and 3.5 using the results of Henson et al.® and Helm.® Figure
3.3 gives the maximum volume contraction as a function df'temperature.‘
Figure 3.4t and 3.5 give, respectively, as functions of temperatfire, the
total exposure required to achieve maximum graphite volume contraction

and that required for the graphite to expand back to its original volume.

ORNL-DWG 67-10495R

 

 

 

14 l |
e UK ISOTROPIC
A UK PGA
12
0o AGOT
& CSF
10

 

N
Y

 

 

 

6 zy,,
. | % ,4/ |

 

VOLUME CONTRACTION (%)

 

 

 

 

 

 

 

 

 

 

 

0
300 400 500 600 700 800 900 1000
TEMPERATURE (°C)

Fig. 3.3. The Maximum Volume Decrease of Graphite During Reactor Exposure

 

“
 

15

ORNL-DWG 67-10498R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

{x1024) - _ ’
*
30 XA ® UK ISOTROPIC
» ‘QV . A UK PGA
1 © AGOT
s CSF
7\ |
25
=~ ‘%/
2
e .
A 20 7,
W V777N
+ <
£
w
L5 a2
0
O
a
b
w
{0 Ay
>
0 .
- 300 400 500 600 700 800 900 1000
TEMPERATURE (°C)
Fig. 3.4. Reactor Exposure Corresponding to Minimum Graphite Volume
S
a ORNL-DWG 67-10497R
: ' (x402h , '
. PN ® UK ISOTROPIC
7 4 UK PGA
é © AGOT
— ‘ A CSF
| 3 50 A
i x .
! O
F <
| Wy 40 — ' 2,
: "§
g 2
x 30 3
: =2
_ 2
: o
: 5
w 20 %
10 -
o . ,
300 400 500 600 700 800 900 1000
- S TEMPERATURE {°C) " - : »
_ Fig. 3.5. Total Reactor Exposure Required for Graphite to Expand Back
% to Its Original Volume
 

 

16

It should be recognized that these data were obtained from GETR and
DFR experiments; tfie_heutron energy spectra assoclaeted with these reactors
‘differ siénificantly, and‘the fast flux differs by almost an order of
magnitude. The data, however, correlaté well and if & dose-rate effect
exists, it appears to be very small over the temperature range studied.

In estimating the useful lifetime of the graphite for the MSBR, the
present informatiop on tested grades has been used. Some speculation is
required since there is little iInformation concerning the effects of
volume expansion on pore spectrum, gas-penetration characteristies, and
strength of the graphite. It apfiears probable that contraction followed
by expafision back to the initial graphite volume doee not create a
structure less sound than the original unirradiated material. On this
basis, the useful life of the graphite would correspofid to the exposure
required for the graphite to return to its original volume. Therefore,
based upon grades of graphite that have been tested and the results
shown in Fig. 3.5, the lifetime expectancy of graphite at T00°C would -
be sbout 3 x 1022 neutrons/cm® (E > 50 kev).

The graphite temperature in an MSBR varies with core design and
pover density and also with spatial position within the reactor. For
an MSER operating et an average power density of 80 kw/liter, peak .
graphite temperatures would be in excess of 750°C. However, peak tempera-
ture is probably not the proper criterion; rather, the volume-averaged
graphite temperature iIn the viecinity of the highest fast neutron flux
would be more appropriate. The peak volume-averaged temperature would
tend to decrease with increasing numbef,of fuel flow channels, with de-
creasing power density, and upon changing from two-fluid to single-fluid
type MSBR's. A value of T700°C is representative of the effective volume-
averaged temperature to be used in estimating permissible graphite
exposure for MSBR's operating at an average core power density of ebout
40 kw/liter; a more detailed analysis of graphite growth, temperature,
.and associated stresses is.given in Seétion'3 2 vhich verifies the above.

The effect of graphite size on dimensional stability during reactor
exposure has been reported by Nightingale and Woodruff.® Large blocks

 

R, E. Nightingale and E. N. Woodru:t"f, "Radiation Induced Dimensional
Changes in Large Graphite Bars," Nucl. Sei. Eng 19, 390-392 (196k4).

"

*

 
 

 

 

 

r;‘

\)

”

17

have shown a transverse shrinkege rate of up to twice that of subsize
specimens. Although the rationale for such behavior ig very vague, this
"size effect" has occurred. Unpublished data® from ENWL indicate that,
although the volumetric contraction in the transverse direction with large-
size graphite specimens is possibly greater by about 1% than that obtained
with small-size samples, the exposure required to obtain minimum volume
and reversel in volume growth has not been réduced{ Further, published
datayfrom.BNWL9 of & very preliminary nature indicate that extruded pipé
specimens of approximately 3 in. OD and 2 in. ID with about 0.2 in.
machined from each surface had the same growth rate as small-size speci-
mens. The "size effect" would, at the most, only require an allowance

for this increase in tfansvefse Shrihkage in the design. The sbove would
neither increase nor decrease the lifetime expectancy of the graphite.

3.2 Stresses Generated in Graphite During Irradiation

W. P. Eatherly and C. R. Kénnedy.

‘ The above discuasidns concerned the 11m1tat16ns on graphite lifetime
due to irradiation-induced dimenéional changes, for the case of graphite
in a strels-free condition. In actual fact, temperature and flux gradients
in the core will tend to produce differential distortions within the
graphite, thus generating internal stresses. 1In examining thesge effects,
a single}fiuid reactor will be cohsidered in which the core is conétrugted
of cylindricel prisms of graphite (i.e., tubes) running axielly through
the core. The stresses wlll arise £rom two distinct causes. _Within each
prism there will be symmetric neut:on_flux‘and temperature gradients due

to flfix distributions in & reactor "cell." 1In addition, across the prisms

there will be superimposed asymmetric gradients due to the gross radisl
flux and temperature distributions within the core. The symmetric'gradi-
ents will be maximum in the central region of the core vhere the pover
density is high;kthe-asymetric gradients will be maximum in the outer
reglons of the core where the-“bianket" region causes & rapid decrease in

" power density with increasing core radius. The symmetyric gradients will

be considered first.

 

°D. E. Baker, BNWL, privaté commmnication.
 

 

18

In examining stresses it is'neceésary'to relate the dimensional be-
havior of the grephite to the three independent variables of temperature,
flux, end time. In the temperature range of interest (550 to T50°C), the
dimensional behavior for isotropic graphite is approximated by

"_%‘; =31 (0.11 - 0.7 x 10 “T)(x -2x) - (3.1)
where

. T = temperature, °C, |
® = fast neutron f£lux, neutrons em 2 sec - (E > 50 kev),
t = time, second,

" and | | ,

_% = fractional length: change of graphite.

This function is plotted in Fig. 3.6 as a function of fluence with
'temperature'as a parameter; as shown,-A47£ is a strong function of the
irrediastion temperature. - - o |

The maximum internal symmetric flux gradients occur in the central
region,of the core; at this position the salt-to-graphite volume ratio
will be about 20%. An appropriate graphite cylinder size is one having
an internal radius, a, of 1.5 cm and an external radius, b, "of 5 cm; it
is assumed that surface temperatures will be the same on both surfaces.
The fuel salt enters the reector at & temperature of 550°c and exits at
700°C. Also, the neutron flux causing fissions, 9, ié-cénsidered to

vary as |
| | ¢ = 0, sin %E ‘ (3.2)
whére ,' | |
L = core height,
z = axial coordinate,
| ¢ = maximm flux.

With = maximum,core power density of 100 kw/liter, which is considered

here,

 
 

 

 

.t

-’

wilf

o

<%

 

19

ORNL-DWG 68-7974

 

 

 

 

AL /L ()

00°C

 

 

750

 

 

 

 

 

 

_// 550

500~

 

 

 

‘Fig. 3.6. Radiation Induced Dimensional
‘Various Temperatures

1

’ . ‘ 2 . .
" {¢ 1), NEUTRON EXPOSURE (neutrons /cm?2 x4022) (£ > 50 keV)

3‘.

'C'har_xges in _G_ilso Graphite at
 

 

 

ite is considered to flow in the radial direction.

20

Oy = h.s x 10** neutrons ecm 2 gec™t

-

The internal heating within the graphite will be due to energy depo- -
sition by both prompt and delayed y rays. For the assumed pegk power
density of 100 kw/liter, this energy deposition smounts to about 8 w/ce ¢
prompt and 2 w/cc delayed. Thus, the internal energy generation rate,
g, is approximately given by

+ 2 wfee .

q==831nL

This expression combined with the graphite geometry and dimension gives
Q, the heét transfer rate per unit length of graphite between the graph-
ite and the fuel salt. Since the radial temperature gradients are much
greater than the axial gradients, all the energy generated in the graph-

The heat genération in the flowing fuel salt will be.nearly pro-
portiofial to the flux é, and thus the temperature in the flowing salt
will have a cosine dependence on z. TFurther, the temperature drop, Amf,
from the flowing salt to the graphite-salt interface can be calculated

 

from | o ¢

_Q
o, = & | (3-3)

where the effective heat transfer coefficient h has the value,

h~ 0.731 weem 2-°C™2 (1240 Btu/hr-ft2-°F) .

This yields the surface temperature of the graphite. The internal graph-
ite temperatures follow immediately from the equations of heat flow in a
hollow cylinder with a uniformly distributed heat source. The calculated
salt, surface, and central graphite temperatures along the central axis
of the core are shown in Fig. 3.7. _ '
In- the single-fluild MSER under consideration, the fast flux decreases
about 5% from the surface of the graphite to its interior due to energy .
degradation. This relation is represented here by

%3-00531115-:?-:: ‘ (3.1!-) Q

where r 1s the radlal coordinate for the graphite cylinder.
 

L1

i

21

 

Y

<3

ORNL—DWG €8-7975

 

 

 

 

 

 

 

 

 

 

 

 

750
'/—_—-
- GRAPHITE -
700 CENTER
— GRAPHITE
& SURFACE
& BULK SALT
0 650
< /
o
w
o .
E . .
w
- /
600 : /
550 -
0 0.2 0.4 06 0.8 1.0

2/ L, AXIAL POSITION

Fig. 3.7. Axial Temperature Profiles in Center Channel of a Single-Fluid
MSBR : .
 

22

'Based on the sbove, the flux end temperature conditions in the graph-
ite tube are specified as a function of 2z and r, and thus, through'Eq.
(3.1), so is the local radistion-induced distortion. Thus, the induced
stresses can he obtained by solving the stress-strain equations. Before
doing this, it is helpful to review briefly the creep behavior of & uni-
axiaily loaded graphite bar under irradiation, and define terms used to
describe this behavior. Figure 3;8 illustrates the type of relation be-
tween strain and fluence for a constant applied stresé,.a. The material
responds immediately in an elastic mode,* then proceeds to undergo a satu-
‘rating primary creep superimposed on a linear secondéry creep. The primary
creep is essentielly a constant ?olume'creep and appesars to be reversible.
Since it saturates at fluences small compared to those of interest here,
it is valid to treat it as a non-time-dependent elastic strain. With this
simplification, the equations vhich must be sqlved take the form,

.gi - = [Gi - u‘(oJ-+ uk):] + % [0’1 --;" (UJ + Gk)]

v
+f k® [oi —%’.(GJ + ck)&] dat

 

 

o
-t :
+f g dt + (T - T_) | (3.5)
o ‘
vwhere -

€; = total strain in i-th direction (1, 3, k=1, 6, 2z),

oy = stress in i-th direction,

E = Young's modulus,

pu = Polsson's ratio,

k = secondary creep constant (irradiation-induced creep),

g = time rate of radiation-induced dimensional changes,

¢ = differential dimensional change due to thermel expansion;
To = reference temperature.

 

% .

Strictly speasking, graphite has no pure elastic mode, but behaves
inelastically unless prestressed. This detall does not affect the calcu-
lations given later.

 
"3

ok

 

0

 

"

23

ORNL-DWG 68-7976

 

 

 

 

 

 

1
_ SECONDARY
< CREEP
o
I._
w
w.
| '
A PRIMARY
{ CREEP
; ' ! ELASTIC
|
0 - 1078

4’1‘ FLUENCE

Fig. 3.8. Elastic and Creep Strains Induced in Uniaxlally Loaded Graphit.e'
Under Irradiation
 

 

a4

The right side of Eq. (3.5) sums the elastic strain, the saturated primary
creep strain, the secondary creep strain, the imposed radistion-induced
distortions, and the thermel strain. In addition to Eq. (3. 5) the follow-
ing must be satisfied:

€r=% ee=% e =2 | (3.6)

where u and w'are the displacements of the material in the r and z direc-

tions, respectively; also,

S(ro)=q
‘/P ro, dr = O

=0 . ' - (3.8)

(3.7)

and

The ebove re}ationships bave the following significance: Egs. (3.6)

preserve the‘continuity of the material during straining, Egs. (3.7)

define the conditions for static equilibrium within the material, and

Egs. (3.8) define static equilibrium at the free surfaces of the cylinder,
The sbove equations cannot be solved explicitly in closed form.

‘Approximate solutions can be obtained under the conditions Ek®t <<1 and

Ekdt >>1. However, it was possible to obtain numerical solutions to

the éomplete problem using a computer pro'blem10 originally designed to

study stresses developed in spherical coated particles and modifying

it to cylindrical gedmetry. The program uses &n iterative procedure

as follows:

 

. 107, W. Prados and T. G. Godfrey, Stretch, a Computer Program for
Predicting Coated-Particle Irradiation Behavior: Modification IV, ORNL-
T™M-212T7 (April, 1963).

 
 

 

 

 

i

)

l

25
A zero-order approximation is generated by replacing Eq. (3.5) with

€0 = % [oio - p(oio + ako)] + i:]-'- [Uio --i'- (cio + cko)] +h, (3.92)

where t

hy o~ £+ f gdt + a(T - To) (3.9b)

o

and fio’ the secondary creep strain, 1s set equal to a constant. Equations
(3.9) are solved for the 0,'s as functions of position and time, and a
first-order approximetion to fi is generated by setting

t
£ =a/ m[oio - -21-(030 + "ko)] at - (3.10)
Using this expression to replace £, in Eq. (3.9b) yieids values for
0497 ajl’ and Okqs such a process 1s repegted until convergence is
obtained. In general, convergence is achieved in two to three cycles.

-The material constants appropriaté for Gilso-carbon-based graphite
(presumsbly to be used for the first MSBR cores) are

1.7 x 10° psi

0.27

2.0 x 10 27 cm®+neut l.psi”?
6.2 x 1076 °¢™2 -

QR E W
N

Using the ebove values and procedures, the maximum stresses occur at the
surfaces of the graphite cylinder, and to within ebout 1% the axial and
tangentisl stresses are equal. Figure 3,9-gives the calculated eximl
stresses és a function!of axiai position for varicus times; 1t 1s apparent
that the maximum stresses occur at‘z/L ;,0.6. The behavior of the surface
stress at this point 1s.given in Fig. 3.10 ag & function of time. Two

points ere of immediate interest: the thermal stresses.initialiy intro-

duced as the reactor is brought tb povwer disappear in & matter of a few
weeks; further, the maximum stress occurs at the end of the graphite

life, T, and is approximately TOO psi. This is well below the anticipated
tensilerstrength of 5000 péi expected foriMSBR grephite.
 

 

 

- 600 |-

400

200

AXIAL STRESS o, AT SURFACE (psi)

Fig. 3.9. Axjal Surface Stress as a Function of Po

ORNL—-DWG 687977

 

TN

 

71

 

-/

b

\

 

I\\

/)//,/"

7
=

 

 

/

/
12

--_——_——_—'
6

 

 

 

 

 

 

0.2 0
Z/L, AXIAL POSITION

MSBR Graphite

4

0.6

08 1.0

sition and Time in

 
-

¥

)

o, AT SURFACE ( psi)

—-200

Fig. 3.10.

27

ORNL—DWG 68— 7978

 

800

 

600

 

400

 

200 A

 

 

 

 

 

 

 

 

 

¢ 6 42

18 24 30

TIME (months at 100 % plant factor )

Axial Surface Stress as
MSER Graphite

a Function of Time at z/L = 0.6 in
 

 

28

Of interest also is the overall dimensional change in graphite,
determined by | |
b -

g = —2 f_‘rgdr . (3.11)

ba_aaa v

Within the accuracy of the calculations, the distortions u and w at the

free surfaces are given by

ul, = ag ujy = bg V=28 .

Thus, the external dimensions of the graphite cylinder change according
to the average distortion rate g quite independently of the details going
on within the tube. Defining the graphite lifetime, T, as that which

gives a zero overall dimensional change,

 

T
f Edt=0at§=o.6
O

At time T the surfaces of the graphite at highest average exposure are
still in a slightly contracted state, while the Interior is in a slightly
expanded state. This criterion yields a value of T = 26.7 months at

100% plant factor.
The total relative change in length of the graphite cylinder as
a funcetion of time 1ls given by

L t | |
&1 o [ Fa a2
L L
O o

The assocliated results are given in Fig. 3.11; as shown, for the case
calculated, the core must accommodate & net 1.6% linear shrinkage of
the graphite colum.

Attentlon is now given to the second problem mentioned above, namely,' o .
the stresses assoclated with asymmetrical gradients. Denoting by R the
radial coordinate from the.centerline of the reactor. core toward the o
blanket regions, the flux ¢ will die away repidly as R approaches the <=J

blanket. Considering a graphite core cylinder near the blanket region,
 

 

)

o3

o«

1%

S |

 

29

ORNL~DWG 68~ 7979

 

 

 

-=—0.8 |

 

 

AL/L ,TOTAL AXIAL DISTORTION {%)

\

0 6 12 18 24 30
TIMEr(months at 100 % plant factor)

 

 

 

 

 

 

 

—1.6

Fig. 3.11. Total Axial Distor*cibn" of Center Line Tube in MSBR
 

 

30

the exterior surface facing toward the core centerline will be exposed
to a more intense flux than the exterior surface facing away from the
centerline. Specifically, if & is the average flux in the tube, the
surface facing the core centerline will be in & flux given by

3 b

'b-—-—
PSR’

and the surface facing the blanket will be in a flux given by

b0 .
3R

Referring back to Eq. (3.1),_the core flux gradient existing near the -
blanket region will tend to bow the tube concave inward during its con-
tracting phase, and convex inward during its expanding phase. The associ-
ated stresses which develop can be approximated in the foilowing way:
In its bowed condition the tube is eSsentially in a stress-free condition.
If it is constrained from bowing by adjacent tubes, then thesé ad jacent
tubes must produce distributed external stresées Jjust sufficiént to
straighten out the bowed tube. Thus, the problem reduces to & beam under
distributed external loading but undergoing creep, with the maximum
,St;esses being produced in the extreme'radial fibers. Let di'be the
radiation-induced distortion of the innermost fiber and 4, that of the
outermost fiber. Then the strain rate on the extreme fibers will be
given by |

,.él =%.'di"do-| - (3:139)

and the resulting fiber stress by

o) = del o (3.130)

ko

 

 

Flux gradients in radially power flattened cores suggest that

 
 

L]

o

|

#)

 

31

ég-< 2.9 x 102 peutrons cm™® sec”l

SR

at ® = 2.4 x 10'* neutrons cm 2 gec"?
For T ~ 700°C, this yields
4, = —8.10 x 10730 . 1.62 x 10717 ¢

and

d, =—7.29 x 10710 + 1,31 x 1077 ¢ .

Thus, )
€ = =041 x1072° 4+ 0.6 x 10737 ¢ .

Near the end of life (T = 1.0 x 10® sec) the stresses reach a maximm,

namely
l%L = 240 psi
ko

lczl = .
Such & value is relatively small, To this must be added the tensile stress
generated by the symmetric gradients occuring at the position of greatest
flux gradient; however, the latter would be less than the'value at the
core centerline. Thus, it is concluded that there are no serious thermal-
or radiation-induced stresses produced in the graphite during’thellife-
time essoclated with a . -

| | | f gat =0 ,

o

and that a net volumetric growth is pezmissible from the viewpoint of

rermissible stresses per se. Thus, & grephite lifetime associated with

. t_: G
J[\ gdt =0

o

implies that other factors, such as the influence of dimensional changes

on graphite permeebility, limit graphite exposufe.
 

 

 

3.3 Penetration of Graphite by Gases and Salts

W. H. Cook

3.3.1 Penetration by Gases

Numerous gaseous fission. products will be produced in molten-salt
breeder reactors,l! the worst being 125Xe from the viewpoint of neutron
absorptions. Ideally, the graphite shbuld be completely lmpermeable to
135ye, However, reasonably low values of the xenon fraction poisoning
(about 0.5%) can be dbtained'by stripping the xenon with helium bubbles
and/or'by using a graphite in which the diffusion rate of xenon is very
lov.

Two parameters are very importent in controlling the quantity of
xenon iesiding in the graphite at a given time. The first is the void
volume, since the amount of gas present 1s controlled by the space in |
which it can be accommodated. This void volume can be made low by |
multiple impregnations of the graphite during processing. The second
factor 1s the rate at which xenon can diffuse into the graphite, which
is controlléd'by the xenon concentration gradient and the properties
of the‘graphite. The accessible‘vbid volume 1s measured by use of
helium or kerosene, and the diffusion coefficient is obtalned from
pérmeability measurements with helium. Examination of gas transport
phenomena revéals that in graphite having very low penetration character-
istics, the permeability and diffusion coefficlents¥* are numerically
equal. This condition exists when the mean free path of the gaseous
molecules is greater than the diameter of the pores in the graphite,
.corresponding to the Knudsen flow conditions. The value of the diffusion

 

¥The dimensional quantity usually used for permeability coefficient
18 cm®/sec, while £t2/hr is used for the diffusion coefficient, and both
of these units are used here. Numericelly, they have the same order of
magnitude. Also, the Knudsen diffusion coefficient for xenon at 650°C
expressed in ft /hr is approximately equal numerically to that for helium
at 25°C expressed in cm®/sec.

1l¥. R. Grimes, MSR Program Semiann. Progr. Rept. July 31, 196h
ORNL-3708, p. 2LT.

 
 

 

o}

4

1

4

 

33

(or permeability) coefficient at which this equivalence holds is gener-
ally sbout 10~ % cm®/sec or less when the pores are small in size and-
nunmerous. For MSBR graphite, a gaseous diffusion coefficient of about
10"® £42/hr is desirable; for such a value, Knudsen flow conditions
would clearly apply. Under such circumstances, the relation between the
permeability and the Knudsen diffusion coefficient 1s'? (for steady
state conditions): |

 

L B
K =-qzz§ = -91-1{2 + D, (3.14)
where | | )
K = combined Knudsen-viscous permeability coefficient, cm®/sec,
D, = Knudsen diffusion coefficient = %-Kb v, cm®/sec,

q, = volume flow rate of gas measured at p , em®/see,

mean pressure'in porous medium, dynes/cma,'

e
n

: 1ength of porous medium in the direction of flow, cm,

>
i

il

cross sectional area for flow, cm®,

= pressure difference across sample, dynes/cm »
=rv1scous flow parameter for porous material, em®,
gas viscosity, poise, _

= Knudsen flow permeability coefficient, cm,

‘= mean molecular velocity, cm/sec = :\/ Qmfi ,
L

universal gas constant ergs/ K]mole,

ad =& 8
i}

vy
1l

temperature of gas, °K,

+3
"

M = molecular weight of gas, g/mole.

The value of K in the equation 1s easily determined experimentally
by measuring the volumetric flow of gases through a piece of material

 

laG F. Hewitt, "Gaseous Mass Transport Within Graphite " AERE-R-
467 (May, 196h)(Chapter Two, pp. T4-120 in Chemistry and Physics of

‘Carbon, Vol. 1, ed. by P, L. Walker, Jr., Marcel Dekker, New York, 1965);

E. A, Mason, A. P. Malinauskas, and R. B. Evans, III, J. Chem. Phys. h6(8)
3199-3216 (April 15, 1967); and R. C. Carman, Flow of Gas Through Porous
Media, Academic Press, Inc., Publishers, New York, 1956. _
 

3k

- under a differential pressure. The term B oP /fl represents’the viscous

coefficient and is a function of the average pressure and the gas vis-
cosity (laminar flow); the seeond term is the Knudsen diffusion coeffi-

' cient, 18714

‘Having determined DK-for a-giten set of experimental cohditions,
extrapolation to other conditions of interest can be made since the
Knudsen flow coefficient, K , is a function only of the porous medium.
Thus, through permesbility measurements of helium in graphite, the
diffusion coefficient of xenon in graphite can be celculated.

Methods for reducing void volumes and diffusion coefficients for
gases in graphite, as well as values associated-with these parameters,

are given in subsequent sections of this chapter.

3.3.2 Penetration by Salts

The efforts being made to obtain graphite with a low gas permeability
should yield a material with high resistance to penetration by salts.
The resistance to salt penetration into the graphite pores results from
the relatively high surface tensions of the molten salts such that they
do not wet graphite.j The molten fluoride salts at TO0°C have surface
tensions ebout 230 dynes/cm and a contact angle with graphitels of
approximately 150°. It is inherent that massive polycrystalline graphite
will have some accessible porosity, but the pore entrance'diemeters can
be'held reasonebly small, < 1 p. Therefore, if there is no preséure
differential between the helium-flilled pores and the salts, the saltis
should not intrude into the accessible pores gince they obey the Washburn

 

relationl® given by

 

133, F. Hewitt and E. W. Sharratt, Nature 198, 95l (1963).

14p, P, Malinauskas, J. L. Rutherford, and R. B. Evans, III, Gas
Transport in MSRE Moderator Graphite. 1. Review of Theory and Counter
Diffusion Experiments, ORNL-41L8 (September, 1967), Pp. 34-35.

15p, J. Kreyger, S. S. Kirslis, and F. F. Blankenship, Reactor Chem.
Div. Ann. Progr. Rept., ORNL-3591, pp. 38-39.

15H L. Ritter and L. C. Drake, Ind. Eng. Chem. Anal. Ed. 17(12),
782 (1945). B

 
 

u}

*

o

s}

i}

t}

 

35

- ooef (3.15)
vhere |
A@ = the pressure difference,
7= - the surface tension,
§ = the entrance diemeter of pores penetrated and
6 = the contact angle. '

Several observations support the applicability of this equation to
fluoride salt systems. Calculations indicate that a pressure difference
of approximately 300 psis would be required to start the intrusion of
fuel salt into the larger pore entrances (epproximately 0.4 p) of the
grade CGB graphite used in the MSRE. In out-of-pile standard salt-
screening tests in which a 165-psia pressure differentisl was applied to
a salt-CGB graphite system, the salt was limited to small penetrations
of the surface and to cracks which intersected exterior surfaces. In
the latter, the salt was confined to the crack and did not penetrate the
matrix.?? In-pile tests'® and the experience to dete with the MSREI®™2%
suggest that radietion does not alter the nonwetting characteristics of
the fuel salt to the graphite. Finally, the effects of compositional
differences in the fuel and blanket fluoride salts, of metal fission-
product deposition on the graphite, of fission product fluorides or
minor contamination of the salt do not appear to make important changes
in the nonwetting characteristic.22

 

3708 p. 38k.

 

1”w. H. Cook, MSR Program Semiann. Progr. Rept. July 31, 1964, ORNL~-

8 " 18SR Program Semiann. Progr. Rept. Feb. 28 1965 ORNL-3812, pp.
7-1200 . , ]

‘1935, S. Kirslis, MSR Program Semisnn. Progr. Rept. Aug. 31, 1966,
ORNL-h03T, p. 172-189 ,

 ®9g, s. Kirslis and F. F. Blankenship, MSR Program Semiann. Progr.
Rept Feb. 28 1967, 0RNL-h119, pp. 125 130. o |

215. S. Kirslis and F. F. Blankenship, MSR Program Semiann. Progr.

225. E. Beall W. L. Breazeale, and B. W. Kinyon, internal corre-
spondence of February 28, 1961,
 

 

 

 

36

The pressure difference:appears to be the controlling factor for
selt penetration as long as the wetting characteristics ere not altered.
The maximm anticipeted operating pressure of the fuel salt in the MSER
will be about 50 psig. The helium cover-gas pressure prior to filling
the reactor with fuel will be approximately 20 psia. Consequently, the
pressure will not be gble to force salt into graphite pores having
openings of 1 p. Steps being taken to reduce the gas permeability will
prdbably reduce the entrance diameters of the accessible pores to cone
siderably less then 1 e

There are no data at this time which suggest that the salt will
ever wet the graphite. However, if for some reason wetting occurred,
gome data'suggest'that pehetration'by a semiwetting or Wefting 1iquid
would be limited'by frictional effects®® and/or'by the pore strubturé
of the graphite involved. This should be particulariy'true for the type
of graphite sought for MSBR's becguse it should have very small pore
entrances, The friction concept has been referred to by E'arl:herffl..y'."2:3
This effect was illustrated by tests with molten sulfur, which wets
grephite. The sulfur penetrated only to an average depth of approxi-
mately 0.25 in. in & previously evacuated block of grade CGB graphite.2*

3.3.3 Pore Volume Sealing Techniques

A graphite which‘prevents salt and fission products from entering
is desired for improved neutron economy, as indicated previously.
Several techniques show promise for producing such a graphite. These
involve treatment of base-stock graphite by (1) impregnating with carbon-
aceous liquids that are carbonized and graphitized, (2) impregnating
with salts, (3) sealing with pyrolytic carbon or graphite, and (l4) seal-
ing with a chemical-vapor-deposited metal. All should be of some value
in 1limiting gaseous end liquid transport into the graphite; the latter
two appear the most promising for MSER application. . R

 

20y P. Eatherly et al., "Physical Properties of Grephite Materials
for Speciel Nuclesr Applications, Proceedings of the Second United
Nations International Conference on the Peaceful Uses Of Atomie:Energy
Geneva, 1958, Vol. T, pp. 389-401, United Nations, New York, 1959.

 

24MoR Program Semiann. Progr. Rept. Feb. 28, 1965, ORNL-3812, pp. T7-80.

 
 

)

»

"}

Ay

o}

¥

 

37

The base stock for all processes should have & narrow range of pore
entrance diameters 5 1l p. This pore structure is finer than that found
in most high-density grades of graphite. However, with proper grain
sizing, this type of base stock has already been fabricated by graphite

manufacturers.

Liguid Impregnations

Hydrocerbons. The clessical approach for reducing the porosity
and increasing the density of graphite has been to impregnate the base
stock with coal ter pitches that are subsequently cafbonized.and graphi-
tized.25 Recent work hés.used a variety of carbonaceous materials such
as thermosetting resins. During the pyrolysis of the impregnants, a
variety of gasés; primarily hydroéafbons, are driven off. The pore: spaces
created by these escaping gases will also be available to fission gases.
Also, these impregnanté,usualhy decrease appreciebly in volume during
pyrolysis and slightly during carbonization; so, the finel volume of the
impregnant does not completely fill or block voids. Since a graphite is
desired in which the gas flow is controlled by diffusion (Knudsen flow),
the hydrocarbon gases formed during pyrolysis must escape by the same
mechaniem. - Consequently, the carbonization cycle has to be long and
carefully controlled. 8palling end cracking are common fabrication
problems of such high-quality graphite. For example, the grade CGB
grephite with a nominal permesbility of 3 x 10™* cm®/sec developed tight
cracks during its final stages of fabrication because of the quality of
the sealing. Graham and Price reported only a 38.3% yield of fuel element
graphite for the first charge of the Dragon reactor,es even though a fine
carbon black, an amorphous carbon, wae used in the fabrication of their
base stock to give them a starting fine-pdreflétructure. We are not con-
sidering the use of amorphous carbon in the graphite for MSBR's until we
evaluate its aimensional stability under irradiation. | ’ '

 

- 3L, M. Curie, V. C. Hamister, and H. G. MacPherson, "The Production
and Properties of Graphite for Reactors," Proceedings of the First United
Nations International Conference on the Peaceful Uses of Atomic Energy,
Geneva, 1955, Vol. 8, pp. ¥51-473, United Nations, New York, 1956.
 

38

- Liquid impregnation has been, used to produce pieces of graphite
having very low fiermeabilities;26’27 permeability values reported have
been < 10°¢ cm®/sec. |

At this time, a permesbility of sbout 10~2 cmZ/sec appears to be
readily obtainable in fine-grained, high-denSity anisotropic or isotropic
graphite. As indicated above, decreasing this permesbility by hydrocarbon
Impregnation techniques becomes increasingly difficfilfi as permeebility is
decreased. The low permeabilities given above were for anisotropic grades
of grephite, but a large part of the associated technology should be use-
ful for the fabrication of low-permeEBility"iSOt:opic graphite. It would
be desirable to prodfice & structure which is uniform‘throughout, however,
it may be satisfactory to have a shallow surface impregnation plus

: graphitizing treatment.

 

Metals and Salts. Previously we emphasized the need for a premium
grade of base stbck. If metals or salts are used-as impregnants, hewever,
the restrictions on the fine-pore-diameter sPectrum of the base stock
could be relaxed. However, the impregnation of the pore volume with metals
is not being seriously considered for the MSER because it might introduce
intolereble quantities of nuclear poisons. At the same time, impregnating
graphite with selts such as LiF, CaFg, or LiaBeF4'is a possibility. ©Such
salts would not constitute intolerable nuclear poisons. The first two
would be solids, and the third would bé liguid at the reactor operating
temperatures. Althdugh not measured, it 1s probeble that the diffusion
rate of”uianium, other fuel-salt and blanket-salt components, and fission
products into the impregnant would be quite low,28

A small of work was done some years ago in which CaFp was used as an
impregnant. However, attendant experimental problems are_difficult,'
since'the fluoride salts are hygroscopic, and a graphite impregnated with

 

L. W. Graham and M. S, T. Price, "Speclal Grephite for the Dragon
Reactor Core, Atompraxis 11 549-544 (September-October 1965). -

K. Worth, Technique and Procedures for Evaluating Low Permeability
Graphite Properties for Reactor.Application, GA 3359 (March 1, 1963), P. T

 

aePrivate commmnications from R. B. Evans, III, of the Reactor .
Chemistry Division, who called our attention to this approaeh. '

 
)

N

0

 

39

- such salts would have to be protected from the atmosphere until installed

in the core and the core sealed. . .

Finally, there is the possibility of using counter diffusion of gases--
a concepted worked on for some time by the British. A counter flow of
helium cover gas from the graphite to the salt could help block diffusion
of 135ye and other gases into graphite. This method would also supply
helium bukbles to the core region to help remove 135Xe from the fuel salt.
However, such &an approach requires speclel core designs and gas flow

- through the graphite, and appears less desireble than the development of

improved graphite.

3.3. i Surface Coatings and Seals

In addition to using liquid hydrocarbon impregnants for dbtaining
improved graphites, a promising method involves sealing the graphite
surface by deposition of pyrolytic carbon (or graphite) or pure metals,

- Such a sealing method has been applied successfully to graphite to give

an. improved oxidation resistance. Much of this work has been associated
with rockets ‘and missile applications. The approach has been to apply &
coating on & massive substrate of porous graphite. Similar work has been
done on nuclear reactor graphite to decrease helium.permeability from
3.7 x 10™2 to less than 10”7 ?/sec.?g Coatings of carbides, oxides,
silicides, pure metals, pyrocarbon, and pyrographite have been investi-
gated. Not all were applied by the pyrokytic technique. The usual
problems.o were cracking of the coating or loss of the coatings because
of differences in rates of thermal expansion._ In some instances the

| graphite substrate was manufactured specifically to match the thermal ,

expansion of & particular coating.
A low-permeability pyrocarbon-graphite material has been reported
hy_Bochirolst_in vhich graphite was sealed with,pwrolytic.carbon_formed

 

29R, L. Pickerdike and A. R. G. Brown, "The Ges Impregnation of EY9
Graphite,” Nuclear Graphite, European Nuclear Energy Agency, Paris (1961),
rp. 109-128,

80, 7, Clarke, R. E. Woodley, end D. R. De Halas, "Ges-Graphite
Systems,” Nuclear Graphite, R, E. Nightingale (Ed ), Academic Prese, New
Yfll‘k 3 1962 2 pp . l"32 -E3T * .

2811,. Bochirol of CEA Seclay, France, personal commmnication.
 

 

 

Lo

from methane or a sulfur-free natural gas at 900°C. Such material, even
1f heat treated to 3000°C, may not be steble enough to radiation damage
for MSER applicafion because the crystallites are small, approximately
100 R.~ However, it does suggest that pyrocarbon can be deposited into

graphite substrate to a significant depth. The gross permeabilitieé

approached 10”7 cm®/sec as deposited, but were increased to 10°> cm®/sec
by a graphitizing heat treatment. Since the reduction in permeability of
the sample was obtained by sealing the surface, the gas diffusion coeffi-

cient essociated with the surface seal was much lower than the gfoss,

permeability coefficient, by the ratio of seal depth to sample thickness.
As indicated above, coatings or surface sealing can be employed.
Surface sealing, which injects the sealant a short distance into the pore
structure of the graphite, appears prefersble to minimize the effects of
radiation demage on the seal effectiveness. This type of sealant would
be more adherent-than & simple surface layer. | -
The surface sealant approach using pyrolytic carbon is in early
stages of study at ORNL.32 Pyrolytic carbon is deposited from propylene,
CsHe, on graphite specimens in fluidized beds at approximately 1100°C.
In one test the helium permeability of a graphite having two peaks in the
pore spectrum was decreased from approximatély'10'3*to'about'2'x 10”7
cm?/sec. This was the average permeability dbtained by considering the
graphite to be homogeheoué;'the permeebility of the materiai near the
surface vas estimated to be sbout 10™® cm®/sec. The carbon penetrated
the pbres as well as forming a surface layer approximately 15 u thick.

‘The low permeability was maintained when the semple was heated to 3000°C

and cooled to room temperature. Additionsl vork on surface sealing 1s in
progress using an isotropic graphite that has a narrow'rhnge of pore
sizes with entrance dismeters near 1 . Specimens of this materiel have
been sealed and irradiated to high reactor exposures in the High Flux

- Isotope Reactor (about 1022 nvt), but the results have not yet been

évalfiated._

 

52, Beutler, MSR Semiann, Progr. Rept. Aug. 31, 1967, ORNL-4191,

 

 
 

 

 

"

"

)

%

L1

The metallic surface sealing studies carried out at ORNL involve
use of molybdenum or nicbium.®® The metal is deposited on a heated graph-
ite substrate by reducing the metal halide with hydrogen. Initial re-

 sults have shown that a molybdenum coating approximately 0.05-mil thick

decreased the permeability'of a porous, molded graphite sample from
epproximately 10”2 to 1076 cmZ/sec; the permesbility of the coating
itself would be much lower. The coating maintained its integrity during
thermal cycling, and more extensive testing is‘planned. .

3. Near-Term Industrial Production Capabiligy

- W. P Eatherly

Discussions have been held with several vendors on the possibility
of producing from Gilso-carbon flour an isotropic graphite meeting the

- initial MSBR requirements and having the radietionAhehavior character-

istics of the British graphite._ Two vendors have made Gilso-base material
into large blocks having the above radiation characteristics, the blocks,

»however, have a coarse-grained ‘structure which would not meet the perme-

ability requirements of the MSBR. Both vendors also have active programs
aimed at producing fine-grained materials, and one vendor has made a
production run on tubing approximately l in. OD.

Production equipment was exhibited by one vendor which is capeble
of producing tubing up to 15 £t in length, with processing parameters
appropriate to Gilso-carbon flours, i’law-free structure ’ and low _perme-

ability. Several vendors have expressed their confidence in ‘being sable

‘to produce the required material on & firm price basis in from'18 to 2h

months. ) _
| It appears that at least two vendors would be eble to produce a
material which would be useable in an MSER. Producing this material re-
quires 1ittle extension of existing technology, and the uncertainties ley
mostly in the region of processing yields and cycle times rather than in
basic productlformulation‘or process.

 

33W. C. ‘Robinson, Jr., MSR Semiann. Progr. Rept. Aug. 31, 1967,
onmL-higi.
 

 

 

 

L2

Thus, an lsotropic graphifie capable of operating up to an MSBR
dose of about 3 x 1022.neutrbns/cm? (E > 50 kev) appears available with
moderate extensions of existing technology. The base material would
probably have & helium permeebility of sbout 10 ° cm®/sec, and it eppears
that pyrolytic carbon can.be ueed to seal the surface. Present york
indicates that the surface'of grephite can be sealed to obtain & surface
permeability of a‘bout'lo'9 cm?/sec; the techniques presently being used
can be scaled up to seal MSBR-size tubes. However, additional work may
be required in order to develop a seal which is resistant to radiation
damage. | | - |

| L. FISSION PRODUCT BEHAVIOR IN MOLTEN-SALT REACTOR SYSTEMS |
S. S. Kirslis
The removal of fission products from the resctor core is required in

MSER systems in order to attain good fuel utilization performance. The
ability to continuwously remove such nuclides is dependent_upon their Bé—
havior in reactor enviromments and, in particular, upon the reténtion
characteristics of graphite for fission products. In this chapter the
behavior of important fission products in molten-salt-graphite—metal
systems is considered; fission gases such as 125Xe, however, are treated
‘more specifically in Chapter 5. |

 In order to use unclad graphite in direct contact with fissioning
molten fluorides, some rather stringent chemical éompatibility'requife-
ments must be met. First, there must be no destructive chemical reaction
between graphite and the fuel salt with its contained fission products.
Second, the fuel must not wet the graphite surface since this would lead
to permeation of the graphite pores by bulk fuel end also fiséion'products.
Third, individual fission products of appreciable CIroOSE section mist not
leave the salt phase and accumulate on the graphite surface or penetrate
into the graphite interior to a degree which significantly affects the
neutron economy of & breeder reactor. This chapter summarizes recent
experimental information on fission‘product behaviqr in MSR systems.

 
 

 

 

 

)

n

N

&

L3

4,1 In-Pile Capsule Tests

 

In-pile capsule tests carried out early in the MSRE program showed
that there was no significant chemical damage to graphite in contact with
fissioning molten salt under reactor operating conditions. There were
compatibility problems only when the molten fuel was allowed to freeze
and ‘cool below 100°C during the course of the experimental measurements.
Undér these'conditions'the solid fuel'was radiolyzed by the fission
product radiations, yielding elemental fluorine and reduced species in
the salt. A final in-pile capsule test. (ORNLJMTR h7-6) showed no graphite

damage ‘and no uranium deposition when the fuel was not allowed to freeze.

fCover-gas samples taken during this test showed no Fa or CFs generation

from the irradiated capsules. There was also no permeation of fuel salt

into the graphite in the final test nor even in the previous tests where

some fuel radiolysis occurred.

No detailed observations on fission product behavior were made in

jthese early tests. However, there were 1ndications that 1°3Ru and 1°6Ru

deposited on the submerged metal and graphite surfaces and some evidence

that 31T and 12Te deposited on the capsule_walls above the liquid level

and on the walls of the cover-gas lines.- :

b, 2 Exposure Tests in the MSRE Core

More detgiled studies of the 1nteraction of graphite with fiss10ning
molten salt were carried out in the MSRE reactor environment. A 5-ft-long
test assembly of graphite and Hastelloy N specimens, shown in Fig. k.1,
was exposed to circulating fuel salt in a central position of the reactor
core for T800O Mwhr of: reactor operation. A second similar assembly was
exposed subsequently for 24,000 Mwhr of reactor operation. These aSsefihlies-
were removed from the reactor, dismantled in a hot cell, and the specimens
subJected to a series of examinations and analyses.

Three rectangular graphite bars were selected from each assembly for
examination, these bars being taken from the top, middle, and bottom parts
of the core. Adjacent Hastelloy N specimens were cut from the perforated

metal basket surrounding each specimen assembly. Visually, the graphite

specimens appeared undamaged except for occasional bruises incurred

during the dismantling. Metallographic examination showed no radiation
 

 

 

PHOTO 81671

7 PR

 

 

 

 

 

 

Fig. 4k.1. Hastelloy N and Grade CGB Graphite Surveillance Specimens

and Container Basket. (a) Specimens partially inserted into
the container. (b) Container and its lock assemblies.
(¢) Location of surveillance specimens in the MSRE,

i

 
 

 

 

+

»

”

A

o

N

k5

‘or chemical demage to the graphite structure and no evidence of surface
- films. X-radiography of thin transverse slices showed occasional salt

penetration into previously existing cracks which extended into the speci-
men surface. This penetration probebly accounted for the slight gein in
velght (~13 mg out of about 30 g). Similer penetration was observed in
the control specimens vhich were exposed to molten salt in the absence

of radiation. No new cracks were caused by the exposure to radiation.

A suggestion?of a very thin leyer of denser_mafierial on the graphite

‘surface exposed to salt was visiblq_in.the x-radiographs of the irredi-

ated and the control specimens. X-ray diffraction analyses of the graphite
surface exposed to fuel showed & normal graphite pattern, with a very
slightly expanded lattice spacing. A few very weak foreign lines, probebly
due to fuel salt,'weréldbserved.' Autoradiography of the graphite specimens
showed & high concentration of activity within 10 mils of the surface,

with diffuse irregular penetratiofis to the center of the specimens (the
resolution of these measurements was sbout 10 mils). An eleétxon prdbe
examination of the graphite epecimene (carried out at Argohng Nationel
Leboratory) detected no impurities in the graphite at or near the -surface
exposed to fuel, with detection limits of 0.04 wt % for fission products
and 0.02 wt % for uranium. These series of observations, based on samples
having T900- andlah,OOQAMWhr,reactor exposures, indicated satisfactory’
compatibility,of_graphite with fissioning molten salt relative to damage
by chemical reaction end to permeation of bulk fuel into graphite.

The three rectangular graphite bars from each of the two MSRE runs
vere also used to study fission product deposition on graphite in, more
detall. Thin 1ayers of graphite, 1 to 10 mils thick, were milled from
the flat surfaces of the bars to e final depth of sbout 50 mile. These

jsamples-weré.dissolvedfand.analyzed radiochemicaliy.--The predominant

activities found deposited on and in the graphite were the 1sotopes of

- molybdenum, tellurium, ruthenium, end niobium. These elements may be

clessed as noble metals since thelr fluorides are relatively unstabdble.
Their‘dep091tion on graphite is,ofupractiéal concern since—severalJiso-
topes in this class (in particulsr, %Mo, ®7Mo, ®%Tc, and 392Ru) have
relatively high neutron cross sections; if the total fission yields of
these fission producte were retained in the graphite core, the long-term
.active isotope of the same element or that of & radioactive noble-metal

 

 

 

L6

neutron economy of an MSBR would be adversely affected. It is difficult
to snalyze directly for these stable or long-lived species; 1t was assumed
that their deposition behavior was indiceted either by that of a redio-
presursor of appreciable half-life. oo ‘
-~Analyses of the milled graphite samples showed that over 99% of the
deposited noble-metal setivities were concentrated within 5 mils of the
graphite sfirfacee; Conversely, the daughters-of the kryptons and xenons
were more uniformly distributed throughout the g:aphite specimens with

shorter 1ived rare gases having steeper concentration gradients through

the graphite (as expected). Elements with steble fluoridee and no gaseous

' precursors (Zr, rare earths) showed low surface concentrations end were
- @bsent from the interior of the graphite. ' |

Relatively heavy deposits of noble-metal fission products were ob-
served on the Hastelloy N specimens adjacent to the graphite samples.
The deposits of other fission producfs:on Hastelloy N’were'relatively
light. The deposition of noble metal fission preducts'en'flastelloy N

“and graphite can be quantitatively described.in terms of the fraction of

the total fission products produced during reactor operation which was
deposited. It was essumed that deposition on the specimens is repre-
sentative of depesitioq on all the reactor graphite and Hastelloy N
surfaces in the MSRE system. On this basis, 14% of the ®°Mo, 13% of
the 1321e, 9% of the 1°3Ru, and 45% of the ®SNb produced during the first
7800 Mwhr of MSRE operation deposited on the graphite core. During-the
same period, 4% of the %Mo, nearly sll the 32Te, and 23% of the 1°%Ru
produced deposited on the metal surfaces.

The deposition of fission products on graphite and metal after sbout

'32,000{Mwhr of MSRE operation is shown in Teble k.1 as percentages of the

totael of each species'generated in the reactor system. The results are
again based on the assumption that deposition on specimens is represen-
tetive'of all surface deposits. The relative activities of'ggMo,'lszTe,
and °3Ru found on the grephite and metal specimens were sbout the same
as those found after the first 7800 Mwhr; however, the relative activity
of ®°Nb was distinctly higher after the second expesure. Aleo, after
the 2k,000-Mwhr exposure,-the‘ratio effgsflb deposited per cm® on metal

 
 

 

»

p

9w

-

h

T

Table 4.1. Approximate Fission Product Distribution in
: MSRE After 32,000 Mvhr of Operation

 

 

Isotope % in Ffiel % on % on Cover Gas™
Graphite Hastelloy N (%)
%Mo 0.9% 10.9 © k0.5 T7
132pe 0.83 110.0 70.0 66
193Ru 0.13 6.6 1.9 40
1 0.04k 36.4 34.1 5.7
957y 9.1 0.03  0.06 0.1k
89gr T7.0 0.26 33
1311 64,0 | - 1.0 16

 

®The figures in this colum represent the percentage of
the daily generation rate lost to the cover gas per day. The
sum of all columms does not add to 100% because of time vari-
ations in behavior, nonuniform concentrations 1n the gas phase,
and analytical inaccuracies.

to that on graphite was about 2 on the average. The correéponding ratio
was 8 for %Mo, 1k for 1321re, and 4 for 103R),__each somevhat higher than
for the 7800-Mwhr‘exposure. It had been expected that the ratio would
fall toward unity as both graphite and metal became coated with noble
metals, but this apparently did not occur. In the preéeht MSBR designs
the ratio of metal surface to .graphite surface ‘is about 1.5 to 1, rather
than 1 to 2 as in the MSRE, Thus, based on ‘these test results, only s
small percentage of the noble metal- f1551on products should deposit on

the graphite in the MSER core.

4,3 Tests in the MSRE Pump Bowl
The'behavidr.of fission products was further investigated by means
of test samples from the MSRE pump bOWl."ACCESS.tO the fuel salt and
the cover gas is provided by the salt sampllng facility shown in Fig.

4.,2. Samples were taken of fuel salt and of the helium cover gas; in
 

48

ORNL-DWG 67-10766R

 

 

12 in.
LATCH
STOP.
’ FINAL CLOSURE
_ WELD
| NORMAL OPERATING -~ A
: SALT LEVEL ~
CAGE FOR

SAMPLE CAPSULE

j SALT INTAKE SLOT

Fig. 4.2, Salt

LATCH ASSEMBLY CABLE SHEARED OFF
Z APPROXIMATELY AT THIS LOCATION

" CABLE

 

  

LATCH ASSEMBLY NORMAL POSITION

 

L L L & & &y S S S A S A A SN N N RS S N S s S STSsS IS Y

   

   

TOP OF FUEL PUMP

 

   

/s

 

 

   
 

 

 

 

 

 

 

 

 

 

 

PO DL IVEMENT SAMPLE CAPSULE
SAMPLE CAPSULE
CABLE
e b A
1| L
,/j/
L/
CAPSULE CAGE
BAFFLE
I SECTION A-A
| A
o SAMPLE CAPSULE
? ==,
{1 A
o
“\ ‘\

- E3/|5 in. MAX
i 0 1 2 3 4
< l INCHES

 

Sampling Capsule in MSRE Fuel Pump Bowl

 

 
 

 

¢

»

»

*

»

k9

eddition, metal snd graphite specimens were exposed to the fuel salt and
to the cover gas. , ' :
. Early fuel-salt samples, taken in open copper ladles, were found to

‘be highly contamdnated with noble-metal sctivities because the open

sampling ladles passed through the cover-gas region. Contamination'was
avoided in later fuel samples by sampling into an evacuated cepsule pro-
vided with & freeze valva which melted when the capsule was lowered into
the molten salt. The later resuits showed that less then 1% of the noble-
metal nuclides produced remain in the fuel-salt phase; species-with stable
fluorides (zr, alkaline earths, rare earths), however, remained predomi—_
nately in the fuel. : '

It was further found that high concentrations of noble-metal fission
products existed in the MSRE cover-gas volume. The metal specimens ex-
posed to -the cover'gas_picked'up activities associated with noble metals
several times that contained in & gram of fuel salt. The ‘fundamentals of
why these materials transfervénd remain in the gas phase are not fully
understood; however, inert ges flow may prove to be an effective way to
remove significent fractions of fissiofi products, and this action may
account for the relative decrease in fission product deposition on graph-
ite with time, which is discussed below.

In aenother test, sets of graphite and Hastelloy specimens in the pump
bowl were exposed to the gas phase and to the fuel phase for 8 hr during
full power reactor operation. Within & factor of‘ten,.fhé same amount of

;each'nuclide deposited on all the-specimens independent of location.
The deposition'of.ndble metals on Hastelloy N in this test appeared to
proceed at the same constant rate in the 8-hr run es in the 24,000-Mwhr

(3340-hr) exposure in the MSRE core. However, the average deposition

- rates of noble metals on:graphité were gbout & factor of ten lower in the

3340-hr exposure than in the 8-hr test, except for 9SNb, where the factor
was;about-l.s; ;This'could-1ndicate-that the deposlition rate of noble =
metals (except ®SNb) on grephite decremses with exposure time, which is-
an advantage from the viewpoint of neutron economy. However, results to
dete should be treated &s preliminary, and further investigations are
needed. Semples of the gas from the MSRE pump bowl indicated thet the
helium cover gas contained sbout 5 ppm by mole of *?Mo (i.e., 5 moles

 
 

 

 

 

50

of Mo per 10° moles of helium) and 1-2 ppm each of ®2pe, 19%py, 1%Ry,
and ®5Nb. If these concentrations are present in the gas leaving the |
pump bowl and are'multiplied by the flow of helium through the pump bowl
(6000 liters/day), the losses of °Mo and *®2re to the cover ges are those
given in Table 4.1. As shown, these calculated losses are appreciable
fractions of the generation rate of these species in the MSRE.

h h Chemical State of Ndble—Metal Fission Products |

The results above indicate that the noble-metal fission products
rapidly leave the fuel-salt phase by depositing on solid surfaces and by
entering the cover-gas volume. In order to help determine the mechanisms
of volatilization, two hot-cell tests were carried out. These tests
involved passing helium or & helium-hydrogen mixture either over or
through a fuel sample from the MSRE. It was found that passage of hydro-
gen gas had no effect on fission prbduct vblatilization, which indicates
that the voletile species of the noble metals were not high-valent gaseous
fluorides. Some salt mist was swept from the Sampie, but the concentrations
of noble metals volatilized were one to three orders of their concentration
(if uniform) in the salt. Further, significant amounts of noble-metal
fission products were swept from the fuel sample by gasrpassage elther
over or throuéh the molten sample.  The amounts of activity were the
same whether or not the gas contained hydrogen, indicating that these
"noble" fission products were present in metallic form. It was also
found that about 20% of the volatile noble metals passed through a filter
vhich held back all particles larger than 4 microns. These results suggest
that noble-metal fission products are injected into the gas phase as tiny
metal partiecles and form stable gaseouS‘Euspensions. '

4.5 Results from ORR Loqp Experiments

In addition to the study of- Pission product behavior in the MSRE,
fuel-salt-material tests have also been carried out with thermal con-
vection loops containing fuel salt and graphite. These loops were operated
~ in the Oak Ridge Research Reactor (ORR) to investigate fuel-behevior at
high power densities. The firstlloop experiment was terminated after
generation of 1.1 x 108 fissions/cc (0.27% 225U burnup) because of

 
 

®

p

"

 

51

g8 break in a sample line. A second loop operated at an average fuel power
density of 165 w/cc until & line leading from the "core" cracked; a fuel
dose of épproximately 8 x 10'® fissions/cm® was achieved. The test arrange-
ment employed in these runs is indicated in Fig. 4.3, The "core" in these
loop tests consisted of a 2-in.-~diam by 6-in.-long cylinder of graphite
(from MSRE stock). Vertical holes were bored through the graphite for
salt flow. A‘horifiontal gas separation tank cbnnected the top'of the
core to & return line (cold leg) which, in turn, was connected to the
bottom of the core;-completing'the loop. A fluid'flow rate of 30 to 50
ce/min (a2 min circuit time) was maintained at e "core" temperature of
ebout 650°C. B |

The surfaces‘in the secbndiloop were snalyzed thoroughly to determine
the deposition of fission pfoducts. This leyers were machined from the
core graphite surfaces, and these layers were analyzed to determine .the
concentration profile of the fission producfs within the graphite. The
results obtained for noble-metal fission products resembled»verj closely
those given gbove for the MSRE surveillance specimens. For‘reasons that
are not clear, the salt seemed tp have wet the grephite and penetrated
to a distance Qf a few mils. This‘apparently was caused by the presence
of a emall amount of,watér vapor. No such wétting behavior has been
observed during MSRE operations.

4.6 Eveluation of Results

A principal_intéractidn between graphite end fissioning molten salt
appears to be the partial deposition of noble metals on graphite; We
infer from the results that the percentage of noble-metal fission pro-
ducts deposited on graphite depends on the ratio of graphite surface to
metsl surface, with deposition-décreasing with decreasing ratio of
grephite-to-metal surface. Finally, test results indicate that signifi-

_cant fractions of noble-metal fission products can be present in the gas

phase. Such behavior could provide & convenient means for their rapid
removal from MSER systems. Experimental studies are continuing in order

~ to verify the present indications.

 
 

ORNL—DWG 66—965

SALT FLOW PASSAGE

PRESSURE MONITOR LINE

HEATER

 

 

 

SALT RESERVOIR

52

 

 

GRAPHITE CORE

 

THERMOCOUPLE WELL

 

 

 

 

 

 

 

 

 

V77
i
I
|
i
|
1
|
i
i
t
}

- ————

 

 

 

 

 

 

 

L L

 

 

 

 

 

 

 

 

 

 

 

 

 

 

HEATER

COOLING JACKET
COOLING COIL

 

 

COLD LEG

 

w
=
3
uy
-
&
=
g
»
5
g
»

 

|
|
\
D
7
;: .\';:‘\“\
THERMOCOUPLE

40
7

— e = o e
_ o .

GAS SAMPLE LINE
THERMOCOUPLE
WELL
COOLER—3
HEATER

 

In-Pile Molten-Salt Convection Loop No. 1

Fig. 4.3.
 

 

9

P

n

"

iy

"

53

5. NOBLE-GAS BEHAVIOR IN THE MSER
R. J. Kedl Dunlep Scott
As pointed out previously, the graphite in the MSER core 1s uncled

and in intimate contect with fuel salt. Thus, noble gases generated by
fission and eny other gaseous compounds may diffuse into its porous struc-

 

ture where they can act as heat sources and neutron poisons. Although
fission products other then xenon are involvéd, the greatest gain can be
made by removing 3°Xe, and later discussions refer primarily to *°Xe
poisoning.

In order to estimate neutron poisoning effects, a steady-state
analytical model waé developed to estimate the transfer of noble gases
in the MSER to the graphite. The verious factors considered included

. decay, burnup, migration into grephite, and migration to circuleting gas

bubbles. Gas generation direct from fission and generation from decay

- of gas precursors wefe considéredras.SOurce.terms. " The mgdel utilizes

éonventional mass transfer concepts and is used to compute nuclide con-
centrations and 235Xe poison fractions. The steady-state model for the
MSRE is developed in reference 3k, while the time-dependent model is
given in references 35;38 When'appliedfto very short-lived ndble'gases,

‘the model has given calculated results in agreement with MSRE values®®

measured under reactor operating conditions.

 

34R. J. Kbdl and A. Hbutzeel, Develqpment of a Model for Computing
13576 migration in the MSKE, omu.-hosg (June 1967).

“‘SMSR Program Semienn. Progr. Rept. _Feb. 28 1966, 0RNL-3936
36MsR Program Semienn. Progr. Rept. Avg. 31, 1966, ORNL-LO3T.
- STMSR Progr&m Sem:lann. Progr. Rept. Feb. 28, 196'[, ORNL-L4119.

281, R. Engel end B. E. Prince, The Reactivity Balance in the MSRE,
ORNL-TM-1796 (March 1967).

39R. J. Kedl, A Model for Comzting the Migration of Very Short-Lived
Noble Gases into MSRE Graphite, ORNL-TM-1010 (July 1967).
 

 

 

54

Using a model similsr to that indicated above, steady-state 135Xe
poisoning calculations were made for a modular two-fluid MSBR /556 Mw(t)/
to show the influence of several design parameters on xenon poisoning.
The reactor design concept considered here is essentially thet described
in reference 40; design perameters pertinent to Xe poisoning are given in

‘Teble 5.1. Xenon stripping from the fuel salt islaccompliehed by circu-
| lating helium bubbles with the salt; the bubbles are injected near the

pump at the inlet to the heat exchanger. Xenon-135 is considered to
migrate to the bubbles by mase transfer, with the mass transfer coefficient
controlling the rate of migration. The circulating bubbles are then ‘
stripped from the salt by a fiipeline gae separator located near the heat
exchanger outlet. | '

- With regard,to mass traflsfer of xenon to the graphite, the prinecipsl
parameters considered were the diffusion coefficient of xenon in graphite,
the mass transfer coefficients and areas associated with the circulating
bubbles, the time that bubbles are in contact with the salt, and the surf-
ace area of graphite exposed to salt in the core.

In Fig. 5.1 the diffusion coefficient of xenon in graphite at 1200°F
(650°C) is given in units of £t2/hr. As mentioned in Chapter 3, the | ]
numerical value of this coefficient in ft2/hr is sbout equal to the
more commonly quoted permeability of He in graphite gt room temperature
with units of cm /sec, if Knudsen flow prevails. Knudsen flow should
dominate for permeabilities < 10~ % cm®/sec.

The gas bubbles circulating through the fuel system were considered
to be made up of two groups of bubbles. The first group, referred to a8
the "once-through" bubbles, were injected at the bubble generator and
removed with 100% efficiency by the gas separator. The seccnd group;
referred to as the recirculated" bubbles, were also injected at the
bubble generator but completely bypassed the gas separator on their
first pass; it was assumed that bubbles in the second group were removed
with 100% efficiency on their second pass through the gas separator. , »
The particular parameter used to indicate the amount of circulatlng

bubbles was the bubble surface ares; for orientation purposes, note | -

 

4%paul R. Kasten, E. S. Bettis, Roy C. Rdbertson, Deslign Studies of
1000-Mv(e) Molten-Salt Breeder Reactors, ORNL-3996 (August 1966).
 

 

 

*

h

55

Table 5.1. MSER Design Parameters Used in Estimating
35Xe Poison Fraction®

 

Reactor power /Mw(t)/ 556
Fuel | £33y
Fuel salt flow rate (£t3/sec) | 25.0
Core diameter (ft) | | 8.0
Core height (ft) 10.0
Volume fuel salt in core (f£t3) | 83.0
Volume fuel salt in stripper reglon-heat exchanger (£4°) - 83.0
Volume fuel salt in piping between core and heat | 64.0
exchanger (ft

Fuel cell cross section

3-7/8-in. holes —_

 

Total graphite surface ares exposed to salt (£t2). - 3627
Mass transfer coefficient to graphite, upfiow (f£t/hr) 0.72
Mass transfer coefficient to graphite, downstream (ft/nr)  0.66
Mean thermal flux (neutrons/sec en®) 5.0 x 1014
Meen fest flux (neutrons/sec cm?@) 7.6 x 101%
Thermal neutron cross section for 233U (barms) a252.7
Fast neutron cross section for 2%y (barns) | 36.5
Total core volume, graphite and salt (£t3) | 502.6
83y concentration in core, homogenized (atoms/barn-cm) 1.11 x 10”5
Graphite“vbid‘afiailablé to'xenoh (%), o 10
Xenon-135 parameters S fli |
Decay constant (1/hr) | | 7.53‘x 1072
Generation direct from fission (%) - - 0.32
Generation from iodine decay (%) = - - 6.38
Crose section for MSER neutron spéctrum (varns) 9.94% x 10°

Nominel core power density (kw/liter) | | ko

 

SThe parameter values given should be considered ag representative
values; they would vary with MSER design conditions.

 

 

 
 

 

56

ORNL — DWG 68— 7980

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107

1078

1 -3
DIFFUSION COEFFICIENT OF Xe IN GRAPHITE

1073 104

1072

AT 1200 °F {12/ hr)

1 :
Fig. 5.1. Effect of Diffusion Coefficient in Graphite on T5’Xe Poison

Fraction

 
 

 

9

5T

that 3000 £t2 of bubdble surface area corresponds to an average vold
fraction of 1% in the stripper region of the fuel loop with bubbles
0.020 in. in diameter, vhen the gas flow rate is about 40 scfm.
Figure 5.1 shows the xenon-135 poison fraction as a function of
the diffusion coefficient in graphite with other parameters having the
values specified. The top curve in the figure is for no circulating
bubbles. The other curves consider that about 10% of the bubbles re-

circulate. From Fig. 5.1 it appears that the xenon polson fraction 1is

not a strong ‘function of the diffusion coefficient when it ranges from
10~ to 107€ £t®/hr. Thus, for these values of the diffusion coefficient,
the mass transfer coefficlent from salt to graphite is the controiling
resistance for migration of 335Xe into the graphite. The mass transfer
coefficients from salt to graphite were computed from the Dittus-Boelter

 equation. as modified by the heat-mass-transfer analogy. Since 125Xe

in the graphite is the greatest contributor to the total neutron poison
fraction, the parameters that control xenon migration will, in turn,
control the poison fraction. For diffusion coefficients less than 10°€
ftayhr, the resistence to xenon diffusion in graphite starts becoming
significant.

Figure 5.2 shows the effect on pbison fraction of the xenon mass
transfer coefficient from selt to helium bubbles. This mass transfer
coefficient is one of the least well known parameters and can be a most
significant factor. Avallable information indicétes its value to lie

between 0.7 end 6 ft/hr, vith & value of 2-4 ft/hr appearing reasonsble
 to expect. Values of ebout 0.7 — 0.8 £t/hr were estimated, assuming that
- the buhbles behave as solid spheres having a fluid dynamic boundary

layer. Values of about 3. 5 ft/hr vere estimated on the basis that the
interface of bubbles is continually being replaced by fresh fiuid
(penetration theory). Both of these cases consider a bubble rising at

1ts terminal veiocity in & stagnant £luid. Theré 1s very little infor-

mation in the literature concerning the effect of fluid turbulence on

the bubble mass transfer coefficient, but from‘turbulencé theory it

hae been pOstnlated that, under MSER conditions, mass transfer éoeffiqients
of 6 ft/hr or more could be realized. The analyses that lead to such
values are generally optinmistic in their assumptions. Figure 5.2 also

 
 

 

 

58

ORNL — DWG 68—7981 -

 

 

 

 

 

5
4
DIFFUSION COEFFICIENT OF Xe
IN GRAPHITE = 10752/ pr
3

 

 

135¢e POISON FRACTION (%)

 

SURFACE AREA OF "ONCE THRU" BUBBLES (ft2)
SURFACE AREA OF "RECIRCULATING" BUBBLES ($1%)

RN

I/
|
|

 

 

 

 

 

 

 

 

 

 

\“ \
3000 0
——— 3000 300
- 3000  60C
0
0 1 2 3 4 5 6
MASS TRANSFER COEFFICIENT TO BUBBLES {ft/hr)
! . l
Fig. 5.2. Effect of Bubble Mass Transfer Coefficient on 35Xe Poison
Fraction

 
 

 

»

y

 

29

indicates that & small amount of recirculating bubbles is as effective
as a large amount of once-through bubbles in reducing xenon polsoning;
thie result is due to the increased contact time for "yrecirculating"
bubbles reletive to "once-through" bubbles.

Anothér'variable that will strongly affect the xenon poison fraction
is the graphite surface area in the core. Calculations indicate that if
the graphite surface area were doubled, all other pérameters'remaining
constant, the poison fraction would increase by 50-70%.

The target poison fraction for the MSER is 0.5%. Referring to Fig.
5.2, if the bubble mass transfer coefficient were 4-6 ft/hr, gas removal
in itself appears to be a feasible method for attalning low xenon poison

fractions. If, however, the bubble mass transfer coefficient were 2-3

ft/hr or less, it appears that the target polison fraction ie not attain-
able under the specified conditions. Under the letter case, alternative
méthods for reducing xenon poisoning are to develop graphite having a
very low gaseous diffusion coefficient (Fig. 5.1 indicates & value of
1078 ftEth would be satisfactory), or to coat the bulk graphite with a
thin leyer of graphite having a very low permeablility.

) Calculations fiere.performed to determine the effectiveness of low-
permeability graphite coatings on xenon poisoning; Fig. 5.3 gives the
results obtained along with the parameter values used in the computations.
Tt vas assumed that for a coating of the indicated thickness, the specified
diffusivity and availsble void would apply to all graphite surfaces
exposed to fuel salt. The various xenon migration farameters vere chosen
to yield a *®5Xe poison fraction of 2.25% with no coating, so that Fig.

5.3 indicates the effect of coating parameters relative to this poison
fraction.‘_It was assumed that'thé“avaiiable void fraction in the graphite
coating decreased by one order of magnitude when the diffusion-coefficient
decreased by two orders of magnitude, vhich i & conservative assumption
relative to experimental results. As shown in Fig.'5.3, it appears that

a coating 10 mils thick and having a aiffusivity of sbout 107 £t3/hr

~ and en availsble void of approximately 0.3% would bring the **Xe poison

fraction down to the target value. A diffusion coefficient of 10™° £t2/hr
would require a coating thickness of only one mil. As stated in Chapter 3,
graphite coatings having the above characteristics have been produced, and

 
 

 

 

60

ORNL — DWG 68— 7982
PARAMETERS

CORE POWER DENSITY & 20 kW / liter

REACTOR POWER = 556 MWt '

BULK GRAPHITE DIFFUSION COEFFICIENT =40 >f12/hr
BULK GRAPHITE AVAILABLE VOID =10 %

MASS TRANSFER COEFFICIENT TO BUBBLES= 2 ft/hr

BUBBLE SURFACE AREA = 3000 12 (NO RECIRCULATING
BUBBLES )

FUEL CHANNEL GEOMETRY = CONCENTRIC ANNULUS

 

 

35xe DIFFUSION COEFFICIENT IN GRAPHITE COATING (f12/ hr)

AVAILABLE VOID IN GRAPHITE COATING (%)

 

 

 

 

 

 

 

 

 

 

 

 

 

4 N
9
=~ 3
<
O
'._
| & ]
<
o -5
L 10 10
z 2 0% 3
> \
m -
-7
£ —) o
\ \
‘ \
T 1008 0.3
, 10079 0.1
0 _
0 5 10 15 20

COATING THICKNESS (mils)

Fig. 5 3. Effect of Graphite Surface Seal on 135 Xe Poison Fraction.

 
 

 

b

61

these would keep xenon polsoning in the MSER at & very low level if the
coatings retained their integrity during reactor operation.

6. INFLUENCE OF GRAPHITE BFHAVIOR ON MSBR PERFORMANCE AND DESIGN

6.1 Effect of Core Pover Density on MSER Performance

A. M. Perry

Limitations on core pover density due to graphite radiation damage
will influence reactor performance. The performance of en MSBR mey be
Judged both in terms of the estimated power cost and also in terms of the
annual rate of net fissionsble material production (the annual fuel yield)
and the fuei specific pover. The fuel yield depends not only on the
breeding gain (breeding ratio minus one) but also enlthe specific powver;

 that is, on the thermal power of the reactor per unit mass of fissionsble

material chargeable to the plant (including material in the core, heat
exchangers and piping, and in the chemical processing plant) All three
factors of cost, breeding gain, and Specific power depend on the power
density in the core, but the dependence in each case is not unique.

That is, the extent to which each factor varies with power density depends
on other reactor parameters'such'es the fuel-salt and fertile-salt volume
fractions in the core, the'concentration cf'fissiOneble material in the
fuel salt, chemical processing rates, etc. An evaluation of the effect

of power density on MSER performance must therefore be based on a search
for the optimum combinetions of ell of these varisbles for each fixed value
of the average power denSity; The optimnn combination is defined here 1in
terms of a_camposite figure of merit, F, such that

F=Y+100 (C+Xx)* |,

uhere Y 1s the.annual fuel yield (the annuel percentage increese in fuel
inventory due to breeding), C 1ie the sum of all elements of the power

cost which depend on the parameters being varied, and X is en ad justable

parameter vwhose value determines the relative sensitivity of F to Y end
to C. Thus, F increases with increasing yield end increases with de-
creasing cost, and may be made to depend almost entirely on one or the

 
 

 

 

62

other. An optimum configuration is considered here to be one which maxi-
mizes F, and by repeating the search procedure with different values of X,
curves may be generated showing the minimum cost corresponding to each
(attainable) value of the annual yield. In practice, the variation in
cost is dominated by the changes in fuel-cycle cost (raw material plus
inventory plus processing costs less production credits), and the curves
derived from our calculations have therefore been plotted as fuel-cycle
cost versus annual fuel yield. Such curves are shown in Fig. 6.1 for
average core power densities of 80, ho 20, and 10 w/cm . These results
apply to a two-region, two-fluid MSER such as given in ORNL-3996. However,
preliminary results-obtained for single-fluid MSBR!s (considexing direct
pfctactinium removal and fission product discard using liquid bismuth
extraction_processes) indicate that‘compareble performance is-feasibie .
for such systems also. For convenience in relating the annual fuel yield
to the potential power doubling time, Fig. 6.1 also indicates the compound-
interest doubling time as a function of yield. .

It is apparent from Fig. 6.1 that there is an incentive to keep the
power density as high as possible. However, 1f the useful 1ife of the
graphite is limited to a fixed fast neutron dose, it is desirable also to
avoid the necessity for too frequent replacement of the graphite. The
influence of graphite replacement on plant availability and on pcwer cost
and the technical problems associated with this‘operetion;are discussed
in Section 6.3.

6.2 Effect of Graphite Dimensional Changes on MSBR Performance
A. M. Perry

During reactor exposure the graphite moderator in the MSBR is expected
to experience dimensional chengee approximately like those shown in Fig.
3.1, i.e., a period of shrinkage followed by increasingly rapid growth.
These dimensional changes must; of course, be allowed for.in.the mechani-
cal design of the core. In addition, the dimensional changes of the graph-
ite will alter the volume fractions of the three core constituents--
moderator, fuel salt, and fertile salt--and these changes, even though
accompanied by changes in uranium and thorium concentrations, may have

an adverse effect on reactor performence. There are two such effects
 

 

b

y

63

 ORNL-DWG 68-7983
POWER DOUBLING TIME (years)

100 50 25 20 5 12 10

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Q‘B | i , | { I i |
207 / I
=

< QOWU%mi// :

n
':E 0.6 // 20 '
- ] / 40
)
O
O A /80
w 0.5
d /
>
QO
) e
w 0.4 : o
2 —1

0.3 , ' ‘

0 | 3 4 5 6 7

ANNUAL YIELD (%/year)

Fig. 6.1. Effect of Power Density on MSER Performance.

 
 

 

 

6l

which especially require atfention. First, changes in graphite,difiensions
will cause a departure of reactor parémeters from the optimum combination
required to minimize costs and maximize fuel yield. Second, the spatial
distribution of neutron productions and ébsoxptions, which governs the
pover density distribution, may be appreclably altered because of éhanging
graphite volume fractions, making it difficult to maintein as flat a
power distribution as would be possible with a dimensionally stable moder-
ator. These are both rather complex questlons, and the extent to which
the MSBR performence might be compromised, when averaged over a peridd of
years, has not been fully analyzed. .However, the results dbfained to date
are sufficient to indicate approximately the effects to be expected.
With fertile salt filling the spaces between the graphite "fuel
elements" (two-fluid MSBR), it is clear that a 5% reduction in graphite
crosg sectional area gives rise to & large fractional_increése in the
fertile-salt volume fraction in the core--from an initial value of 0.06,
for example, to & maximum value of 0.11. Such a large volume fraction of
fertile salt is not optimum end, if uniform throughout the core, would
occasion a loss in annual fuel yield of about 0.0l and an increase in
fuel-cycle cost of approximately 0.1 mill/kwhr(e). The actual penalties
would not be this large, because the dimensional changes in graphi@e
would not occur uniformly throughout the core and because the time-averaged
volume change would be not much more than half the maximum change. The
average loss in performence, therefore, does not appear excessive - if
graphite dimensional changes are no more than 5 vol %.

A potentially more serious difficulty arises in connéction with the

power density distribution in the core, which should be maintained as flat

as possible throughout the core life to increase the time interval between
graphite replacement.' Calculations sfiow that the spatial pOweridistri-
bution is very sensitive to details of core composition, and that the
distributions of fertile and fissile materials in the core must be quite
closely controlled in relation to each other. In the presence of large,
spatially dependent changes in fertile-salt volume fraction, adjustments

in uranifim ahd thorium concentrations in the two salt streams do not

appear sufficient to maintain both criticality and a flat power distribution.

-
 

 

¥

L)

65

As.a consequence of the above considerations, the original concept
of the two-fluid MSBR4© was revised so that the fertile salt stream, as -
well as the fuel stream, flows in annuler passages defined by the spacing
between concentric graphite pipes. The interstitial spaces‘between graph-
ite assenblies would be filled with helium. For such & design, the relative

~volume fractions of the importent core constituents--the solid moderator

and the two selt streams--then remain nearly constent, while the variation
in helium volume has littletinfluence on reactor performance. This
approach largely eliminates penalties'in breeding performance in pover
flattening that might othervise result from dimensional changes in the
graphite.

Alternatively, use of a single-fluid MSBR would alleviate the in-
fluence of graphite volume changes on reactor performance. ‘The single-fluid
reactor contains fissile and fertile materials in the same salt stream,
and so changes in graphite dimensions influence both fissile and fertile
concentrations in the reactor equally. At the same time, fiesile and
fertile‘concentrationa can he'controlled independently due to use of on-

.stream processing. These conditions permit considerable flexibility

with regard to material concentrations, such that there is little change
in nuclear performance with expected graphite dimensional changes, based
on equilibrium physics -~ fuel-cycle calculations.

6.3 Mechanical Design Factors and Cost Considerations

n&kms‘”~mwnmmM'

As shown in Chapter 3. when graphite is exposed to a high neutron
flnx it first undergoes & period of shrinkege followed by swelling at an
ever-increasing rate. These effects occur both with and across the grain
structure of the graphite, although not necessarily at the seme rate in

~each direction, and are related to the energy of the. neutrons and to the

total accumulated dose. Such dimenaionel changes in MSBR graphite impose
mechanical and nuclear design problems, for example, it is necessary to
prevent overstresaing of the core graphite. Also, particulerly, for the
two-fluid design, the volumetric ratios of fuel-to-graphite need to be
maintained within 1imits in order to obtain good nuclear and economic

 
 

 

 

66

performance. Thus, the useful life of the MSBR core graphite and the
associated power production cdsts can be significafitly'influenced 5y the
neutron-radiation-induced damage to the graphite. The influence that
graphite volume changes and a finite permissible exposure have on reactor

- design features and performance are discussed below with respect to the

twoffluid and also the single-fluid MSBR condepts. The maximum permissible
radiation exposure to MSBR .graphite, based on presentlyrtested grades,

‘appears to be sbout 3 x 1022 neutrons/cm® (neutron energies > 50 kev)..

This exposure corresponds to & final graphite volume about equal to its
initial volume (see Chapter 3). . |

The two-fluid MSBR coref® is designed with re-entrant type fuel
channels in order to minimize the likelihood of mechanical failure of the

- graphite. Each fuel channel consists of concentric grephite pipes such

that the fuel salt flows upward through the center pipe and downward

" through the anmuler passage; the outer pipe is closed &t the top. At

the bottom of the core, the graphite pipes are brazed to Hastelloy N
nipples, with the other ends of the nipples being weldedftb the fuel
plena at the bottom head of the reactor vessel. Each fuel channel is
thus free to expand and contract in the axial (vertical) direction to
accommodate the dimensional changes in the graphite caused by thermal
effects and radiation-induced damage. |

In order to accommodate dimensional chenges in the core radial
direction, it is necessary to locate the fuel channels with sufficient
clearance to prevent interference when the graphite expands. Thus, the
top ends of all the graphite elements in the core are mechanically inter-
locked‘to'assure that they will maintain the same position relative to
each other while at the seme time not restrictihg the axiasl movement.
There are no unattached graphite elements or filler pieces in the core.
Also, in order to decrease the influence of graphite diménsional changes
on reactor performance, the two-fluid design was modified so that fertilé
and fissile streams are contained in separate annular flow regions B
definedlby the spacing between concentric graphite pipes. Helium was
used to £ill the intefstitial spaces between graphite assemblies, so
that changes in graphite volume have only a small effect on reactor
performance (see Section 6.2). |

 
 

 

 

 

e

67

fAs'pointed out previously, MSBR's can &lso operate as single-fluid
reactors, with features analogous to those of the MSRE. The performance
of single-fluid MSBR's can be as good as that 6f the two-fluid concept
so long as the fuel stream is processed on sbout a S5-day éycle to remove
protaectinium, and fission products are removed on &bout a 50-day cycle.
Recent chemicel discoveries suggest that processing methods which perform
the above functions are féasible, and indicate that such fuel processing
can be performed economically at a rapid rate, These methods utilize
liquid bismuth to selectively extract uranium, prbtéctinium,-and fission
products from fuel salt, and depend upon the relative nobilities of the

. varipus metals involved. Present information on relative ndbilities

indicates that reductive extraction processing effecting the desired
separations is possible, and thet the equipment involved ie small in

- size. Since protectinium is of intermediste nobility to thorium and

uranium, reductive extraction effectively holds Pa out of the reactor
uhtil it decays to uranium, after which it returns to the fuel system. -
Fission products are removed by concentration in a salt stream followed
by salt discard; alternative methods are also avallable for fission -
product removel from the fuel circuit. '

In the single-fluid concept, the fuel salt flows into the bottom
of the reactor and out the top in a once-through afrangement that permits
use of grafhite having simple geométny.y One of the present design con-

- cepts places the.graphite-eiements on a supporting grid at the bottom of

the reactor; these ‘elements are supported by this grid vhen' there 1& no

‘salt in the reactor. Also, a metal grid is used et the top of the
reactor to maintain proper spacing and alignment of the graphite elements;
a strengthenéd top plenum 1s used to react to the buoyant force of the

graphite when the reactor is filled with salt and operating. The top of
the reactor vessel and/or portions of it ere removeble so that graphite

“can be withdrawn vertically and replaced'as*needed.f"Changes in the graph-
ite dimensions in the axiel (vertical) direction are easily asccommodated

since the graphite is not-restrained.'_The graphitéielements-are“long
enough o0 thet 1f axiel shrinksge occurs, the graphite to fuel retio in
the active portion of the core due to this effect remains essentially
unchanged. Changes in nuclear performance due to radiesl shrinkage or

 
 

 

68

expansion of the graphite can be accommodated by changes in the fuel-

salt composition.

-After the MSBER graphite has received the maximum permissible exposure,
it must be taken out of service and replaced. In the two-fluid concept,
it appears that this would be done by replacing the entire reactor vessel

- and core. In the single-fluild concept the graphite itself would be re-

placed, with the reactor vessel remaining in place throughout the life of
the plant. The time required for this replacement, the replacement cost,
and the time between replacements all influence the power cost penalty
assoclated with graphite replacement. Also, for & given fiermissible
exposure, the time between graphite_replacements'can be increased by -
lowering the reactor power density. The influence of these factors on-
reactor power costs 1s discussed below.

Lowering the core power density to increase the useful life of the
graphite requires that the reactor be made larger, thus increasing the
cost of the initial reactor as well as that for replacément equipment.

For the two-fluid MSBR, the cost of replacing a spent reactor with a

new one appears to be a strong function of the reactor vessel size an&
wveight. Also, all the graphite is replaced in the operation. For the
single-fluid'concept, the reactor vessel would not be replaced and only

a pért of the total graphite would be removed during one replacement
operation. For both concepts, increasing the reactor vessel size leads
to higher fissile inventories and larger fuel-storage tanks, which incresse
fuel and capital costs. At the same time, lowering the core power density
leads to longer graphite life end reduces the number of times the graphite
must be replaced over the useful life of the power station. As a result,
thére ié & minimm in the curve of pofier cost versus core power density
for a specified maximum permissible exposure of the graphite.

. The effective cost of graphite replacement is also influenced by
plent downtime requirements associated with the replacement operation.
Since the MSBR would be fueled on & continuous or semi-continuous basis,
this concept has & potentially high load factor. Thus, if graphite re-

‘placement can be scheduled at times of regular turbine plent maintenance,

total reactor downtime should be no greater than normally expected in &
base-load power plant. This appears to be the case so long as graphite

 

 
 

69

replacement does not eceur'et intervals shorter than 2 to 2.5 years.
However, 1n order to determine the effect on costs of-iosing power pro-
duction due to graphite replacement, the term "effective downtime" was
treated as a parameter, where effective downtime is the time during which
pover production is lost due solely to graphite replacement requirements.
During the "effective downtime", it was considefed that pover would be
bought at 4 mills/kvhr(e) from an outside source. Values of zero, 1/2
and 1 month were used for the effective downtime. This nonproductive
time does not include plant downtime require& for normel maintenance
operations, which time could also be used for replecement operations.
Lebor costs associated with replacing the graphite were those for 18

men working in three shifts for two months at a cost of $10/br, including
overhead, etc.; these costs amounted to $259,200 per replacement.

Tables 6.1 and 6.2 summarize pover costs calculated for two-fluid and
eingle-fluid[MSER'e, respectively, as & functien of average eore pofier
density, on the bases given sbove; effective downtime fot replacing graph-
ite was considered to be 1/2 month in these cases. The results in Table
6.1 consider replacement of the entire reactor vessel and its contents
when the graphite exposure has reached & maximm value of 3 x 1022 nvt
(E > 50 kev); Table 6.2 considers a single—fluid MSBR with replacement
of grephite elone. Since costs and revenues occur st different times,

e "levelized" cost calculation was performed, using a 6% per year
discount fector. The fuel cycle rerformance for the two MSER concepts
- appear to be compareble, and so the same fuel cycle cost was used for
- each concept for a glven average core power density.

The capital cost data shown in Tebles 6.1 end 6.2 were based on cost
estimates made for a two-fluid, 80-kw/11ter, 1000-Mw(e) MSER station.
Rether'broad adjustments were made to these base costs in estimating
 costs associated with other core power densities and with the single-
fluid concept. While there ie'censidereble unce;tainty-essocieted with
_ the ebsolute costs given, the relative coste for the two concepts as &
function of core power density appear to be significant.,

Cost estimates were aleo mede on the basis that the effective plant
downtime associated with graphite replacement wasg elther one month or
zero. (The latter aesfimes that graphite replacement is performed during

 
 

 

 

T0

Teble 6.1. Effect of Core Power Density on Power Costs? in a 1000-Muw(e)
MSBR Station if Reactor Vessel is Replaced After Graphite
Reaches a Maximum Exposure of 3 x 1022 nvt (E > 50 kev)

 

Average Core Power Density, kw/liter

 

 

80 4o 20 10
Life of graphite plus vessel, years 2 oy 8 16
Costs per replacement, $10®
Reactor vessels (U4 cores) k.0 5.3 7.6 10.1
Graphitéb 1.2 1.9 3.1 6.3
Labor | ' 0.3 0.3 0.3 0.3
Power loss for 1/2 month 1.2 1.2 1.2 1.2
| ‘Total 6.7 8.7 12,2 17.9
30-year replacement cost, $10° 43.h 26.L4 15.5 7.0
Remote maintenance equipment, $10° 5.0 5.0 5.0 5.0
Total depreciating capital cost, 137 140 149 160
$/iu(e) \
Total power production costs,
mills/kwhr(e)
~ Capital costs® 2,34 2,50 - 2.5%  2.73
Reactor replacement costs 0.50 0.30 0.18 0.08
Fuel cycle costs® 0.b4  0.L46 0.52 0.62
Operating costs 0.29 0.29 0.29 0.29
Total,f

mills/kvwhr 3.57 3.L45 3.53  3.72

 

®Costs shown consider a four-module 1000-Mw(e) plant and include in-
spection and installastion costs plus 41% indirect charges.

Poraphite cost 1s based on $5/1b and & density of 112 1b/£t3.

®rime levelized replacement costs using a 6% per year discount faétbr.

dBased'on-la% per year fixed charge rate for depreciating capital and
80% plant -load factor. : |

®Fuel cycle costs include investment for fuel and blanket salts and
fuel recycle costs. The fixed charge rate for nondepreciating fuel was

10% per year.

fOn comparable bases, light water reactors would have capital costs
of 2.3 mills/kwhr(e), fuel cycle costs of 1.4 mills/kwhr(e), and power
production costs of 4.0 mills/kwhr(e).
 

 

 

T1

‘Table 6.2, Effect of Core Power Density on Power Costs™ in a 1000-Mw(e)
MSBR Station if One-Healf of Graphite is Replaced After Reaching a.
Maximum Exposure of 3 x 1022 nvt (E > 50 kev)

 

Average Core Power Density, kw/liter

 

 

80 Lo 20 10
Life of graphite, years 1.6 3.2 6.4 12.8
Costs per replacement, $10° |
Graphite® | 0.6 1.1 2.1 3.3
Labor 0.3 0.3 0.3 0.3
Power loss for 1/2 month 1.2 1.2 1.2 1.2
- Total 2.1 2.6 3.6 1.8
30-year replacement cost, © $10° 17.5 10.3 6.2 3.k
Remote maintenance equipment, $10€ 5.0 5.0 5.0 5.0
‘Total de reciating capltal cost, 128 131 134 136
$/ute) |
Total power production cost,
mills/kwhr(e)
Capital costs® 2,20 2,24  2.29 2,33
Grephite replacement coste - 0.20 0.12 0.07 0.04
Fuel cycle costs® | O.44h  0.U46 0.52 0.62
- Operating costs , _0.29 _0.29 _0.29 _0.29
Total,T mi11s/kvhr(e) 313 3,11 3.17 3,28

 

 ®costs ehown consider a lOOOéMw(e) plant utilizing & single reactor
vessel, and include 1nspection and installation coste plus 419 indirect
charges.

Paraphite cost is based on $5/1b and & density of 112 1b/ft3
qTim.e levelized replacement costs using & 6% per year discount factor.

dBasecl on 12% per year fixed charge rate for depreciating capital and
804 plent load factor.

, “Fuel cycle coets include investment for fuel &and blanket gelts and
fuel recycle costs. The fixed charge rate for nondepreciating fuel was
10% per year. | |

' On ccmparable bases, 1ight vater reactors would have capital costs
of 2.3 mills/kwhr(e), fuel cycle costs of 1.k mills/kwhr(e), and power
production costs of k.0 mills/kvhr(e).

 
 

 

 

no net load-factor penalty applied to MSBR's relative to other systems.)

nomic acvantage in deve;bping an improved radiation-resistant graphite and

 

T2

normal plent maintenance operations, and this is considered to be the
reference condition. Graphite replacement can be considered equivalent
to refueling dperations in other reactor types, and so there should be

Further, the influence of graphite permissible exposure on power costs

was determined by considering the permissible exposure to be either

6Ix-10?2 nvt (E > 50 kev) or 30 years (versus 3 x 1022 nvt for reference

case).i In these latter studies no effective downtime was associated

with graphite replacement. The results obtained, including those given

in Tebles 6.1 end 6.2, are summaerized in Fig. 6.2. | |
The overall results given in Fig. 6.2 indicate that there is an eco- N

that, for a gliven exposfire lifetimé, maintenance concepts and méthods that
reduce effective graphite replacement costs and replacement downtime are
economically desirable. In utilizing these results, it should be remem-

bered that a maximum core power density of 100 kw/liter for 2 to 2.5 yeérs

corresponds to & zero net change in graphite volume and to an nvt'(E > 50
kév),for graphite of about 3 x 1022: peutrons/cn=. For the two-fluid con-
cept, if graphite had a permiésible'exposure'lifetime of 30 years at an .
average core povwer density of 80 kw/liter, the minimum power generation
cost would be about 3.03.mills/kWhr(e); the minimm cost would fie about
3.41 mills/kvhr(e) based on a permissible graphite exposure of 3 x 1022

-pvt and zero effective downfime. :The difference between 3.03 and 3.41

mills/kvhr(e) power cost amounts to sbout $80 million of revenue over.
the 30-year life of a single 1000-Mw(e) power station. If the eléctric
utility industry were to employ 100 such molten-salt breeder reactors at
a glven time, about $265 million per year would be:associated with re-
moving exposure limitations on the graphite. Doubling the grafihite life
from 3 x 1072 to 6 x 1022 in the two-fluid reactor would reduce power
costs'by about 0.2 mill/kwhr andrbe worth about $125 million per year for
one hundred 1000-Mw(e) MSBR's. For the single-fluid reactor, the comparable ,
inceptives would be about $28 million per year for doubling the graphite
life, and about $90 million per year for removing reétrictions on graphite

~1ife, Thus, even considering a reasonable_d;scbunt_factor, a significant ksJ
‘effort for graphite improvement can be economically justified if such work

leads to a graphite with improved irradiation characteristics.

 
e

T3

ORNL~DWG 68-7984

-FLUID MSBR
{GRAPHITE REPLACEMENT)

POWER PRODUCTION COST (mills/kWhr)

© TWO-FLUID MODULAR MSBR
(VESSEL PLUS GRAPHITE REPLACEMENT)

 

o 10 20 30 40 S50 60 70 80O 40 20 30 40 50 60 VO 8O
; AVERAGE CORE POWER DENSITY (kW/liter)

CURVE A. ONE MONTH CHARGEABLE POWER OUTAGE 22 2
CURVE B. ONE-HALF MONTH CHARGEABLE POWER OUTAGE zszHI:'?::;é‘grmg?;ifi:gkevl
CURVE C. NO CHARGEABLE POWER OUTAGE :

CURVE D. NO CHARGEABLE POWER OUTAGE, euo22 neutron/cm

, MAXIMUM GRAPHITE EXPOSURE -

CURVE E. 30-yr LIFE GRAPHITE (REQUIRES NO REPLACEMENT)

Fig. 6.2. Effect on MSBR Power Production Costs of Core Power Density,

Graphite Life, and Duration of Power Outage for Graphite
Replacemen‘b

 
 

 

Th

The power cost results given in Fig. 6.2, for which effective down-
time was treated as a parameter, show that, for the cited conditions,
lower effective downtime for graphite replacement leads to lower power
costs. Increasing the effective downtime from zero to one month in-
creased minimum power costs by 0.08 to 0.15 mill/kwhr for permissible
graphite exposures of 3 x 10%2 nvt (E > 50 kev).

The ebove results indicate that exposure limitations for MSER graph-
ite lead to less economic penalty to the single-fluid MSER than to the
two~-fluid concept. Nevertheless, an improvement in graphite behavior is
desirsble for both concepts. o '

6.4 The Influence on MSBR Performance of
Noble-Metal Deposition on Graphite

A. M, Perry

It has been recognized for several years that uncertainty in the
chemical behavior of certain of the fission products--notably niobium,
molybdenum, technetium, and to a lesser extent ruthenium and tellurium--
constitutes one of the principal,uncertainties in estimates of the
breeding capdbilities of molten-salt reactors. The Molten-Salt Reactor
Experiment 1s being used to reduce or remove this uncertainty, and it
has already yielded much encouraging information of value in this regard.

The essential question is whether these fission products will remain
in the core, or whether, as we have assumed in our MSBR performance
estimates, they will be removed from the melt dfirifig fuel processing, or
perhaps be deposited as metals on the Hastelloy N surfaces outside the
core. Experlence in the MSRE indlcates that most of the noble-metal
fission products appear in the gas phasé of the pump bowl.

Should they all remain in the core of an MSBR, they would signifi-
cantly reduce the breeding ratio. While the cross sections of these
1sotopes are not especially large, thelr combined fission yields account
for nearly a quarter of the total yield of fission products, and the
cross sections of the stable isotopes in the group sre, in several instances, -
sufficient to allow saturation to occur in a few years. At saturation, kiJ
the rate of production by the fission of uranium equals the rate of removal
 

 

™

by neutron capture, and the total qpantity of the materisl in the core
becomes constant., The neutren loss--and hence the reduction in breeding
ratio--then'depends onlyvon_the fiss;on product yield, not on the'cross
section. - The neutron poisoning, 'i’ at any time t after startup, due to
a particular stable isotope designated.by subscript i is expressed
approximately'by

-o‘itbt -
6.1)
Il + od (; ‘) ! (6.
where ¥y i the fission yield of nuclide i,'oi is its effective spectrum-

averaged cross eection, 0. is the flux in the reactor core, & is the
cepture-to-fission ratio for the fuel, and f, is the fraction of this -
fission-product specles that 1is deposited from the fuel salt and remains
ih the core. The value of Pi gives directly the loss in breeding ratio
assoclated with this fission product.

Estimates of the amount of poisoning that could result from depo-
gsition of these fission products in the core have been made from time to
time during the evolution of the MSBR design.t While the fully saturated
polsoning depends very little on details of the reactor design, the rate
of approach ‘to saturation does depend on detailed‘design paremeters, and
this accounts for some differences in the estimates that have appeared.

Table 6 3 gives the maximum reduction in breeding retio assoclated
with the stable and very long-lived isotopes of Mo, Tc, Ru, Rh, Pd, and
Te, 85 & function of time after reactor startup or after the installation
of fresh core graphite. Thesge numbers correspond to complete deposition
on the graphite of the entire yield of each of these 1sotopes. In some
cases the prdbebility of deposition of the etable poison is assumed to
be associeted with the chemicel behavior of -its precursor. For this
reason, niobium deposition behavior,_as vell as_thet_of_m01deenum, is
important--The quantityi(oQ)';-ih Table 6.3 is the time required for'e
nuclide to reach ebout TO% of its saturation value. These time constents
are computed for noble fietel fission producte in the core region of a
single-fluid MSBR, considering a 90% ‘plant factor end & fuel specific
pover in the "core" of 10.7'Mw(t)/kg fissile. The total poisoning in
Table 6.3 is the lose in breeding raetio at the given time after startup;

 
 

76

Teble 6.3. Loss of Breeding Retio Corresponding to Cdmplete Retention

of Certain Fission Products in & Single-Fluid MSER

 

: (U@)-l

Time After Core Startup (years)

 

 

Nuclide (yr)a 1 5 \ 8 16

- 95M0 k.3 0.0062  0.0111  0.0186  0.0272  0.03k5
Mo 29 0.0009 0.0018 0.003% 0.0086  0.0110
Mo | 93' 0.0003  0.0005 0.0010  0.0021  0.0040
1000 %5 1 0.0002  0.0005  0.0009 0.0018  0.0035
99pc 3.1 0.0067  0.0116  0.0183 0.0252  0.030k4
101, 7.3 0.0019 o.0037'_ 0.0066 - 0.0107 o.oisq.
102p, 42,5 0.0003 0.0005 0.0010 0.0020  0.0037
104py, 66 0.0001  0.0002  0.0005  0.0009  0.0011
103y 0.41  0.0096  0.0117 0.0138  0.0158  0.0168
10Spq 6.0 0.000k = 0.0007 0.0012  ©0.0019  0.0026
107pg 9.1 _ 0.0001  0.0002 0.0003  0.000k
1257e 46 - 0.0001  0.0002  0.0003  0.0006
1287, 230 0 0 0.0001  0.0002 = 0.0003
130me 154 0.0001  0.0002  0.000k  0.0007  0.0013
To£a1 0.026T  0.0427  0.0662  0.0977  0.1252
P (average) 0.015 0.026 0.041 0.088

 

 

 

0.063

 

®These saturation time constants (time required to reach about TO%

of the equilibrium value) epply in the "core" zone, which contains approxi-
‘mately half the graphite area exposed to fuel salt. The time constents

for the "blanket" zone are ebout ten times longer.

 
 

 

 

 

 

-

(i

in the last row of the table, however, the average loss of breeding ratio,

P, over time, t, is given, where

_F =%f P(tt)at* . i - (6.2)

O

Results obtained from graphite samples exposed in the MSRE regarding
the behavior of these fission products are discussed in Section 4.,2. From
Teble 4.1 it is noted that, on the assumption that the graphite samples
are typleel of all graphite surfaces exposed to the selt, approximately
10.9% of the ®°Mo produced in the MSRE was reteined on the graphite as
well as 10.0% of the *33pe, 6.6% of the °°Ru, end 36.4% of the ®SWb. In
using these results to estimate the fraction of the stable fission product
poisons retained on the graphite surfaces in an MSBR, account is teken of
the difference in the ratio of graphite-to-metal area in the two reactors.
In the MSRE, the graphite comprises'63% of.the area exposed to salt, where-
ag in the single-fluid MSER, the "core" graphite represents about 40%. 1In
addition, the MSRE results indicate a considerably greater affinity of the
noble metals (except for Nb) for ‘the metal su:face than for the graphite
surface. Thus, 1t is expected that the pefcentage'depositibnlof noble
metal fiselon products on graphite in the MSER would be less then 1n the
MSRE, vith the ratio dependent upon the kinetics of the deposition process.
On the basis that fiseion products have eccess to all surfaces equally,
their relative deposition on MSER graphlte would be less than ore-third
that observed in the'MSRE,;however, since many of the fission products

are generated in the core region, the factor 1s probably about one-half.

Thus, in this enalysis, the percentage of noble metals retained on ‘the -
MSER graphite 1is considered to be 5% for 99Mo, 52pe, and 1°SRu, and

20% for ®5Fb. It is further postulated in view of the small fraction of
these muclides found in the galt (see Table k.1) that the deposition is

relatively rapid compared to the decay rate of radioactive precursors

of the steble noble metal poisons, conseqpently, the deposition fractions
of the etable poleone ere those of their precursors where the fission
yield is zero. Thus, ®°Mo is assumed to be deposited in accordance with
its precursor SSNb, while the other Mo isotopes and ®%Tc are assumed to

 
 

 

 

 

 

78

~ behave like ®%Mo. Similarly, the behavior of °®Rh 1s assumed to be
governed by that of 1ts precursor 1°3Ru; Pd is &lso assumed to-behave
like Ru, although its contribution is very small. Finally, in view of

- the marked difference in neutron flux intensities (sbout tenfold) in

the "core" zone and in the "blanket" zone of the single-fluid reactor,
the expression for saturation of the deposited fission prbducts was modi-
fied by 1hcluding a separate term for each of the two zones. For the
ccmbined poisoning of all the noble metal fission products 1isted in
Teble 6.3, the above conditions give the results shown in Table 6.k,

with P(t) and P(t) defined as before.

Table 6.k, Anticipated Roble-Metal Fission Product Poisoning in -
a Single-Fluid MSBR (Loss of Breeding Ratio)

 

. Pime After Startup (years)

 

 

1 2 L 8 16
' P'(t) | 10.0022 0.0038 6.0061' | 0.0089 0.011k
P(t)  0.0012  0.0022  0.003  0.0056  0.0079

 

It mfiy‘be seen from Table 6.4 thet for exposures of up to 10 years!
duration the degradation in breeding ratio due to deposition of noble- -
metal fission products is expected to remain less than 0.0l, and the |
cumilative average will be smaller still. Inasmuch-as the graphlte will
probably be replaced because of radiation damage.considerations at inter-
vals shorter than 10 years, it appears that the average loss in breéding
ratio wiil.be in the range of 0.002 to 0.005 due to fission product
deposition on the graphite. |

Thus, although complete retention of the ndble-metal fission pro-
ducts on core graphite leads to a significant reduction in MSBER breeding
ratio, the deposition behavior inferred by MSRE tesults glves only a
small reduction in MSER perfbrmance, Addifiional experimental results
are needed to confirm these,prfiliminary indications.

 
 

 

 

 

or

T9

6.5 Conclusions

‘Graphite dimenEional'changee due to exposure in an MSER can alter
the relative volume fractions of moderator, fuel salt, and fertile selt
in the reactor. Such changes influence the design of & two-fluid MSER
more then a single-fluid reactor, since in the latter the fertile and
fiseile materiels ere mixed together end their ratio does not change
when the graphite volume chengee. By constructing a two-fluid resctor
such that the fissile and fertile.materiale are_confined to channels
within the graphite assemblies end the spaces between graphite assenblies
ere filled with helium, changes in graphite volume fraction lead largely

'torreletive volume change in the helium space. Such volume cnanges have

only e small effect on fuel cycle performance end on power distribution.
In a single-fluld MSER, graphite dimensional changes would have little
effect on nuclear performance since the fissile and fertile ealt ‘volumes
ere equally effected. Aleo, the ebility to 1ndependent1y adjuet fissile
end fertile materiel concentrations in both two-fluid and single-fluid
MSER's permits adjustment in reactor'performance es changes in graphite
volume occur. Thus, 1itt1e change in nuclear performance is expected
because of radiation damage to grephite so long as the graphite volume
does not increase much beyond its 1nit1a1 value and the graphite diffusion
coefficient to gases remains low during reactor expoeure (the latter con-‘
dition neglects the possibility of removing xenon efficiently'by gas
stripping)

A 1imit on the permiseible expoeure of the graphite ¢an have & sig-
nificant influence on reactor design conditions, If there vere no ex-

- posure limit, the aversage core ‘powver density corresponding to the minimum
‘power cost would be in excees of 80 kw/liter. Eowever, 1f a limit exists,

high power density can leed to high cost because of graphite replacement
cost. At the seme time, decreasing the core power density leads to an
increase in capital'cost and fuel cycle cost. 'Thue, a 1imit on permiesible

.grephite exposure generally requires & compromise between various cost

items, with core power density choeen'on the basie of power cost. The
optimum pover density also varies with MSER concept, since only graphite
requires replacement in a single-fluid MSBR, while both the reactor vessel
and graphite appear to require replacement in a two-fluid MSBR because of

 
 

 

 

80

the complexity of constructing the latter core. Further, reéctor'power
outage due'solely to graphite replacement requirements can be a signifi-
cant!cost'factor, However, if graphite were replaced at time intervals
no less than two years, 1t appears feasible to do the replacement operation
during normal turbine maintenance periods, such that no effective down-
time is.assigned to graphite,rerlacement. A two-year time intervel is .
associated with an average power density in the power-producing "core"
of sbout k0 kw/liter and a graphite exposure of about 3 x 1022 nvt
(E > 50 kev); For the above "réferencg“.conditions, the single-fluid
MSBR has power costs ebout 0.35 mill/kwhr(e) lower then the two-£luid
MSBR. Doubling the pernfissibie'graphite exposure [to a value of ebout
6 x 1022 nvt (E > 50 kev)/ would be more important to the two-fluid con-
cept and wbuld reduce powver costs by about 0.15 mill/kwhr(e); the corre-
sponding change for the single-fluid MSER would decrease power costs
by ebout 0.07 mill/kwhr(e). If a two-week effective reactor downtime
were asgigned solely to grafihite replacement operations, the associsted
povwer cost penalty would be about 0.07 mill/kvhr(e) for elther concept.
Deposition of noble-metal fission products‘in the core graphite of
an MSER would tend to lower the nuclear performance of an MSBR. Based on
the results obteined in the MSRE and taking into account the higher
metal/graphite surface area in an MSER relative to the MSRE, it is esti-
mated that deposition of fission products on the graphite in an MSBR
" 'would reduce the breeding ratio by about 0.002 on the average if graphite
vere replaced every two years, and about 0.004 if replaced every four
yesars. Thus; although complete retention of the noble-metal fission
- products on core graphite'would lead to a significant reduction in MSER
breeding ratio, the deposition behavior inferred from MSRE results corre-
sponds to only & small reduction in MSER performance.

T. FROGRAM TO DEVELOP IMFROVED GRAPHITE FOR MSBR'S

W. P. Eatherly C. R. Kennedy
D. K. Holmes R. A. Strehlow
Recent work on graphite implies that materisls can be developed in
the near future having improved properties for reactor application. The
avallable information supports the hypothesis that resistance to radiation

 

 

 
 

 

[ L3

[ LY

81

damage 1is strongly connected to large crystellite sizes and to minimal
binder content. Since the binder phase is, in general, dominated by
small and highly disoriented crystal structures, these two bases may

‘indeed by synonymous.

In connection with the graphite problem, representatives of ORNL
have visited all U.S. centers where active research on graphite is belng
undertaken and all vendors who have - expressed interest in the molten-
salt reactor program; As a result of these visits and our own analyses
of the problem, we have concluded that a graphite research and develop-

‘ment program conducted largely (but not exclusively) at Oak Ridge

Nationel Leboratory is desirable and essential to furtherance of the
fiolten-salt reactor concept. For convenience the program is divided
into five areas: (1) Fundamental Physical Studies, (2) Fundemental
Chemical Studies, (3) Febrication Studies, (4) Engineering Properties,
(5) Irradietion Program. This program is aimed not only at the devel-
opment of & suitable type of graphite, but also at establishing an
improved model for radiation damage which will aid in guiding graphite

| development. -

At the present time 1t appears that a radiation damage model can
probably be esteblished which will possess predictive capacity and define
the limits of materisl capability in withstanding ifradiation. Such a.
model is desirsble not only in guiding the development of superior
materials, but also to define the ultimate material limitations on the
reactor concept and design. Our confldence in the establishment of such

~ a model rests on the emergence of recent techniques offering increased:.

control over graphite miérbstrubture, on the continuing development of
néw diagnostic techniques ihich'enable one tb'dbtain both quantitative -
and qualitative information on microstructures, snd on the present indi-
cation that radiation damage at elevated temperatures may'be more tract-
able to analysis.

‘As indicated above, the attainment of improved graphite for molten-

 salt reactors (viz., lifetimes of 5 to 10 x 1022 neutrons/cm®) appears
~ possible to ephancement of crystallinity end by minimization of binder -

content. These postuletes rest primarily on British theories based on
single~crystal experiments, work on pyrolytic graphites at Gulf General

 
 

 

82

Atomic and irradiation data on certain relatively binder-free graphites.
The most promising routes of attack appear to be catalysis and pressure
carbonization, methods not largely explored by the graphite industry,
particularly with regard to radiation damage.

The development program is summarized,in\more'detail below.

7 1 Fundamental Physical Studies

The ultimate solution to the problem of increasing the resistance
of graphite to radiation demage may depend upon a fundemental under-
standing of the defect processes underlying the observed property changes.
A coordinated effort should he planned for establishing the basic mechan-
1sms of radiation damage. Damage models studied to date do not seem to -
offer a cdmpletely-acceptable_explanatipn of all aspects of the,damage
observed at. high doses end reiativel&rhigh temperatures; héwever, such
models do indicate general directions for further investigation.

The crystalline composition of a given graphite seems to play an
importent role in the finel results of the damage; thus, it appears |
Important to study single erystals, polycrystelline samples, and pyro-
lytics (as transition materials) in crden to hetter understand this

crystallite~size effect. Because of the high exposures required, it
 also seems important to utilize cherged particle bombardment (atong with

fast neutron irrediations in high flux reactors) in order to permit the
accumulation of irradiation data in & reasonable time. This, of course,

necessitates the use of thin specimens which may require careful devel-

opment in some cases..

Various property changes (with irradiation) can be studied in each
graphite material as deemed expedient for best identification of basic
defect strnctures. Among the most importent are dimensional changes,

| lattice spacing changes, and chané&s in thermal expansion‘coeffieientsnf

and elastic moduli., Obtaining these properties (and others) may require
supplemental work in developing technigues ‘and esteblishing the precise
property values of material in the unirradiated condition. In particulsr,
use of electron microscopy in investigating defect clusters and their
growth has already been sghown to be of ‘great velue and would be of
immediate utility, especially in association with single-crystal

 

 
 

 

 

oy

83

irradistions. Additional valuable techniques which have not yet been
exploited adeqnately are x-rey line shape analyses, optical transmission
and decoration. - '
Theoretical support of the experimental work is required &t three
levels. Any realistic damage model must first involve a set of complex
rate equations which would best be solved by automated analysis. Secondly,
the basic defect energetics and interactions employed in the rate equations
must be studied from the viewpoint of solid state theory. Finally, the
entire model must be related to the directly observable parameters char-

- acterizing polycrystalline graphite.

T.2 Fundamental Chemicel Studies

Recognition of the experimentally observed relationship between
radiation-induced growth rate and crystallite size give reasonable assur-
ance that an‘imprbved graphite can be developed. Crystallinity is strongly
influenced by chemical chenges occurring throughout the graphite manu-
facturing process. Three chemical approaches to the tailoring of the
crystellite size dietribution are: (1) alteration of carbonization con-
ditions for filler-residual binder systems (e.g., carbonization pressure);
(2) elimination of residual binder; and (3) modification of the graphite
by catalytic recrystallization.

Residual binders (those yielding part of the carbon 1n a graphite
body) carbonize and begin to develop their crystalline habit primarily by

free radical mechanisms with evolution of the gases Ez0, CO, Ha, etc.

This habit of texture perSistS‘throughout the graphitizetion process.
Changes in the crystallinity of the final product may'be accomplished
by chemical slteration of the binder materiel and by application of
pressure dnxing the critical,baking operation.

" In order to eliminate the residual binder one can utilize fugitive
binders during green article febrication which can evolve before sub-
stantial hardening of the article occurs. The use of raw or semicalcined
cokes presents a prbmising'ccurse of action because of the inherent
chemical activity of those material. The study of solvent action on
these filler materlals 1s a necessary first step.

 
 

8L

- Cetalytic modification of graphite has been demonstrated to yield
an incressed crystallite size. Either of two rather distinct mechanisms
may be involved. The firet is via a solid-state diffusionsl path..;
Thorium, uranium, and titanium cerbldes, for example, in the presence of

 excess carbon have been observed to improve the graphite crystallinity.
A second mechanism eppears to be operative for carbides et temperatures
above the eutectic (or peritectic) temperature where a solution-reprecipi-

tation process can be readily driven by the free-energy differences between

large end smell crystallites,

The free-energy differencés can 8lso be expected to result in reaction
rate differences; measurements of those rates could augment x-ray studies
of the crystallite size. In view of the difficulty of obtaining crystal-
lite-size distribution data from x-ray analysis, some effort in the field
of chemical kinetics is desirable. Studies of gas evolution and catalyst
removal from carbons at temperatures above 1500°C are expected to assist
further in improvement of process control as well as to provide fundemental

- information.

T.3 Fabrication Studles

The fébrication of grafihite samples for irradiation and physicél
pProperty evaluation 1s aimed in two complementéry.directions: first,
to provide the more fundamental programs with controlled test materiels,
and sécond, to take quick advantage of any information developed by these
programs. It would also include the development of suitable graphité- |
Jjoining techniques and-pyrolytic-carbon sfirface impreghation techniques
for control of gas'penetration into the gréphite. It is envisioned that
the scope 6f grephite fabrication would not proceed beyond semple prepa-
ration, with scaleup beilng left to commercial vendors.

The highly specialized nature of graphites sultable for molten-salt
applications required advanced fabrication techniques and strains the

limits of current graphite technology. For these reasons, it has been

our experience that vendor participation can be successfully_secured
only if their claims to protection of proprietary information are |
respected. On this basis two companies ere actively scaling up processes
to supply a graphite applicable to first cores in an experimental MSER,

 

 
 

 

 

i

[ L

85

two other companies are actively supplying samples of more advanced
materials, and several others have expressed an interest in subsequent
participation. | | |

Under these circumstances, an in-house cepsbility of supplying
materials for irradiation becomes essential. Only on this basis do
materials become availsble of known character end controlled variability.
Conversely, as long as vendor interest remains asctive and substantive,
the difficult problems of ProCcess scaleup and control ‘can remain with
commercial suppliers. It is obvious that this approach to the graphite
problem will require close and continued cooperation between ORNL and
commercial suppliers.

T.4 Engineering Properties

Candidate graphite materials must be evaluated and engineering date
generated to obtain the data required for proper design of an MSER core.

_ This willrrequire that sufficlent property values be determined within
 reasongble confidence intervals for specifying the design parameters.

The bulk physical properties of the materials must be determined with
particular_emphssis on any effects that surface costingsvmay have. The
mechanicsl end thermal prbperties must be critically‘eveluated,vith respect
to'possible enisotropic behavior. Sensitive properties determining the
compatibility of the graphite with the MSER environment, such as entrance
pore diameter, accessible pore volume, and penetrstion characteristics,
mist be examined very carefully. Also, the effects,of irradiation of
these properties mist be studied carefully. ;
Sound methods of quality control must be developed to ensure the )
soundness of all material to be used in an MSBR core. Techniques devel-
oped to ensure the integrity and effectiveness of coetings and of metal
and/or graphite joints must have a high degree of reliebility. There
must also be development of nondestructive testing techniques and property
interrelstionships to reduce the amount of destructive testing required to

~_ensure total integrity of the fsbricated_parts.

 
 

 

 

 

86

7.5 Irradistion Progrem

Initially the irradiation program.will be directed to provide eritical
information assisting both the fundamental and developmental programs.
Eventually the program.will be devoted to evaluating candidate materials
and to generating necessary engineering_data. These studies require |
graphite irradiation exposures to‘a level where failure occurs or which
exceeds the lifetime requireménts of an‘MSBR; This necessitates that
irradiation be done in reactors having high flux levels. Preliminary
experiments"in'target rod poSitions in the core of thé'HFIR have already
been performed and demonstrate the ability to maintein an irradiation
temperature between 690 and T30°C over prolonged periods. This faciiity
has the capability of accumlating a maximm of 4 x 1022 neutrons/cn®
(E > 50 kev) per year; even with recycling losses, the exposures will ve
gbout 3 5 x 1092 neutrons/cm (E > 50 kev) per year.

‘The main disadvantages of the HFIR irradiation facility is the small
size which limits the experiments to a 1/2-in.-OD tube. Therefore, it
will be necessary to consider the use of other irradiation facilities for
studies requiring larger samples. These atudies will be designed to
determine the combined effects of stress and irradiatidn on-the properties
of graphite and to investigate the possibility of size effect on dimen-
sional stability.

Ion-bombardment testing is also planned as a meens of acreening
graphite samples. This treatment would be used either as an ad junct or
as a substitute for high-flux neutron irradiations of grapnite. It is
proposed that the feasibility of ion-bombardment testing be examined
thoroughly to determine whether such studies can feed back information
to both fundamental ‘and developmental studies.

T.6 Conclusions

Irradiation iesults for different grades of graphite have shown that
gross volume changes are a function of crystallite arrangement as well as
size of the individual crystallites. Also, in graphites cnntaining'binder
materials, it appears that the binder region has little capacity to
accommodate or control_particlé strein and thus fractures because of
buildup of mecnanical stresses. This indicates that graphites with

 
 

 

 

87

improved radiation resistance might be obtained by developing graphites
having little or no binder content. Further, improved radiation resistence
appears to be associated with :lsotrop:l.c graphites made up of large crystal-
lites. Consequently, a research and development program aimed &t producing
improved graphite would emphasize development of graphite having large
crysta:_l_.lit_e sizes and little or no binder content. Such a program would
involve physical , chemicéi fiiechanical fa'br:l'catioh , and irrsdiation
studies » and could 1ead possibly to graphites with permissible fast neutron
exposures of 5 to 10 x 1022 neutrons/cm® (€ > 50 kev).
 

88

APPENDIX

’Graphite Exposure Messurements and Their Relationships
to Exposures In an MSER
A. M. Perry

Irradiations of near-isotropic graphites have been carried out in
the Dounreay Fast Reactor (DFR), providing information on dimensionel
changes as a function of fast neutron dose in the temperature range and
at the high neutron doses of interest in the MSBER. The DFR irradiations
are reported in terms of an Equivalent Pluto Dose (EPD), which investi-
gators in the Uhited Kingdom employ as a standard dose unit in order to
express results of experiments carried out in several different facilities
in directly comparsble terms. In order to apply the results of the DFR
irradistions to the MSBR, we must establish a connection between the
Equivalent Pluto Dose and the irradiation conditions to be expected in
the MSER.

Rather than computing an Equivalent Pluto Dose (EPD) for the MSER,
vhich would require detailed information on the reference spectrum in

Pluto, it is convenient to establish a correlation between neutron-

induced demage and the integrated neutron flux above some standard refer-
ence energy. ©Such a correlation is extremely useful if it can ba shown
that there exists an energy E, such that the ratio of observed damage rate
to the flux above energy E, is essentially the same for all reactor spectra
in which graphite damage is measured or needs to be known. Mathematically
this can be written as,

 

Q
f ® (E) D (E) aB
R(E,) = 0 = (A.1)
f<b (E) aE
EO

vhere D(E) is a "damage cross section™ giving the relative graphite damage
per unit fast neutron flux as & function of neutron energy, ®(E) is the
fast flux per unit of energy, E is neutron energy, and R(E,) is relative

 
 

 

. %

by 2.

89

demage to the graphite. Figure A.l shows the value of D(E) as a function
of energy, based on the carbon scattering cross section, the energy dis-
tribution of primary-carbon-recoil atoms following a neutron collision,
end on the number of carbon atoms displaced from thelr normal lattice
positions by a primary carbon recoil atom ag & function of the recoil-
atom energy. This lest function has ‘been calculated by Thompson and
Wright, end predictions based upon 1t have compared well with experimental
observations.l |

Figure A.2 shows neutron spectra produced by a fission source in
four widely different neutron moderating materials, these materials being
Hz0, D20, C, &and & mixture of equal volumes-of gsodium end uranium =
(enriched to 20% in the 235U 1sotope). The last composition is intended
to be representative of a fast reactor core. In Fig. A.3 the function
R(E,) is shown as a function of E, for each of these four spectra.
Since a1l four curves mcross within a 4% band at 50 kev, it appears that
the desired correlation exists.

In order to utilize Eq. (A.l)'we need the neutron flux exposure above

50 kev corresponding to the EFD in DFR. The total neutron dose in DFR

exceeds the EFD by a factor of 2.16. This is just the reciprocal of the
factor that was used to infer the EPD from the total dose in the first
place.® In sddition, it is estimated that epproximately 94% of the
neutron flux in DFR is sbove 50 kev; while this fraction is not accurately
known to us at present, the uncertainty involved is believed to be small.
Thus, the exposure in the DFR to neutrons asbove 50 kev is (2.16)(0.9%) x
(EFD) = 2.0 x (EPD); 'That 1g, the EPD scale on the damage curves obtained
from Harwell is converted to dose (based on E > 50 kev) by multiplying

Results of graphite damage experiments in the GETR have been reported
in terms of the dose &bove 180 kev, The spectrum in these experiments was

 

1M, W. Thompson and S. B. Wright, J. Nucl, Matls. 16, 146 (1965).

2\, J. Perks and J. H. W. Simmons, "Dimeneional Chenges and Radi-
ation Creep of Graphite at Very High Neutron Doses," Carbon L, 8 (1966).
 

 

 

RELATIVE NUMBER OF DISPLACEMENTS

90

ORNL-DWG 68-7985

 

102 10° ot 10° 10° 10
£, , NEUTRON ENERGY (eV) -

Fig. A.1l., Number of Atom Displacements in Graphite per Cm>-Sec per
Unit Flux as a Function of Neutron Energy.

 
 

 

 

 

¢ (U) (arbitrary units)

91

NEUTRON ENERGY (MeV)

ORNL-DWG 6€8-7986

 

 

 

 

 

 

 

 

 

 

 

 

 

10 8 6 4 3 2 15 1008 06 04 03 02 045 040 008
W T T T T T T T T T T T
1 H,0 SPECTRUM
2 D,0 SPECTRUM
3 CARBON SPECTRUM
4 FAST SPECTRUM
o (50 vol % Na, 50 vol % U—METAL ,20% 233y, 80 % 238y)
[\N‘ \
05 | A N\
' / 2 A N_ >
—\—3
3 4
/ \1
0
0 1 2 3 4 5
LETHARGY U
Neutron Flux per Unit Lethargy Versus Neutron Lethargy

Fig. A.2.

Normaliz’ed for Equal Damage in Graphite
 

 

R(£g), (102 displacément-s/uni'r flux)

ORNL~-DWG &7-10499

 

 

 

 

 

 

  
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

11 '

. . I L...I

. @« 1l
. {qb(E)a(E)dE 7 / I
0 —— Rlbyle 5——— I /
1=H,0 SPECTRUM /' / L—3 ! /
2=D,0 SPECTRUM £, PLEVIE / .
- 3=CARBON SPECTRUM
9 —— 4=FAST SPECTRUM
8 ‘I’ —— T — 1 —_—f— - ——————
7 ~23%
i ’
L LA A
6 71 Lo
~ 397 [ /l' -/
~ T3

5 L e |

i 3 ~7 3

{ ,\»” o
1" 7
Eo (GV)

Fig. A.3. Graphite Damage
Spectra .

per Unit Neutron Flux in Various Reactor

 
 

 

93

such that the dose (E > 50 kev) is 1.18 x dose (E > 180 kev).® Thus,
results of the several experiments can be placed on the same dose scale,
for which equal dose should imply equal damage (other factors also being
equal) even for widely different spectra such as those in the DFR and in
the GETR. , |

Based on the sbove analysis, the permissible dose (E > 50 kev) for
tfie MSER spectrum is equel to twice the Equivalent Pluto Dose. Thus;
an EPD of 1.5 x 10®2 nvt 1n DFR, associated with what appears to be
pérmissible graphite dimensional changes, corresponds to a permissible
MSER dose (E > 50 kev) of 3 x 1022, The flux sbove 50 kev at any point
in an MSER core is very nearly proportional to the power density per unit
of core volume in the vicinity of that point. For an MSER with a central
pover density of 100 w/cc, the associated flux above 50 kev is about
k.5 x 10 neutrons/cm®-sec, vhich would produce a dose (E > 50 kev) of
sbout 1.1 x 10°2 in one year at 80% plant load factor. Thus, if the
permissible dose (E > 50 kev) is 3 x 10°2, and if the maximm power
density is 100 w/cm®, then replacement of at least a portion of the
graphite would be required at approximstely 2.T7-year intervals. Alter-
natively, if the average "core" power density is 80 w/cc and the power
peaking factor is 2, the time between graphite replacements would be
ebout 1.7 years.

 

Sprivate communication from E. Yoshikawa, Pacific Northwest
Laboratory, 1967.
 

 

 

 

 
 

 

 

1-50.

51.
o2.
23
5k,
53«
56.
57.
29.
600
61.
620
63,
6l .
65.
66.
67,
68.
69.
TO.
T1.
T2
T3
Th.

6.

81.

88.'

89.
0.
9.
92.
93.

960
970
98.

 

%

Internal D_istribution

MSRP Director's Office 99.

Bldg. 9201-3, Rm. 109 100.
R. K. Adams ‘ 101.
G. M. Adamson ' 102,
R. G. Affel ' 103.
L. G. Alexander 10k.
J. L. Anderson 105.
R. F. Apple 106.
C. F. Baes _ 107.
J. M. Baker 108.
S. J. Ball f 109.
C. E. Bamberger 110.
C. J. Barton ' 111.
H. F. Bauman 112.
S. E. Beall 113,
Ro Lo Beatty ’ lll'l'o
M. J. Bell ' 115.
M. Bender | ' 116.
C. E. Bettis ' - 117.
E. S. Bettis ‘ 118.
D. S. Billington . 119.
R. E. Blanco 120.
F. F. Blankenship - 121.
J. 0. Blomeke ' 122.
R. Blumberg | 123.
E. G. Bohlmann | 12k,
C. J. Borkowski 125,
G. E. Boyd 126.
C. A. Brandon 127,
M. A. Bredig 128.

“R. B. Briggs | \ 129¢
H. R. Bronstein 130.
G. D. Brunton 131,
D. A. Canonico _ 132.

So Cantor . 133-
R. S. Carlsmith 13k,
W. L. Carter o 135.
G. I. Cathers ) 136.
0. B. Cavin o 137.
A. Cepolino - 138.
W. R. Cobb | 139.
C. W. Collins | , 1%0.
E. L. Compere | 141,
K. V. Cook 142,
W. H, Cook . 143.

"D, F. Cope =~ 1k,
L. T. Corbin - 1bs.
W. B. Cottrell 146,
B. Cox . 147,

F.

D.
J.
D.
R.
c.
J.
S.
A.
I.
D'
WQ
J.
E.
D.
L.
J.

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H.
J.

_c.

R.
R.
A,
J.
L.
H.

B.

W.
A.
R.
RC
JQ

B.

D.

W.
c.

P.

R.
P.
D.

M.
H.
D.
P.
R.
T.
H.

L.
R.
M.
G.
J.
B.
H.
J.

'So
T,

A.
PO
R.
P.
E.
M.
E.
P,
A‘
E.
H.
Bl
E.

Culler
Cuneo
Dale
Davis
DeBakker
Deering
DeVan
Ditto
Dworkin
Dudley
Dyslin
Eatherly
Engel
Epler
Ferguson
Ferris :
Fox, AEC, Wash.
Fraas
Friedman
Frye, Jr.
Gabbard

‘Gallsher

Gelbach

Giembusso, AEC, Wash.

H.
0.
E.
L.
R.
G.
W,

H.

P.
A.
H.
G.

0.

S.

‘N.

E.
G.
N.
R.
R.

‘W,
'Ko'

P.
W
L.

Gibbons
Gllpatrick
Goeller
Greenstreet
Grimes -

‘Grindell

Gunkel
Guymon
Hammond
Hannaford
Harley
Harman
Harme -
Harrill
Haubenreich
Helms
Herndon
Hess
Hightower -
Hiil
Hoffman
Holmes
Holz
Horton
Hudson

Inouye
 

 

 

-1k9,
150-169.
170.
171.
172,
173,
17h.
175.
176.
177.
178.
179.
180.
181.
182,
* 183,

181'-0.

185,
186.
-187.
188,
189.
190-
191.
192 .

193,

19k
19%5.
196.
197 *
198.
199,
200.
201.
202.
203.
204,
205.
206.
207-208.
209.
210.
211,
212,
213,
21k,
215.
216 »
217.
- 218,
219.
220.

w.
P,
R.
M.
M.

- C.

T.
H.
SO
J-.
A.
T.
JO
c.
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W.
c.
J.
M.

R,

A,
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E.

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R.
H.
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D.
c.
T.

He

R'
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HI
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T.

L.

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Al
R.
SI
D'
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J.

Ho
R.
J.
To
Jo
R.

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W.
I.

S.

W.

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J,
E.

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B.
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I.

N.
L.
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C.
L.
Do
H.
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C.
F.
K.
Je.
Wa
Eo
R.
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J.
S.
Jde
IJ-
E.
Mo
R-
An
Po

Jordan
Kasten

Kedl

Kelley
Kelly
Kennedy
Kerlin
Kerr

Kirslis

Koger
Krakoviesk
Kress
Krewson
Lanb

Lene

Laxrkin
Larson
Lavwrence
Lin
Lindauer
Litman
Llevwellyn
Long
Lundin
Lyon
Macklin
MacPherson
MacPherson
Mailen
Manning
Martin

-‘Mauney

McLain
McClung
MceCoy
McCurdy
McDuffie
McGlothlan
McHargue

MeIntosh, AEC, Wash,

McNeese
McWherter
Merkle
Metz

Meyer

Miller
Moore

Moore

Moulton
Mueller

‘Nelms

Nichols

- E.

E.

L.

P.
Ac
T.
H.
B.
E.
G.
J.
Mo
Go
R.
Wc
Hc
K.
R.
Ho
So
Ho
A,
W.
C.

L.
D,
c.‘

Nicholson
Noguelra
Oakes

Patriarca

M.
W.
B.
E.
ED
L.
L.

Perry.

Pickel

Piper S
Prince o
Purvis, AEC, Wes
Ragan :
Redford

Richardson

D.
C.
c.
c‘
A.
G.
M.
R.
c‘
W.
F,
E‘

Robbins
Robertson
Robinson
Roller
Romberger
Ross

Roth,. ARC, ORO
Sepirie, AEC, ORO
Savage
Savolainen
Schaffer
Schilling

Dunlap Scott

J.
H.
c.
Je.
M‘
WI
J.
E.
M.
G.
W.
AO
F.
G.
00
P.
I.
R.
W.
H‘
RO
D.
R.
J.
E.
W.
R.

L.
E.
E.
H.

Scott
Seagren
Sessions
Shaffer

Shaw, AEC, Wash.

H.
M.
E.
J.
M.
L.
N.
JO
P.
LO
G.

Sides

Simmons, AEC, Wash.,
Sinclair, AEC, Wash.
Skinner

Slaughter -
Smalley, AEC, ORO
Smith

Smith

Smith

Smith

Smith

Spiewak

c.
c.
H.
A‘
A.
FC
R‘
H.

Steffy
Stoddart

Stone

Strehlow

Sundberg L
Sweek, AEC, Wash.
Tallackson
Taylor

Terry

E.

Thoma

 
 

 

 

273.
27h ®
275,
276.
277.
278.
279.
280.
281.
282,
283 .
28,
285,
286 [
287.
288 ®
289,
2% *
291 .
2% .
293-294,
295.296.,
297-299,
300.

301-315%
316.

 

97

P. F. Thomason
L. M. Toth

D. B. Trauger
J. S. Watson
H. L. Watts

C. F. Weaver
C. E. Weber, AEC, Wash,
B. H. Webster
A. M. Weinberg
J. R. Weir

W. J. Werner
K. W. West

M. E, Whatley

J. C, White

L. V. Wilson

Gale Young

H. C. Young

J. P. Young

E. L. Youngblood

F. C. Zapp

Central Research Library
Document Reference Section
Laboratory Records
Laboratory Records (LRD-RC)

External Distribution

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