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ORNL-4076

Contract No, W-7405-eng-26

REACTOR CHEMISTRY DIVISION ANNUAL PROGRESS REPORT
For Period Ending December 31, 1966

Director

W. R. Grimes

Associate Directers

E. G. Bohlmann
H. F. McDuffie
G. M. Watson

Senior Scientific Advisors

F. F. Blankenship
C. H. Secoy

MARCH 1967

OAK RIDGE NATIONAL LABORATORY
Quk Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION

: RN

RIIN MARTS .>Fl N5 J

e wmenc st o IHTIEINNY

3 4Y5L 0L3407Y 1
Reports previously issued in this series are as follows:

ORNL-2931
ORNL.-3127
ORNI.-3262
ORNL-3417
ORNL-3591
ORNL-3789
ORNL.-3913

Period Ending January 31, 1960
Period Ending January 31, 1961
Period Ending January 31, 1962
Period Ending January 31, 1963
Period Ending January 31, 1964
Period Ending January 31, 1965
Period Ending December 31, 1965
Contents

PART |. MOLTEN-SALT REACTORS

1. Phase Equilibrium and Crystallographic Studies

THE EQUILIBRIUM PHASE DIAGRAM FOR THE SYSTEM LiF-BeF -ZrF4
R. E. Thoma, H. A. Friedman, and H. InSley ..o v s e e
Investigations of the equilibrium phase diagrams of the systems LiF-BeF «ZrF and Bel -Zr[‘ were
completed. Both systems exhibit liquid-liquid immiscibility, behavior which has neretofore hecn cons

sidered to be very unusual in molten fluoride systems.

PRELIMINARY STUDY OF THE SYSTEM LiF—ThF4~PaF4

C. J. Barton, H. H. Stone, and G. D. Brunton ..o i it e et sttt e e anes e s enenee e sen st st s

Optical examination of two slowly cooled mixtures of LiF, ThF4, and Paf*‘4 indicates the probable

existence of the compounds LiPa¥F _ and Li4PaF8.
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APPARATUS FOR DIFFERENTIAL THERMAL ANALYSIS

L. O. Gilpatrick, R. E. Thoma, and S. CAntOr ... .o iieimieiiitcont i nes sasies et smie sreecs sars seesssseniassies sanssasssrnens

Automatic DTA apparatus was developed, tested, and found to be suitable for the study of phase

transitions in mixtures of fluoroborate salts.

SOLID-PHASE EQUILIBRIA IN THE SYSTEM Srn)':2~.'§mF3-UF3
R. E. Thoma and H. A, Friedman .. et .
Extensive mutual solidestate solubility of components and intermediate phases were found in the

Sml}‘ZuSmFB-UF3 system.
PHASE RELATICNS IN THE SYSTEM I(F»CeF3

C. J. Barton, G. D, Brunton, D. Hsu, and H. INS1EY .o i i e e et e i et e e

An incomplete investigation of the system KF-CeF3 showed the existence of one esutectic composition

and two incengruently melting compounds, SKFP'CEI?s and KF~CeF3.

THE CRYSTAL STRUCTURE OF L:4UF

G. D. Brunton

The U*T ion in this structure is surrounded by ¥ ions at the corners of a 14-faced polyhedron. The Li-F
coordinations are irregular octahedra, two of which share faces with the U‘Vr polyhedron.
THE CRYSTAL STRUCTURES OF NarF'-!._uF3 SOLID SOLUTIONS
D. R. Sears and G. D. BIUNTON o et ittt it e ar st e et e as e ta e e rrtes et et
Crystal structure analyses of two specimens near the 50:50 composition are described. A model
structure based on that of CaF2 with cation vacancies and anion interstitials appears to fit the intensity

data best, but there are anomalies in the thermal motion.

THE CRYSTAL STRUCTURE OF }/nCsBeF3
H. Steinfink and G. D. Brunton

This structure is similar to that of the high-temperature form of BaGeOs.

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THE CRYSTAL STRUCTURE OF f3-KLaF ,

D R S IS ittt ittt ettt ettt ottt et et een tes de e eese ittt atatne deeiaea et eneate teeea e et et ahties et eeaentenn et naes  an et ttne e tten e te e anenar nrneas

This compound forms merohedral twins which are almost isostructural with NaNdF4.

CENTRAL CATION DISPLACEMENTS IN THE **TRIPYRAMIDAL® COORDINATION

Interatomic potential calculations suggest which anion configurations favor displacements of the cation

from the medial plane and are correlated with the structures of ,81-KLaF4 and NaNdF4.

PREPARATION OF FLUQORIDE SINGLE CRYSTALS FOR RESEARCH PURPOSES

R. E. Thoma, R. G. Ross, and H. A. Friedman .. .......ciiiiiiiii s et e et e At v EiEeE v e e rEr reean e tarr e vaan

e : 7y : , 714 NaT .
Pure single crystals of the fluorides ‘LiF, L12B8F4, Na 7Zr6F31, L12Na"1h2F11, B CsBer, and

LictUFB were grown from the melt and furnished for use in research programs at ORNL and elsewhere.

2. Chemical Studies of Molten Salts

A POLYMER MODEL FOR Lii’f-BeF2 MIXTURES

C. F. Baes, Jr cciiiiiieeniee e e et ettt e et et eeeae e eee et eeeees—ameeseeee teeetaeanetasedetetetaretan taoeten et ane srraee nerndneate e et rirarat ataan on

A model which assumes polymeric Bean(b-za)- ations containing BeFdz" tetrahedra, bridging F™™, and

terminal F~ ions is found to be consistent with measured activities of BeF‘z.

PHASE EQUILIBRIUM STUDIES IN THE UOZ-ZrO2 SYSTEM

K. A. Romberger, C, F, Baes, Jr., and H. H. Stone ... i e e e

New results from equilibration of the oxides in the presence of molten fluorides, while confirming the
nearly complete exsolution of the oxides from one another below the eutectoid at 1110°C, show evidence

of nonideal behavior in the dilute solid solutions just above this temperature.

THE OXIDE CHEMISTRY OF ThF4-UF4 MELTS

B, F. Hitch, C. E. L. Bamberger, and €. F. Baes, Jlu i e e et e et s a e b aen srrn s e

The oxide phase at equilibrium with 2LiF-BeF2+UF4+ThF4 is a (U-Th)O2 solid solution into which

the uranium is strongly extracted.

THE OXIDE CHEMISTRY OF LiF-BeFZ-Z'rF4 MIXTURES

B. F. Hitch and €. F. Baes, JIu ittt s ceetteties tieeartes st ataastaateaaste surs seatne sens s mees mar s rnees otaeeemaensnan aas reamee aerens e e

Measurements of the solubility of BeQ and Z'rO2 indicate the oxide tolerance of MSRE flush salt and

fuel salt.

CONSTANT-VOLUME HEAT CAPACITIES OF MOLTEN SALTS

AN L EY A T oot iiiitiie ittt ettt e et et e et et e et e et e e e e e e et e n e e i T e e e rE by e pa s e aeanan et e

Values of CV were obtained by combining published Cp values, sonic velocities, and density-tempera-=
ture data; in almost every case experimental CV exceeded that calculated on the basis of simple clas-

sical and/or quantum contributions.

TEMPERATURE COEFFICIENT OF Cv FOR MOLTEN SALTS

SN L ANt OT Lot iiiiiies ittt e it et iees e e soettrbbes riettere e et em e e e e e tee s rae s ee et e eee e eeae fane eeean eeeea e teea e teetateeearrtnr s atans aetanenrenbeas o

By using an empirical equation in which compressibility is a linear function of pressure, the temperature
dependence of CV for 34 salts was calculated; where compression was necessary to sustain a fixed volume,

CV increased with temperature.

TEMFERATURE COEFFICIENT OF COMPRESSIBILITY FOR MOLTEN SALTS

BEANTEY CAIIOT Lot it et e et et e e et eeee et e eee e es et e et eaae taeire e ore et e eeeaeeeee et s ereeea s ety beereen e rn i aean s

A simple empirical equation, ,8,?, == AebT, was found to hold fer all molten salts. (Bg is the iscthermal

compressibility at 1 atm, A and b are constants, and T is the absolute temperature.)

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VISCOSITY AND DENSITY IN THE LiF-B\eF2 SYSTEM

C. T. Moynihan and Sanl ey Camtor i it ettt ettt itet e e et eae tenaee e reeene teeene catean e men eeneteenant s

Viscosity and density measurements show that the temperature coefficient of viscosity decreases when
the volume expansion coefficient increases; the volume expansion coefficient is directly correlated to the
temperature dependence of “*free*’ volume in these melts.

VAPQOR PRESSURES OF MOLTEN FLUORIDE MIXTURES

Stanley Cantor, W. T. Ward, apd e B RO IS Lottt it ettt e et te et e e e eat e et e n e aeeeseentraes et aran et aann

Vapor equilibria that are involved in the reprocessing by distillation have been measured. Decontamina-
tion factors of the order of 1000 for rare earths were evidenced. The vapor pressure of the composition of

MSKE fuel concentrate was also measured.

POTENTIOMETRIC MEASUREMENTS IN MOLTEN FLUORIDES
A. R. Nichols, Jr., K. A. Romberger, and C. F. Baes, Jr. i i i et s e eeneramen e e e

Preliminary results for niobium in 2LiF-BeF2 indicate the formation of stable, insoluble NbFZ.

APPEARANCE POTENTIALS OF LITHIUM FLUORIDE AND LITHIUM BERYLLIUM FLUORIDE IONS

R. A. Strehlow and J. D. IREAMAD ..ottt ie ettt aes s araes smes st ae 2 etb s s ee st etn as e anean eers mree aeaers AL narnas ot L

A study was made of appearance potentials of jong formed by electron impact from LiF and Lif)BeF‘4

vapor, and a surprising amount of structure was found in the ionization efficiency curves.

3. Separations Chemistry and Irradiation Behavior

REMOVAL OF IODIDE FROM LiF-BeF2 MELTS
C. E. L. Bamberger and C. F. Baes, Jr.
The efficiency of HF utilization during sparging shows an increase with decreasing pressure caused

either by as vet unidentified side reactions or by a rate effect.

REMOVAL OF RARE EARTHS FROM MOLTEN FLUORIDES BY SIMULTANEQUS PRECIPITATION
WiTH UF3
F. A, Doss, H. F. McDuffie, and J. H. Shaffer ... i
Expasure of LiF-—Ber (66~34 mole %) containing about 107 % mole fraction CeF3 or NdF | to excess solid
UF3 caused removal of the rare earths from the molten solution, '
EXTRACTION OF RARE EARTHS FROM MOLTEN FLUORIDES INTO MOLTEN METALS
J. H. Shaffer, W. P. Teichert, D. M. Moulton, F. F. Blankenship, W. K. R. Finnell, W. R. Grimes . ..............
The distribution of rare earths (lanthanum, cerium, neodymium, samarium, and europium) between molten
Li.T*‘-Ber (66-34 mole %) and molten bismuth was studied at 600°C as a function of the concentration of

lithium netal added as a reducing agent.

REMOVAL OF PROTACTINIUM FROM MOLTEN FLUORIDES BY REDUCTION PROCESSES

The removal of protactinium from a simulated molten-salt breeder reactor blanket was demonstrated in a
six-week experiment in which liquid bismuth was recirculated through the blanket salt, a bed of steel wool,
and a bed of thorium metal chips. The evidence suggested that the protactinium was transported as a4 sus-

pension, perhaps associated with high«melting metallic compounds of iron, chromium, and thorium.

REMOVAL OF PROTACTINIUM FROM MOLTEN FLUORIDES BY OXIDE PRECIPITATION
J. H. Shaffer, W.:P. Teichert, W. K. R. Finnell, F. ¥. Blankenship, and W. R. Grimes ... ...
The removal of protactinium from solution in molten LiF-BeF2 (66-34 mole %) by oxide precipitation upon
the addition of Zr02 at 600°C was studied with a variety of Zr(_)2 powders of differing surface areas (from
1.3 to 8O mZ/g). The results were not consistent with simple theories of either complete solid solution for-

mation or precipitation on the surface of the ZrO,‘2~

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PROTACTINIUM STUDIES IN THE HIGH-ALPHA MOLTEN-SALT LABORATORY
C. J. Barton and H. H. SEOME Lt et i it ettt ettt st e sttt e e o e bt ben abare s eaasstaes bt bbb et s e a e ae e ae o
Reduction of protactinium dissolved in a molten LiF-ThF4 breeder blanket mixture by exposure to solid
thorium, followed by adsorption of the reduction product on an iron surface, is the most promising of the

several recovery methods studied to date.

GRAPHITE-MOLTEN-SALT IRRADIATION TO HIGH FiSSION DOSE
H. C. Savage, J. M. Baker, E. L.. Compere, M. J. Kelly, and E. G. Bohlmann .........c.cccei i
Irradiation of the first molten-salt thermal convection loop experiment in the ORR was terminated August
8, 1966, because of a leak through a broken transfer line after achieving power densities of 105 W/cm3 in
the fuel channels of the graphite core. A second loop, modified to eliminate causes of failure encountered

in the first, begins long-term irradiation in January 1967.

4. Direct Support for MSRE

EXTENT OF UF4 REDUCTION DURING MSRE FUEL PREFPARATION
B. F. Hitch and C. F. BAES, JIv i i it e et i e e e s et e e e s et e e e e s e e s
It is estimated that 0.16% of the uranium introduced into the MSRE had been reduced to UFS during salt

purification.

CHEMICAL BEHAVYIOR OF FLUORIDES DURING MSRE OPERATION
) S L I U o - TSP P PRSP
Current results of the chemical analyses of MSRE fuel, flush, and coolant salts show that after approxi-
mately 20 months in the MSRE, the molten salts have retained their original chemical composition and have

not induced perceptible corrosion in the reactor.

FISSION PRODUCTS IN MSRE FUEL
S. S. Kirslis and F. F. BlankensShip ..o it e iiiris iee reeavte it eimt s ietas et s e atetite sareaeness erntseaats roee sesetbnate sesnss sars o1
Radiochemical analyses for fission products in MSRE fuel salt samples indicated that appreciable
fractions of the 99Mo, 132’Te, 1c's’Ru, and 1%0Ru produced by fission had left the fuel phase.

FISSION PRODUCTS IN MSRE EXIT GAS
Equilibrium Pressures of Noble-Metal Fluorides Under MSRE Conditions
G B B, T ittt i it e et e e eee i L b e e e e b e e eL s Le s b s L b deeeee s h e ee e an e e e e e ne s tateeeaten beba o
Thermochemical data indicate that, with increasing oxidizing power, the order of appearance of volatile
fluorides should be NbFS, MOFG, RuFS, TeFS.

Analysis for Fission Products in MSRE Exit Gas
S. S. Kirslis and F. F. BlanKkenShip ciiiiciome i it ciiessiniecnms seemsssmmeocssescssnrs somesssaeesenestsees sieees sorsssasns ssnsssnassasssssanss oo
Small metal samples exposed to the gas phase of the MSRE pump bowl demonstrated qualitatively an ap-

preciable volatility of 9QMO, 132’1‘6, 103Ru, and 19%Ru presumably as high-valent fluorides.

FISSION PRODUCTS ON METAL AND GRAPHITE FROM MSRE CORE
S. S. Kirslis and F. F. BlankensShip ..o oo it eiieetirt et cres ceeetians sreresins sermrs saie s sere thetenssanenensmne essssen sasons vs os
Samples of MSRE graphite removed from the reactor core after 7800 Mwhr of operation showed no radiation
damage effects but were found to be significantly permeated or plated by noble-metal fission products and
those with noble-gas precursors. Adjacent Hastelloy N samples were also undamaged and were more heavily

plated with noble-metal fission products.

XENON DIFFUSION AND FORMATION OF CESIUM CARBIDE IN AN MSBR
C. F. Baes, Jr., and R. B, Evans TII i e e e e
Carbide formation in the moderator graphite should occur, but not in significant amounts; 135xe poisoning
could be reduced effectively either by iodine removal or by some means which reduces the salt-graphite film

coefficient.

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PART 1. AQUEOUS REACTORS

K. Cotrosion and Chemical Behavior in Reactor Environments

NASA TUNGSTEN REACTOR RADIATION CHEMISTRY STUDIES
G. H. Jenks, H. C. Savage, and E. G. BORIUALI .o iieiieies cer s cireaneiettsaes aess s raeemmesamnaee te s eesnm st a2 mnane aa s 57
Experimental results showed that electron irradiation produces a small loss of cadmium from CdSO‘{l solu-

tions under conditions of interest in the NASA Tungsten WatersModerated Reactor. Equipment was designed

for additional studies of the effects of agitation on the radiation stability of the solution.

CORROSION OF ZIRCALOY-2 BY DILUTE HYDROGEN PEROXIDE AT 280°C
R. J. Davis, T. H. Mauney, and R. J. Harl i et et saries s e s ras aeeees tmeeararas e enrass s neeatess sbesas ae 58
The corrosion of Zircaloy-2 in oxygenated water at 280°C was shown to be unaffected by the presence of

107° M 11202, and it was concluded that the radiation effect on zirconium=-alloy corrosion in these solutions

is not a direct result of the peroxide formed during irradiation.

ANODIC FILM GROWTH ON ZIRCONIUM AT ELEVATED TEMPERATURES
A. L. Bacarella, H. 8. Gadiyar, and A. L. SULTOM Lo i i i s et treaas e raes o ts Sammaeuntaes ossnas e ansntrsnunss 2eee on 58
A new expression for the anodic film growth current on zirconium was derived using the triple-barrier

model with a field-dependent activation distance in the oxide phase, and our experimental data were fitted

with this expression,

AU IMPEDANCE OF OXIDE FH_MS IN AQUEQUS SOLUTIONS AT ELEVATED TEMPERATURES
G. H. Jenks, A. L. Bacarella, R. J. Davis, and H. 5. Gadiyar ...t et e 61
FKquipment, methods, and techniques are being developed and tested for measuring ac impedance of cor-

rosion films on zirconium alloys in agueous solutions at elevated temperatures. The immediate objective of

such measurements is the detection of film porosity.

CORROSION SUPPORT FOR REACTOR PROJECTS
J. C. Griess, Jr., J. L. English, and P. D. NeUMANN e e smenis s i e e meianee s soes omire s sae s 63
Corrosion investigations conducted for selecting structural materials for use in the High Flux Isotope Re-

actor and the Argonne Advanced Research Reactor were completed. Generally, both reactors should operate

many years without major corrosion problems providing the chemistry of the coolant is properly maintained,

6. Chemistry of High-Temperature Aqueous Selutions

ELECTRICAL CONDUCTANCES OF AQUEOQUS ELECTROLYTE SOLUTIONS FROM 0 TO 800°C AND TO
4000 BARS
A. S. Quist, W. Jennings, Jr., and W. L. Marshall ... i e e i e e e e 65
Coentinuing, extensive conductance studies on aqueous electrolytes to 800°C and 4000 bars have provided
limiting eguivalent conductances and dissociation constants of sodium chloride, differing sharply from bLe-

havior at 25°C, and measurements on 16 other 0.01 m 1~1 electrolytes.

DISSOCIATION CONSTANT OF MAGNESIUM SULFATE TO 200°C FROM SOLUBILITY MEASUREMENTS
B, L A ALl it it miciries cre tats srries mr e n e sereet R na Ao eeenen s ean e neen g ries faee ssageneene teaaneseen ena g aatentan ntrere b na e senean 66
From the differences in solubility of calcium sulfate in sodium chloride and in sea-salt solutions, dis-

snciation gquotients, constants, and other thermodynamic quantities have been calculated.

DISSOCIATION CONSTANT OF CALCIUM SULFATE TO 350°C OBTAINED FROM SOLUBILITY
BEHAVIOR IN MIXED ELECTROLYTES
L. B. Yeatts and W, L. Marshall i et i e e e men e et eass saemts samcabntas senseia e s s s ne s e et anns enne s 68
In perhaps the first extensive study of a four-component, mixed electrolyte system to high temperatures,
solubilities of calcium sulfate were determined from 25 to 350°C from which dissociation quotients, solu-

bility products, their respective constants, and thermodynamic quantities were calculated.
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SOLUBILITY OF Fe304 AT ELEVATED TEMPERATURE
F. H. Sweeton, R. W. Ray, and C. F. Baes, ]Jr.
The soclubility of Fe‘304 in dilute HCI solutions containing dissolved H2 has been measured at 200, 260,

and 300°C, and solubility products for formation of Fe2+ and FeOH™ have been calculated.

HYDROLYSIS OF BERYLLIUM ION IN 1.0 M CHLORIDE AT 25°C
R. E. Mesmer and . F. Baes, Jru ot ittt e ettt bt eees eesmas = aeaas eae s e ean sas o ebs e aesans men s sbe ae e snbn mans ae
The previously reported hydrolysis schemes for beryllivm are not fully supported by the present data at

25°C. The uniqueness of other possible schemes is being tested.

7. lInteraction of Water with Particulate Solids

SURFACE CHEMISTRY OF THORIA
s H o S B OOy iiiitiiiis it tirt et teee et e e et tee et eeee eete e e ete beeeeettbn etnes eeeies hes eaeeneenre et eemetneae e ea e teee baen e etee e e een i e eareaneeneteaen e

Heats of Immersion and Adsorption
E. L. Fuller, Jr., H. F. Holmes, amd S, AL T Ay lor oot oo i et ee s ot it r ettt e ete stes s reransin e e antesrs sebtrsrannsnrenesren
Thoria powders composed of crystallites with an average size greater than about 1400 A yvield a constant
amount of heat per unit surface area upon immersion in water after outgassing at a given temperature. Powders
composed of smaller crystallites react more energetically and release a portion of the heat by kinetically slow

processes.,

Adscrption of Water and Nitrogen on Porous and Nonpoious Thoria
H. F. Holmes and E. L. FULLer, Jr. o oo coeet e e o ettt eeee seeaes e atn sen s e eae sate s s eetasent s seerenmaeemenneaens
The concept that chemisorbed water decreases the pore velume is not adequate to explain the observed
decreases in nitrogen and water surface areas of nonporous thoria, nor is the smaller size of the water
molecule compared with nitrogen consistent with the observation of water areas much smaller than nitrogen

arcas.,

Infrared Spectra of Adsorbed Species on Thoria
C. 5. Shoup, Jr.
Infrared spectra of the ThOQ-HQO interface obtained by both adsorption and desorption have confirmed the

nonequilibrium nature of the surface interactions of thorinm oxide and water.

BEHAVIOR OF GASES WITH SOL-GEL URANIUM-THORIUM OXIDE FUELS
D. N. Hess, H. . McDuffie, B, A. Soldano, and C. Fu Weaver e e e et e et et e e
The gases released when sol-gel microspheres of ThO2 or UO2 were heated in vacuum were identified,
and the temperatures of maximum gas evolution were established. A conditioning procedure was developed
which, when applied to wet, unfired microspheres, converted them into satisfactory reactor-fuel-element

products of high density, low carbon content, and low O:U ratic.

PART lll. GAS-COOLED REACTORS
8. Diffusion Piocesses

TRANSPORT PROPERTIES OF GASES

Gaseous Diffusion Studies in Noble-Gas Systems
AL P MalinausKas ..o e e et et o s e e e

Diffusion data are reported for the systems He=Kr, Ar-Kr, and Kr=-Xe over the temperature range 0 to 120°C.

Thermal Transpiration
B. A. Cameron and A. P. MalinausKas ... i it et e s ce et et ee e e e tee s avae e ein e nen s .
Thermal transpiration measurements using a porous septum have been attempted., Although steady-state
conditions are attained very rapidly, the thermal conductivity of the gas now enters in a pronounced manner

and causes the analysis of the data to be extremely difficult.

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Gaseous Diffusion in Porous Media
A. P, Malinauskas, R. B. Evans III, and E. A, MaSOI (i e i vins siesremman aetsaisass attsetaenass sors sssntstnasssesasrsssions s - 85
A generalized treatment of gas transport in porous media has been developed on the basis of the *‘dusty-

gas’ model.

Gus Transport Studies Related to Vented Fuel Elements for Fast Gos-Cooled Reactors
R. B. Evans IT]T and D. B BIUiiS .ot i oot iiie iiieie e re e eeiis sies st s amke temee s sanam e = raee arnmte s bebse st sane abettasnrsaharebant ataans nars o0 36
An investigation of the possibility of using direct venting devices on fuel elements in fast gas-cooled re-

actors has been initiated.

RECOIL PHENOMENA IN GRAPHITES
R. B. Evans IIl, J. L. Rutherford, and R. B. Perez ..., et et es hatetstee e eanhy A saae e e nn s aes e e e e en e an 86
The effects of density and porosity of graphitic structures on the range of “light’’ and “*heavy?® fission

fragments have been determined.

9. Behavior of Graphite with Reactive Gases

1.. G. Overholser

OXIDATION OF GRAPHITE SLEEVES BY STEAM
Co M. Blood ant G. M. HEDEIt v iiiriirrsrrreeas covsns sersareses sars vsrers sre s sssainesese s smees pasasean sameseasss sessonsonsaonsshaiss sansanose beaassas 92
Oxidation rates of virgin, impregnated, and irradiated AT] graphite sleeves were measured at 1000°C

using a partial pressure of water vapor at ™ 250 torrs.

TRANSPORT OF FISSION PRODUCTS
o M. B0 i ittt et et et e e e aee et eea e eees e aeee f e ua e een e Es T en e et e e §enara e eeuiaere dtne erese o ebs tnaaen seabee shusnreenaee s neaas 93
Deposition profiles for (1) 133Ba transported from barium-impregnated graphite by wet or dry helium and (2)
“‘0Ag, 137, and 13%*Cs transported from previously irradiated graphite by wet helium were established by

sectioning and counting techniques.

OXIDATION OF COATED FUEL PARTICLES BY WATER VAPOR
o oy BaKET oottt et et e e e e ek aee ee e e oaeE A ke e e ere fuEE R RS e e g r gL TSN raapenEReeans hessaareee sans b os 96
Rates of oxidation and incidence of coating failures were determined for various batches of coated fuel

particles at 1100 to 1400°C using helium-—water-vapor mixtures containing 500 or 1000 ppm of water vapor.

10. lrradiation Behavior of High-Temperature Fuel Materials

O. Sisman and J. G. Morgan

IRRADIATION EFFECTS ON PYROLYTIC-CARBON-COATED FUEL PARTICLES
P. E. Reagan, J. G. Morgan, J. W. Gooch, M. T. Morgan, and M. F. OSDOINE .. .ooiiiiiiis iceviii i sieeiin e enns . 99
Pyrolytic-carbon-coated thorium-uranium carbide particles prepared commercially for the German AVR re-

actor withstood irradiation to 10 at. % heavy-metal burnup at 1300°C, and a barrier layer of silicon carbide

added to a pyrolytic carbon coating greatly reduced the release of fission solids.

IN-PIL.E TESTS OF A MODEL TO PREDICT THE PERFORMANCE OF COATED FUEL PARTICLES ,
P. E. Reagan, E. L. Long, Jr., J. G. Morgan, and J. W. GOOCI ... it i it e e s e e cn v paee s 100
A mathematical model developed to predict the bumup necessary to cause pyrolytic-carbon-coating

failure was found to be accurate for the weakest coatings in the batch, and a thick carben buffer layer

caused uranium oxide particles to overheat and attack the coating.

POSTIRRADIATION TESTING OF COATED FUEL PARTICLES
M. T. Morgan, C. D. Baumann, and R. L. Towns
Various types of pyrolytic carbon coatings applied to fuel particles of UO2 and UC2 have been annealed

at high temperatures after neutron irradiation to test for coaling stability, retention of fission products, and

fuel migration.
IRRADIATION EFFECTS ON COMPATIBILITY OF FUEL OXIDES AND BERYLLIUM OXIDE WITH

GRAFPHITE
D. R. Cuneo, C. A. Brandeon, H. E. Roebertson, and E. L. Long, Jr. .o e e e, 105
Graphite is chemically compatible with both (U,Th)O2 and Be(Q; the concentration of 61.i found in BeQ

diminishes in smaller pieces in a direction consistent with a surface-to-volume relationship.

FAST GAS-COOLED REACTOR DEVELOPMENT
D. R. Cuneo, H. E. Robertson, E. L.. Long, Jr., and J. A. Conlin . i i e eeet o ee i aeee e s 107
In low-burnup irradiations no indication of fuel element failures have been found with UO2 in either

stainless steel or Hastelloy X,

FiSSION-GAS RELEASE DURING FiISSIONING OF U02
R. M. Carroll, R. 3. Perez, O. Sisman, G. M. Watson, and T. W. FUllon .o e e e 109
Refinements have been made in the defect-trap model, and clustering of defects at about 1000°C in single-

crystal UO2 was observed as predicted.

THERMAL CONDUCTIVITY OF UOZ DURING IRRADIATION
C. D. Baumann, R. M. Carroll, J. G. Morgan, M. F. Osbome, and R. B. Perez . ...ooccoviiiiiiii et inineniie e 111
The thermal conductivity of a UO2 fuel specimen is being measured as a function of flux and temperature

during irradiation.

11. Behavior of High-Temperature Materials Under liradiation

EFFECTS OF FAST-NEUTRON IRRADIATION ON QXIDES
G. W. Keilholtz and R. E. MOOT@ . ot e e s e e et e e e e e e e e e e e e s ot e st e e ee e 113
Translucent aluminum oxide of high density has been found to be more resistant to irradiation damage than

sintered 11\1203 at irradiation temperatures of 300 to 600°C up to 3 X 1021 110utr0ns/cm2 (>1 Mev).

BEHAVIOR OF REFRACTURY METAL CARBIDES UNDER IRRADIATION
G. W. Keilholtz, R. E. Moore, and M. F. OSDOIme i e e e e e e e e e 114
Irradiation effects on specimens of monocarbides of Ti, Zr, Nb, Ta, and W made by hot pressing, slip

casting, and explosion pressing were investigated at low temperatures (300 to 7000(3) over the fast-neutron

dose range 0.7 to 5.4 x 1021 neutions/cm? {>1 Mev);, W and Ti monocarbides were quite resistant to irradia-

tion under these conditions.

PART IV. OTHER ORNL PROGRAMS
12. Chemical Support for the Saline Water Program

SOLUBILITY OF CALCIUM SULFATE IN SEA SALT SOLUTIONS TO 200°C; TEMPERATURE-

SOLUBILITY LIMITS FOR SALINE WATERS
W. L. Marshall and Ruth SIusher ...........coooomiioiiiiesienn. e et oo e, 119
Solubilities of calcium sulfate were determined in s=a salt solutions from 30 to ZOOOC, and the data wers

used to calculate revised temperature-solubility limits for saline waters in general.

CORROSION OF TITANIUM IN SALINE WATER
E. G. Bohlmann, J. F. Winesette, J. C. Griess, Jr., and F. A. P OSeY i e et e e e v 121
Continuing electrochemical studies of titanium corrosion at elevated temperaturcs have supported the acid
solution crevice corrosion mechanism and suggested that the complex inverse temperature dependence of the

pitting potential is related to effects of alloy constituents on the passive oxide film.
xXi

13. Effects of Radiation on Organic Materials

W. W. Parkinson and O. Sisman

EFFECTS OF RADIATION ON POLYMERS
W, W. Parkinson and W. K. Kirklamd ... i ottt e e te et e e e arre s eereree e e reue ubaae thes tarane e een naaees s 125
The olefin groups of all isomeric forms of butadiene decrease rapidly upon irradiation, with the dis~

appearance of side vinyl groups showing a high enough rate to suggest a chain reaction.

RADIATION-INDUCED REACTIONS OF HYDROCARBONS
K. M. Keyser and W. K. Kirkland ... i cernens e e vy e en e Ees e reee e eeabbaan e h et s aey e e 127
The nonvolatile products from the irradiation of the model system naphthalene in hexane were found by
chromatographic and spectral analysis to be chiefly &~ and fl-alkyl-substituted naphthalenes (assuming no

decomposition on the chromatographic column).

ADDITION REACTIONS OF FURAN DERIVATIVES
C. T, Bopp antd W, W, P aUKIIISOM (oot i e oot it ee ettt e e eeee et ess tieaes e e e et s taeaes 2 tns et s taan trasnnan e anaaenaretanannn 129
The major radiation products from solutions of cyclohexene in tetrahydrofuran have been tentatively

identified as 1:1 adducts and dimers, with yields ranging from G = 0.5 to 2 at room temperature.

DEVELOPMENT OF RADIATION-RESISTANT INSULATORS
W. W. Parkinson, B, J. Sturt, and BE. J. B enedy .o oo iee e e e tvtere st e rr e etrrr et e aren e ran s ana e naaens 130
Many samples of styrenesbase polymers and samples of several other chemically simple plastics have been
obtained and analyzed for common impurities, and an electrical measuring apparatus has been tested for

sensitivity.

14. Chemical Support for the Controlled Thermonuclear Program

2. A. Strehlow and D. M. Richardson

INTERPRETATION OF DCOX-2 MASS SPECTRA ittt s v i e e et e sttt res e e e oot seeetssrta g amssne cnnsmne seeees baaenees . 131
The composition of residual gas in the DCX-2 vacuum system was analyzed in detail from mass spectra ob-

tained during operation. Several low-molecular-weight hydrocarbons were found to be generated during beam

injection.

MASS SPECTROMETER CALIBRATION STUDIES ... et eehe ettt v eae e tretnesate e tas tenane eneaaasaaaae caenan e 133
Improvement was made in quantitative interpretations of residual gas spectra by studies of mass dis-

crimination in the spectrometer; the observed transmission fraction of carbon dioxide ions was one-fourth that

of water ions.

WATER VAPOR CHEMISORPTION ON STAINLESS STEE L e et et ettt e et v et e 134
The desorption of water from stainless steel after short exposures below 10~% torr was studied in an oil«

free system; the results followed chemisorption kinetics.

DECOMPOSITION OF DC-705 DIFFUSION PUMP FLUID ... et e csee s e 138
A white solid accumulation found in the inlet of a diffusion pump was identified as the decomposition

product of the silicone oil pump fluid.

PART V. NUCLEAR SAFETY
15. Activities of Nuclear Safety Technical Staff

W. . Browning, Jr., M. H. Fontana, and B. A, Soldano ... e e e e e 143
The Nuclear Safety Technical Staff, comprised of three persons, was formed early this year to aid in
planning, coordinating, and directing the research and development activities within the Nuclear Safety

Program.
Xii

16. Correlations of Fission Product Behavior

THE LIGHT BULB MODEIL. FOR RELEASE OF FISSION PRODUCTS
G Bl ML, hu ettt ettt ettt oot et v v et e s E e bt e re et e et e et et et ae e et r e et e e aer et et 145
A medel, based on boundary layer diffusion processes, satisfactorily describes the dependence of the

fraction of fission product released from reactor fuels on (1) the composition and pressure of the surrounding

atmosphere, (2) the temperature, (3) the heating time, and (4) the chemical fonn of the fission product species.

EFFECT OF CONTAINMENT SYSTEM SIZE ON FISSION PRODUCT BEHAVIOR
G. M. Watson, R. B. Perez, and M. H. F oontama it e e e e e e e 147
The behavior of iodine in containment systems differing in size by two orders of magnitude has been cor-

related with moderate success using simple mathematical relationships.

CHEMICAL EQUILIBRIUM STUDIES OF ORGAMIC-IODIDE FORMATION UNDER NUCLEAR REACTOR
ACCIDENT CONDITIONS
R. H. Bames, J. F. Kircher, and C. W. T omiiley i it st e et e e e e e e 149
Computerized thermodynamic calculations indicate that there are realistic conditions under which CH 3I

could be generated in reactor accidents.

THE ADEQUACY OF STALEUP IN EXPERIMENTS ON FISSION PRODUCT BEHAVIOR IN REACTOR

ACCIDENTS
C. E. Miller, Jr., and W, . Browiinmg, J o et i ettt et et eee et e e e e e haete e vababs bbe s s ran b ese s barn s s benane s rnarans s 150
A report has been written which describes two possible intermediate-scale experiments at 1 and 10% the

size of LOFT which are needed to extend the scaling range of experiments on fission product behavior over

the five orders of magnitude between small experiments and LOFT.

17. Nuclecr Safety Tests in Major Facilities

FISSION PRODUCTS FROM FUELS UNDER REACTOR-TRANSIENT CONDITIONS
G. W. Parker, R, A, Lorenz, and J. G. Wile i (i e ettt et ee e s ettt e e e e e e et eee s 152
Studies of fission product release and transport from metal-clad UO2 fuel transient-melted under water in-

clude the effect of and pressure rate of steam release.

SIMULATED LOSS-OF-COOLANT EXPERIMENTS iN THE OAK RIDGE RESEARCH REACTOR
C. E. Miller, Jr., R. . Shields, B. F. Roberts, and R. J. DaAVIS ..o eeie e et eeeie s ceeeaertebeeraveean o 153
The interpretation of data from previous experiments on fission product release and a literature review of

iodine deposition have been the main activities during a period when major construction work has been under

way on the reactor facility.

IGNITION OF CHARCOAL ADSORBERS
C. E. Miller, Jr., and R. P, Shields . e e et et et e e e e et et vaae e 155
Results of both in-pile and ocut-of-pile experiments on ignition temperatures of charcoal used in containe

ment vessel air cleaning systems show that the temperature can be affecied significantly by long-term ex-~

posure, slightly by moisture, and very little by adsorption of excessively large quantities of iodine.

FISSION PRODUCTS FROM ZIRCALOY-CLAD HIGH-BURNUP UOZ
G. E. Creek, R. A. Lorenz, W. J. Martin, and G. W, Parker e e e et et erenen s 157
Zircaloy-clad UO2 trradiated to a burnup of 7000 Mwd/ton was melted in the Containment Mockup

Facility (CMF), and the behavior of released fission products in the stainless steel CMF tank was compared

with that released from a stainless~steel-clad specimen irradiated to 1000 Mwd/ton,

BEHAVIOR OF i2 AND H!' IN THE CONTAINMENT RESEARCH INSTALLATION TANK
G. W. Parker, W. J. Martin, G. E. Creek, and N. R. HOrt0n . o i et e e et r et s e et e r e are s 158
The relative deposition behavior of molecular iodine and hydrogen iodide has been observed in the 1200-~gal

CRI stainless steel containment vessel,
xiii

18. Laboratory-Scale Supporting Studies

DEVELOPMENT OF FILTRATION AND ADSORPTION TECHNOLOGY
R. E. Adams, Jack Truitt, J. S. Gill, and W, I3, YUIlle i et et e e e s
The effect of accident environments on the behavior of test aerosols and on the performance of filter

media is being studied in the laboratory.

EXAMINATION OF PARTICULATE AEROSOLS WITH THE FIBROUS-FILTER ANALYZER
M. D. Silverman, Jack Truitt, W. E. Browning, Jr., and R. E. Adams ... e
The fibrous-filter analyzer is being developed as a device for examining the characteristics of radioactive
aerosols in terms of particle response to the major processes of filtration: diffusien, interception, and
inertial impaction.
DISTINGUISHING 1O0DINE FORMS AT HIGH TEMPERATURES AND HUMIDITIES
R. E. Adams, Zell Combs, R. L. Bennett, W. H. Hinds . i e e et st s v reanae s e o
Extensive tests of May packs, which are designed to distinguish lodine forms, have been conducted under

elevated-temperature and high-humidity conditions such as those expected in water reactor accidents.

REACTIONS OF [ODINE VAPOR WITHORGANIC MATERIALS
R. E. Adams, Ruth Slusher, R. L. Bennett, and Zell Combs e s
Laboratory investigations are being made to determine the reactions responsible for the production of

methy!l iodide, which has been observed in containment experiments involving elemental iodine.

BEHAVIOR OF FiSSION PRODUCTS IN GAS-LIQUID SYSTEMS
R. E. Adams, B. A. Soldano, and W. L. Ward oo e et et tret eieeet e her et et b e nn e rea s

A study of the behavior of fission products at the gas-liquid interface has been undertaken.

HIGH-TEMPERATURE BEHAVIOR OF GAS-BORNE FISSION PRODUCTS., TELLURIUM DIOXIDE
M. T Silverman and A. P. MallrmausKas oot eeiis cen serias ee rartaratns s ee e aaie baseursars Frraa s sa saes s nan b ainesataa s aeee s teene s
An experimental investigation of the enhanced volatility of metal oxides in the presence of water vapor has

been initiated.

THE CASCADE IMPACTOR AS A TOOL FOR THE STUDY OF SIZE DISTRIBUTION OF FISSION PRODUCT
AEROSOLS

G. W. Parker and H. Buchholz i ittt o st e e ottt et e e teee e aus et e ataa teee baa e e eeeeae e tiara s e e
Calculations show that operation of the Andersen cascade impactor at pressures in the range 10 to 40 mm
Hg permits extension of its useful range to particles with a diameter less than 0.1 4, and appaeratus has been
devised, and is presently being tested, for this mode of operation,
REACTION OF MOLECULAR IODINE AND OF METHYL 10DIDE WITH SODIUM THIOSULFATE SPRAYS
G. W. Parker, W. J. Martin, G. E. Creek, and N. R. HOTEOD .o e e
We have performed tests in the small (180-liter) stainless steel tank of the Containment Mockup Facility
(CMF) using misting sprays containing 0.1 M sodium thiosulfate to remove molecular icdine and methy!l iodide.
STUDIES OF CSE-TYPE FISSION PRODUCT SIMUL ATION
G. W, Parker, R. A. Lorenz, and N. J. HOrtom o e e e e e
Design, construction, and preliminary testing of equipment for performing CSE-type simulant experiments
in the CMF and CRI have been completed.
RETENTION OF RADIOACTIVE METHYL IODIDE BY IMPREGNATED CHARCOALS
R. E. Adams, R. D. Ackley, J. . Dake, J. M. Gimbel, and F. V. Hensley .
Certain specially impregnated (iodized) charcoals have the capability of effectively trapping radioactive
methy! iodide, by an isotopic exchange mechanism, from flowing air and steam-air over a wide range of cons
ditions includiog 70 to 300°F, 14 to 60 psia, and 0 to 90% relative humidity.

PAPERS PRESENTED AT SCIENTIFIC AND TECHNICAL MEETINGS ...

169

170

175
180
Part |

Molten-Salt Reactors

 
1. Phase Equilibrium and Crystallographic Studies

THE EQUILIBRIUM PHASE DIAGRAM FOR
THE SYSTEM LiF-BeF ,-ZF

R. E. Thoma H. A. Friedman
H. Insley?

Mixtures of 'LiF, BeF , and ZrF, are of especial
interest in this Laboratory because such mixtures
serve as the solvent for 23’5UF4 in the Molten~
Salt Reactor Experiment, Phase behavior of this
ternary system and of its constituent binary sub-
systems has, accordingly, been examined in some
detail. The binary systems LiF—BeF22'3 and Lib-
ZrF44 have been carefully investigated here and
elsewhere and have been described in available
literature. Study of the BeF ,-ZrF, and the LiF-
BeF -ZrF, systems was completed during the past
year.

Most of the data for these systems were obtained
by the technique of thermal gradient quenching
followed by careful examination of the products
by optical microscopy,® though the older technigue
of thermal analysis was of value in some regions.
The regions of liquid-liquid immiscibility in these
systems were defined with the help of high-temper-
ature centrifugation® and careful examination of
the separated products.

The combined data were used in construction of
the phase diagrams shown as Figs. 1.1 and 1.2

 

Yonsul tant.

D, M Roy, R. Roy, and E. F. Osborm, J. Am. Ceram.
Soc. 37, 300 (1954).

AL V. Novoselova, Yu. P. Simanov, and E. 1. Yarem-
bash, J. Phys. Chem. (U.8.5.R.) 26, 1244 (1952).

*a. Insley et al,, Bull, Soc. Franc. Ceram., No. 48,
July ~Sept. 1960.

R. E. Thoma et al., J. Chem. Eng. Data 10(3), 219
(1965).

®Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1965,
ORNI.-3913, p. 3.

Invariant equilibria in these systems were found
to occur at the composition-temperature locations
listed in Table 1.1.

As Fig. 1.1 indicates, the binary system BeF -
ZrF, exhibifs relatively simple phase behavior,
A single eutectic occurs at a relatively low ZrF,

concentration, and the system is free from binary

compounds. Two immiscible liquids occur in
mixtures containing 14 to 25 mole % ZF ; the
upper consolute temperature is near 740°C,

The LiF-BeF -ZiF, system (Fig. 1.2) is, so far
as we are aware, the only temary fluoride system
yet shown 1o include immiscible liquids, As the
temperature is increased above the liquidus, the
composition interval showing two liquid phases
steadily diminishes; it disappears at an upper
consolute point at 25 mole % LiF and 55 mole %
BeF, at 955°C.

This occurrence of two liquid phases in the
system at high concentrations of BeF, does not
prejudice use of materials in the composition region
near 65 mole % LiF and 30 mole % Bel, as fuel
solvents for molten-salt reactors.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

GO0 rovemsemreeee *~‘

800 || LiguiD e

2l 1 ‘

0 LIQUIDS -\
= 700 1\
Ll
@ ‘
2 600 [ L
=L ZrF, + LIQUID
ui 548 \ |
& so0 ] LT T T T
= BeF, + IrF,

400 e - o1t |

F00 Foemeeeees ‘ Lo

 

 

 

 

 

 

 

 

 

B(—EFz 10 20 30 40 50 G0 70 80
Zr, (mote %)

Fig. 1.1. The System Ber-ZrF4.
PRIMARY PHASE AREAS:
@) uir
(B} 6LIF-Bef, ZrF,
(© Li,Beq
(D LiyzrF,
€ 3LiFZrE,
(F) 3LiF-azcF,
@ BeF,
® zrs,

4

Zrk,

ORML-DWG 65— 7324 -

TEMPERATURES IN °C
COMPOSITION IN mole %

 
 
 
 
 
 
 

 

 

- LiBefy

Fig. 1.2. The System LiF-Ber-ZrF4.

Table 1.1. Invariant Equilibrium Points in the Systems BeF ,-Zrf ; and LiF-BeF,-ZrF

 

Composition (mole %)

 

 

Temperature Type of
LiF BeF ZiF, (°0) Equilibrium Solid Phases Present
92.5 7.5 525 Eutectic BeFZ, ZrF4
36 14 645 ZrF4
74 26 645 ZrF4
75 5 20 480 Peritectic LiF, 3LiF - ZrF , 2LiF - ZrF,
73 13 14 470 Peritectic LiF, 61L.iF - BEF2 . ZrF4,
2LaF « ZtF
4
67 29.5 3.5 445 Peritectic LiF, B6LiF » Ber . ZrF4,
21LiF - BeF
4
64.5 30.5 5 428 Peritectic 6L4iF - Bel, - ZrF4, 2LiF - ZrF4,
2LiF » Be¥ 2
48 50 2 355 Eutectic 2LiE - ZrF4, 2LAiF - BeFZ, BEF2
47.5 10 42.5 466 Peritectic 2LiF - Zer, 3LiF » 4ZrF4, ZrF4
44 18 38 460 Eutertic 2ILiF - ZrF4, BeF2, Zrk,
27 46 27 532 Ber, ZrF‘4
2 a8 10 532 BeF ZrF
PRELIMINARY STUDY OF THE SYSTEM
LiF-ThF -PaF

C. J. Barton H. H. Stone
G. D. Brunton

I the protactinium recovery studies described
in Chap. 3 of this report it has been generally
assumed that protactinium is present in the LiF-
ThE, (73 mole % LiF) melts as PaF,. This as-
sumption has seemed plausible since the melis
have, in every case, received a treatment with H,
at temperatures near 650°C, and since H, is known
to reduce pure PaF_ to PaF, at much lower temper-
atures. (The Pa'" state seems to be the lowest
known in fluoride sysiems.) We have, however,
conducted a few preliminary experiments to test
this assumption and to see if the phase behavior
of Pal’, is similar to that of Th¥, in mixtures
with LiF,

About 100 mg of ?*'PaF, was prepared by
evaporating a measured portion of purified stock
solution (9 M in HF) to dryness in platinum and
heating the residue to 600°C in flowing HF-H, mix-
tute. Conversion to PaF, was confirmed by weight
and by the brown color of the material. A portion of
this material was mixed with LiF-ThF4 (73 mole
% LiF) to yield a mix with 68 mole % LiF and 32
mole % (Th,Pa)F,. Another portion was mixed
with LiF and the LiF-ThF  mixture to yield a
mix with 73 mole % LiF and 27 mole % (Th,Pa)F4.
Both mixtures were admixed with ammonium bi-
fluoride (whose decomposition products on heat-
ing help to minimize possible hydrolysis), heated
to 650°C, and cooled slowly.

Examination of the slowly cooled melts showed
that segregation of Pal -rich phases from the
bulk of the LiF-ThF, material occurred in both
cases. Material from the mixture with 68 mole %
LiF is believed to be a solid solution of LiPaF
in LiThF_. One of the phases from the sample
with 73 mole % LiF is believed, because of its
similarity to the analogous uranium compound, to
be Li,PaF,. The PaF, does not appear iso-
morphous with ThF,; the LiF-PaF_  system may,
in fact, be more like the LiF-UF, than the LiF-
Th¥, system. It is obvious that study of the binary
LiF-PaF, system is needed before attempting
further deductions conceming phase relations in
the ternary system LiF-ThF -PaF,.

A portion of the LiF-ThF -PaF, mixture with
73 mole % LiF was transferred to a small thorium

crucible and heated to 650°C in a helium atmos-
phere. Examination of the material with the polar-
izing mictoscope revealed some LiSThF,V but a
large part of the mixture was in the form of opaque
angular fragments, which are probably protactinium
metal. X-ray examination will be required to con-
firm this conclusion.

APPARATUS FOR AUTOMATIC DIFFERENTIAL
THERMAL ANALYSIS

L.. O. Gilpatrick
5. Cantor

R. E. Thoma

Fluoborate mixtures containing high concentra-
tions of NaBF, appear useful as secondary cool-
ants in molten-salt breeder reactors.’
systems show significant BF, pressures at tem-
peratures above the liquidus, our standard tech-
niques of thermal analysis and thermal gradient
quenching are applicable only with difficulty to
these materials. We have, accordingly, developed
(with help from the ORNL
Controls

Since such

Instrumentation and
Division) a sensitive automatic differ-
ential thermal analysis apparatus for study of the
fluoborates.

This equipment, similar to that used by Holm,®
consists of a series of components assembled as
illustrated in Fig. 1.3, Specimens are contained
in sealed thin-walled nickel tubing, having outer
dimensions ¥ x 2.5 in. Temperatures are monitored
by armored 40-mil thermocouples which are posi-
tioned from below in reentrant chambers. Specimen
containers and thermocouple assemblies were de-
signed for minimal heat capacity, The specimen
thermocouple supplies a signal in opposition to
that from a matched cell containing fired Al,O,
as a comparison standard. Both cells are mounted
in a massive nickel block and out of direct thermal
contact with the block by means of ceramic sup-
ports.  An independent thermocouple embedded
in the block provides a signal which programs
the temperature and heating rate of the system as
a whole. Differential temperatures are recorded
as a fuanction of time on a model 7002AMR Moseley
x~y recorder after amplification of from 500 to

 

7R. E. Thoma and G. M. Hebert, ““Coolant Salt for a
Melten Salt Breeder Reactor,”’ Patent Application
CNID-2100, Nov. 16, 1966.

8y. L. Holm, Acta Chem. Scand. 19, 261 (1965).
FURNACE HEATER

 

 

ORNL-DWG 67- 768

L ?HOV

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(e
...... ) —\_ "LaBacC"
C—————Too— L SOLID-STATE
ey CONTROLLER
NICKEL BLOCK
O-10 mv
H fl SIGNAL
TEMPERATURE
o | CONTROL T.C. ) ] CONTROLLER |
T AND
— S PROGRAMME R
CsalT -] [ A0 T
SAMPLE STD
o MOSELEY [
20 ICE BATH v o
|1 RECORDER
AN - {x) -
e R
TEMPERATURE
e L AND N
AT
L || sTEPPING
\\—777 B ot .
vvvvv MICROVOLT =y 5128

 

 

 

 

 

 

 

ICE BATH

 

Fig. 1.3.

20,0600 as desired, via a Leeds and Northrup
microvolt amplifier, model No. 9835B. The record-
ing svstem is calibrated daily by a potentiometric
circuit and a standard cell used to produce a known
signal. The apparatus temperature was
calibrated by measuring the melting points of lead
and bismuth metal standards from the National
Bureau of Standards.

input

The temperature programming
unit was designed to contwl linear temperature
changes over variable time periods from 30 to
300 min.
of a 0« to 10-mv signal, is fed to a solid-state
This
device automatically programs repeated heating
and cooling of specimens within preset limits and
thereby enables automatic collection of phase
transition data for periods of 50 hr or more,

The equipment is being applied to definition of
liguid-solid and solid-solid transitions in the NaF-
BF, system. Initial results show that the melting
point of NaBF, is 383.1°C; this material, which
contains less than 200 ppm of oxide ion, appears

Output from the programmer, consisting

power unit which controls the heat input.

AMPLIFIER

 

 

 

 

L

 

Block Diegram for Automatic DTA Apparatus.

to melt some 15°C higher than the (probably less
pure) material described by Selivanov and Stander.®

SOLID-PHASE EQUILIBRIA IN THE SYSTEM
Ssz-fimF3-UF3

R. E. Thoma H. A. Friedman

The fact that reduction potentials for the reac-
tion Ln*" > Ln?" are lowest for the lanthanides
and samarium {(approximately 0.4 v)
implies that the interactions of Smi,, SmF,, and
UF, possibly are significant in the development
of molten-salt fuel reprocessing methods which
would employ UF, as an ion exchanger for selec-

europium

tive removal of rare-earth fission product fluorides.

Pioducts of the reactions of Sm° or U° with
SmF,, SmF,, UF,, and mixtures of these fluorides,
which take place at 1300 to 1400°C, were analyzed.

9V. G. Selivanov and V. V. Stander, Russ. J. Inorg.
Chem. 4, 934 (1959),
The crystal phases were characterized with micro-
scopic, x-ray diffraction, and electron resonance
methods and were chemically analyzed, Samarium
trifluoride was found to be stoichiometrically
reducible with Sm° to SmF,, a cubic blue phase
with refractive index 1.534 and a, - 5.866 A.
Partial reduction of SmF, with Sm” produces a
birefringent red intermediate phase of unknown
structure, inferred tentatively to be the compound
SmF,.SmF .. The limited data which
rently available indicate that within the system
Sm¥ ,-SmF -UF solubility of
the various components and intermediate phases
prevails except between SmF , and the intermediate
phase SmE, - SmF .

If this conclusion is borne out in future experi-
ments, the reduction potentials of salts which
pass through UF, icn exchange beds will possibly
be significant parameters in process control.

are Ccur~

extensive mutual

PHASE RELATIONS IN THE SYSTEM KF-CeF

D. Hsu'®
H. Insley!

C. J. Barton
G, . Brunton

Previous studies'! !'? have shown that phase
relationships in the LiF-CeF, and NaF-Cel,
systems are very similar to those for their LiF-
PuF, and NaF-PuF, counterparts. A brief ex-
amination of the KF-CeF , system has, accordingly,
been conducted to show probable behavior for
KF-PuF .. Woik will begin on KF-PuF, as soon
as differential thermal analysis eguipment becomes
available with
small amounts of materials.

to provide adequate sensitivity

We obtained thermal analysis data from cooling
curves obtained with a bare platinum—platinum-
rhodinm thermocouple immersed in the melt con-
tained in a small (5 ml) platinum crucible. The
calculated weights of KF and CeF, (to give 3 to
4 g total) were placed in a crucible along with
several grams of ammonium bifluoride, and the
mixture was heated in a flowing stream of helium,
slowly at first until the ammonium bifluoride de-
composed, and then more rapidly until a tempera-

 

IOSummer employee from Universgity of California,

Berkeley.

e, J. Barton and R. A. Strehlow, J. Inorg. Nucl,
Chem. 18, 143—-47 {1961).

IZC. J- Barton, J. D. Redman, and R. A. Strehlow,
F. Inorg. Nucl, Chem. 20, 45 (1961).

ture well above the estimated liguidus temperature
was reached.

Gradient quenches were performed with portions
and the resulting
samples were examined by use of a polarizing
microscope.  The principal findings of the in-
complete investigation are as follows: There is
one eutectic composition at about 19 mole % CeF ,
melting at 625 + 5°C, and two incongruently melt-
ing compounds, 3KF-CeF, and KF.CeF,. The
former melts at 675 + 5°C to give KF.CeF, and
liquid and it also decomposes on cooling at
595°C £ 5° giving KF and KF -CeF,. The latter
compound melts at 755 & 5°C giving liquid and a
Thermal
analysis did not indicate the upper and lower sta-
bility limits of 3KF .CeF,, nor did it provide reli-
able liquidus values for mixtures containing more
than 20 mole % CeF ..

It was interesting to compare the data obtained

of the slowly cooled melts,

cubic phase of unknown composition.

for this system with the proposed diagram for the
system KF-NdF, recently reported by Schmutz, '
He reported a eutectic containing about 20 mole %
NdF,, melting at 625°C 1 10° and three com-
pounds, 3KF.NdF,, KF.NdF,, and KF:ZNdP‘3,
melting incongruently at 695, 750, and 1115°C
respectively. His phase diagram was based on
differential thermal analysis data and examination
of the slowly cooled melts by means of x-ray dif.
fraction. On the basis of cur studies of quenched
melts it” seems probable that the system is more
complex than shown by Schmutz’ diagram. We
plan to perform further thermal analysis and quench-
ing studies to better define the composition of the
high-temperature cubic phase mentioned above,
It is quite possible that it is more closely re-
lated to the 5Nal’.9YF, compound!? than to the
simple 1 to 2 compound postulated by Schmutz,

THE CRYSTAL STRUCTURE OF Li UF,

G. D. Brunton

The compound Li,UF_ crystallizes in space
group Pnma with a, = 9.960, b, = 9.883, and ¢, =
5.986. The x-ray density is 4.71 g/cc, and Z = 4.

 

“?H. Schmutz (thesis), Investigations in the Alkali
Fluoride-—Lanthanide or Actinide Fluoride Sysfems,
Kernreaktor Bau- und Betriebs--Gesellschaft w.b.H.,
Karlsruhe, Germany, KFE-431 (July 1966).

YR, E. Thoma et al., Inors. Chem, 2, 1005 (1963).
ORNL-DWG 85-14450

  
  

3 F & g . b , il

                  
 

   

YW -

e

   

F)

            

       

 

          

£F TG & o W, -‘0 I ‘
g BwET W ) A g A B

vl\w«"qi : -

      

e

g Ay

       

     

X
e
,_-

- e
/AL N or i B g 4.\»y/ ]

 

   
 

) “J% .1.‘ S

     
  

   

o

      
 

&
b

Li,UF

Fig. 1.4, Stereoscopic Drawings of the Crystal Structuse of
Table 1.2, Atomic Parameters for Li4UF3

 

 

 

i 3 3 1. 3 .. 4 (. _ 4 ; 4 R 4
Atom  x L1007 © ytwto ziwto B Fwto B, *wte fot10te B L10%o
U 0.140000.1)  0.2500 - 0.3634(0.2)  0.0007(1) 0.0011 1) 0.0085(3 0,000 2)
(1) 0.027(2) 0.107(2) 0.128(3) 0.0056(7)
F(2) 0.023(2) 0.121(2) 0.608(2) 0.0045(6) Isotropic temperature factors:
F(3) 2.241(2) 0.031(2) 0.375(3) 0.0058(7) 2
F(4) 0.305¢3) 0,250 0.117(4) 0.0059(10) % By s By

2
F(5)  0.292(2) 0.250 0.633(4) 0.0045(9) G, - c* B
33 2 1%

Li(1) 0.376(9) 0,055(9) 0.099(15) 0.0106(39) *
Li(2)  0.395(13) 0.060(16)  6.649(22) 0.0174(73) _/

 

Twenty-four positional parameters, four anisotropic
uranium temperature factors, and seven isotropic
temperature factors (Table 1.2) were determined
from 634 independent reflections measured by the
20-scan  technique with a scintillometer.  The
parameters were refined by least squares to an R
tactor of 0,082. Absorption corrections were made
for Cu Ka radiation on an oblate spheroid with a
short 36-p axis, (001], and a 64-p diameter for the
circular section. The U*” ion has eight I near-
est neighbors with bond distances of 2.21 to 2.39
A, Fig. 1.4, The next three nearest neighbors
are two Li* and another F™ at 3.27 and 3.39 A
respectively. The nine F7 ions are at the corners
of a l4-faced polyhedron which has the form of a
triangular prism with pyramids on each of the three
prism faces, and the two Lit ions are at the centers
of irregular F~ octshedra which share faces with
the uranium polyhedron. The Li'~F~ distances
are 1.82 to 2.28 A.

THE CRYSTAL STRUCTURES OF NaF-LuF,
SOLID SOLUTIONS

D. R. Sears G. D. Brunton

A remarkable feature of the sodium fluoride—rare-
earth trifluoride binary systems is the occurrence
of a cubic phase whose lanthanide-rich composi~
tion limit corresponds to a formula SNaF - 9LnF3..IS
The 5:9 stoichiometry is bizatre, but its independ-

 

3. E. Thoma, H. Insley, and G. M. Hebert, Reactor
Chem. I}v., Ann. Progr. Repf. Dec. 31, 1965, ORNL-
3913, pp. 6 ff.

of cationic
radius) 1s even more surprising, Seeking a crystal-
chemical explanation of this behavior, we have
begun to determine crystal structures of cubic
NaF-LnF, materials of selected compositions.
Complete three-dimensional x-ray intensity data
have been collected for two cubic sodium lutetium
fluorides, whose compositions are 51.2 and 56.6%
LUF.s' Reflections were measured with a spectro-
goniometer and single-crystal orienter, using the
20-scan technique. Using full-matrix least-squares
and difference-synthesis methods,

ence of choice of lanthanide (i.e.,

a variety of
cation vacancy and anion interstitial models were
tested and refined. Basically, all models were
derived from the classical fluorite (Cal,) structure.

Best goodness of fit {(as judged by R factors
and difference syntheses) was obtained for each
crystal using a model consisting of 2 mixed cation
site at the origin, a fractionally occupied fluorine
site at (0.275, 0.275, 0.275), and a second {raction-
ally occupied fluorine site at (1/?, x, x). For
51.2% LuF ,, x = 0.114, and for 56.6% LuF,, x -
0.061. The corresponding R factors are 4.6 and
3.8% respectively.

Fluorine-fluorine distances are impossibly short
and would seem to be unacceptable even if x = 0.
The model is unrealistic also in the constraints
necessarily imposed upon thermal parameters in
order to assure convergence: the isotropic thermal
motion parameters of the two types of tluorine
were constrained to be equal.

We conclude that the anion interstitial model is
tenable only if the static fluorine displacements
from (1/4, Y ,3/4) and (1/2, 0, O) are fictitious and

anharmonic thermal occurs along

A
that motion
10

ORNL-DWG ©66-13040

 

F
i
m‘““*-—v.: o
F[ll: ] s ;
hy
7~ -3
ST
fr%'
FiS

o C
Fa o | Fa

 

Fiy

 

A
-flrru} PR A
Fag Fin) ><
!

 

 

 

 

 

 

 

 

 

Fi

Fig. 1.5. Stercoscopic Drawings of the Structure of CsBeF3.
Table 1.3, Atomic

11

Paroemeters for CsBeF3

 

 

 

Atom x y z B11 B B B B B
22 33 12 13 23
Cs 0.265 0.250 0, 107 0.0234 0.0143 0.0058 0.0 0.0007 0.0
Be 0.705 0.250 0.679 0.0216 0.0319 D.0063 0.0 0.0003 0.0
F{1} 0.244 0.042 0.783 0.0245 0.c211 .0106 0.0031 0.0006 0.0069
F{2) 0.884 0.250 0.679 0.0169 0.0282 0,0033 0.0 0.0007 0,0
tetrahedral directions from these sites. This kind

of motion has been postulated in the case of UO,,
’l’hOz, and Can,”’ and the conclusion is at least
consistent with the appearance of difference
syntheses of the two NaF-LulF, crystals hitherto
examined.,

We will attempt to collect intensity data from
NaF-LuF, crystals near the composition limits
(39 and 64.29% LuF,) in order to reduce the cal-
culational ambiguities and to attack directly the
problems imposed by the puzzling 5: 9 ratio.

THE CRYSTAL STRUCTURE OF »-CsBeF,

H. Steinfink!? G. D. Brunton

The gamma (low-temperature) form of the com-
pound CsBeF, crystallizes in space group Pnma
with a) = 4.828 A, b, =6.004 A, and ¢, - 12.794 A,
The x~ray density is 3,56 g/cc, and Z = 4. Nine
positional parameters and 16 anisotropic tempera-
ture factors (Table 1.3} were determined from re-
flections measured on a Norelco PAILRED auto-
matic crystal data collector. The parameters were
refined by least squares to an R of 0.11.

Each Cs' ion is surrounded by eight F™ nearest
neighbors with bond distances of 3.02 to 3.92 A
(Fig. 1.5). The Be®" jons have four nearest
neighbor F7 ions at the corners of a tetrahedron.
The Be®™-F™ distances are 1.47 to 1.62 A, The
structure of this compound is similar to that of
the high-temperature form of BaGeO,.'® The re-
puision of the doubly charged Re®' jons increases
the Be-F distances where the F7 ions are shared
between two tetrahedra, This accounts for the un-
usually long (1.62 A) Be? *-F™ distances.

 

185, 1. M. Willis, Acta Cryst. 18, 75f (1965).
17Consultant from the University of Texas.
18v7on Waltrud Hilmer, Acta Cryst. 15, 1101 (1962).

THE CRYSTAL STRUCTURE OF 3,-KLaF

D. R. Sears

Hexagonal (5 -KLaF, solidified as mecohedrally
twinned crystals of space group Pb, almost iso-
structural with NaNdF , and not with 8 -K UF,
as previously reported,'? Seven coordinates,
fourteen anisotropic thermal parameters, and one
occupancy fraction (vide infra) were determined
from 258 independent reflections measured with a
spectrogoniometer and single-crystal orienter,
using the 20-scan technique. The structural
parameters were refined to an R factor of 5.8%
using full-matrix least-squares methods and appear
in Table 1.4, The unit cell and some adjacent
atoms are illustrated in Fig. 1.6,

This cell, with a, = 6.530 £0.002 A, ¢, = 3.800 *
0.002 A, contains ‘?/2 formula weights of KLaF,
and is disordered with respect to fluorine occu-
pancy of a pair of unrelated sites on either ‘side
of the twinning planes and potassium occupancy
of a pair of half-occupied, crystallographically
equivalent sites above and below the horizontal
mirror planes. Furthermore, there exists a calion
site filled randomly by both lanthanum and potas-
stum  ions. Ordered lanthanum and disordered
potassium, as well as the mixed cation sites, are
all nine-coordinated by fluorines, each of which
is shared with five additional (but not identical)
coordination polyhedra. These polyhedra, compris-
ing the “tripyramidal’’ coordination, are
structed by erecting

con-
right pyramids upon each
rectangular face of a trigonal prism. In BI»KLaFV
individual polyhedra lack a 6 figure axis because
both fluorine and potassium are disordered.

The La-I¥ distances range from 2.36 to 2.51 A,
and the K-F distances from 2.46 to 3.05 A,

 

1QW. H. Zachariasen, Acta Cryst. 1, 265 (1948).
Table. 1.4, Sites, Symmetry, and Least-Squares Adjusted Parameters” of ,81-KL0F4

 

 

Atom site  Jccupancy x 100 y 10 zt10% 10°8, e 10%p,, r10’c 1038, t10%0 18, t10'c
Fraction

La (a)e 1 0 0 0 4.4(0.2) b 6.3(0.4) b

Y (K+La) Wb 1 % % % 3.4(0,3) b 8(1) b

K a3 Y % 7 0.455(4) 13(1) b 31(9) b

F(1) 3m 1 0.255(2) 0.247(2) 7 %(2) 6{2) 22(4) 4(1)

F(2) AHm  0.57 T0.62  0.380(9) 0.040(7) 0 24(12) 53(23) 52(20) 20(16)
F(3) 3m ¢ 0.358(18) — 0.035(16) O d d d d

Al

 

f8roordinates were calculaied in the lasi cycle of a *‘coordinates-only’’ refinement, thermal parameters and fluorine occupancy fraction in an earlier
cycie when only they were adjusted. All standard errors, however, were calculated in a one-cycle least-squares adjustment in which all disposable
narameters were allowed to vary. For the atom coordinates, these standard errors average 46% higher than the errors calculated in a coordinates-only
calculation.

®These parameters are fixed by symmetry relations: ‘822 = Bli' )812 = 3’2 ,8
9, 679 (1936).

“Constrained to equal 1 minus the occupancy fraction of F(2).

In addition, for all atoms 51 s ,823 =0, Cf. H. A, Levy, Acta Cryst.

11°

9These parameters were artificially censtrained to egual the corresponding parameters of atom F(2).
13

ORNL.- DWG 66-6926

 

Fig. 1.6. Schematic Perspective of the B]'KLQF;L Unit Cell and lIts Environs.

CENTRAL CATION DISPLACEMENTS IN THE
“TRIPYRAMIDAL™ COORDINATION

D. R. Sears

All cations in I\IaI\Ile“‘20 and KLaF, (preceding
section) are located in “‘tripyramidal’’ coordination
environments, Both sodinm and potassium exhibit
targe displacements from their ideal positions,
although it has been assumed®’ generally that the
cation shounld lie on the polyhedron medial plane.
Seeking an explanation of these data, we have cal-
culated potential energy sums 2?2 of the ““exponential-
six’’ form:

1 or.. r. \?
Z e |Bexp( o= ) u(_.?_‘)
: H r, T

Y -

and of the Lennard-Jones form:

; 6
2_1__ 6-59_;,,.3 i)
J.S~m6 r £

 

20], H. Burns, Inorg. Chem. 4, 881 ff. (1965).

?'ISee, for example, D. L. Kepert, J. Chem. Soc

Japan 7, 348 ff. (1952).
22$€:e, for example, T. Kihara and S. Koba, J. Phys.
Soc. Japan 7, 348 ff. (1932).

as well as the Coulombic sum 2 (l/rl.j) for various

anion configurations in the trifyyramidal coordina~
tion. We used values of o and s from 7 through 12
and incremented the central cation position along
the polyhedron figure axis from the medial plane
to the basal plane. Interactions beyond the poly-
hedron were ignored, the anion framework was
assumed to be rigid, and r, was taken to be the
sum of Pauling radii.

For the examples studied so far (Na, Nd, and
mixed Na + Nd in NaNdF4, and K, La, and mixed
K + La in f,-KLaF ) the Coulombic term varied
less than 0.1 to 1% for a 1-A displacement. By
contrast, relatively larger variations in the ex-
ponential-six and Lennard-Jones functions were
found. Results were nearly independent of choice
of function and choice of s or ¢, insofar as the
qualitative conclusions are concerned, and ac-
cordingly the results are presented here only for
o == 12 in the exponential-six case. In Table 1.5,
Az . is the experimental cation displacement,
Az .. is the location of the minimum in the poten-
tial function, Az(1%) is the displacement from the
minimum corresponding to a 1% increase in poten-

tial, and \/L;g; is the rms thermal displacement
derived from the experimental thermal parameters
f3,, corresponding to vibration along the poly-
hedron figure axis (i.e., the hexagonal z axis).
14

 

 

Table 1.5. Results of Potential Calculations in NaNdF ; and fi]-KLmF‘1
2
Atom "zobs L/“\zcalc V uobs Az( 1%)calc
(4) (A) (A) (A)
Na in NaNdF , 0.58 0.71 0.19 0.15
K in KLan,’ 0.17 0 0.15 0. 17
L.a in KLaF, 0 0 0.07 0.06
Nd in NaNdF4 0 0 0.05 0.12

 

We conclude that the cation displacements can
be rationalized in terms of the relative insensitivity
of the Coulombic interactions to cation location
in these structures and that displaced potential
minima, or very shallow minima, in the remaining
potential contributions will suggest which anion
configurations will favor or permit cation dis-
placements,

Calculations are being extended to include other
framework structures containing tripyramidal co-
ordination polyhedra.

PREPARATION OF FLUORIDE SINGLE
CRYSTALS FCR RESEARCH PURPOSES

R. E. Thoma R. G. Ross
H. A. Friedman

Our efforts to develop techniques for the produc-
tion of large (350 g), pure single crystals of LiF
with selected isotopic ratios*® have succeeded in
the production by Stockbarger-Bridgman methods
of crystals of as high chemical purity as can be
obtained with Czochralski techniques. Concentra-
tion of heavy-metal impurities in crystals produced
this year was found to be less than 2 ppb from
activation analysis and thermal conductivity data.
Thermal conductivity of these crystals, as meas-
ured by investigators at Cornell University, was
found to be virtually identical through the tempera-
ture range 2 to 50°K with the best "LiF crystals
grown from our starting materials by the Czochral ski
method.

Since purity control has been developed to the
extent that further improvement is limited by the

capabilities of the analytical methods (spectro-
chemical, infrared absorption, and activation
analysis), most recent efforts have sought to
reduce the population of crystal dislocations. By
programming the reduction of annealing tempera-
tures at a slow steady rate after crystal growth,
in one experiment at 2.5°C/hr in the temperature
range from 840°C to room temperature, the popula-
tion of crystal dislocations in LiF crystals has
been reduced significantly. Preliminary indica-
tion of the effectiveness of such programmed an-
nealing was obtained by neutron diffraction ex-
amination of ORNL-14,%% a 550-g crystal of lithium
fluoride of 99.993 at, % ’Li, which showed a major
volume of the crystal to be essentially free of
disorder.

Part of the effort to synthesize fluoride single
crystals was devoted to the preparation of crystals
for use in x-ray and neutron diffraction experiments.
For such purposes, neither size nor purity stand-
ards are as demanding as those imposed for LiF
preparation. Single crystals (approximately 1 cm
maximum dimension) of each of the three com-
pounds Li,BeF,, Na Zr F, , and "Li,NaTh,F,
were grown by the Stockbarger-Bridgman method.
For other structural investigations?® which employ
even smaller crystals, suitable single crystals
of [-CsBeF, and Li UF, were grown by high-
temperature anneal-quench experiments.

 

B Reactor Chem. Div. Ann. Progr. Rept. Dec. 31,
1965, ORNL-3913, p. 126,

24Performed by H. G. Smith, ORNL Solid State Divi-
sion.

6. D. Brunton, preceding sections, this chapter.
2. Chemical Studies of Molten Salts

A POLYMER MODEL FOR LiF-BeF, MIXTURES

C. F. Baes, Jr.

The salts LiF and BeF, are quite dissimilar in
character; LiF is a normal icnic salt, while BeF
quite evidently is more covalent since it foms
a very viscous liquid' which is obviously poly-
meric.  Solid BeF, has a structure analogous to
510 ,; Be?t jons are surrounded tetrahedrally by
four flucride iens, and each F7 ion is bonded to
two Be?' ions, that is, BeFf* tetrahedra share
comers to form a three-dimensional network., The
liguid would be expected to have a similar poly-
meric structure. As LiF is added to molten BeF ,
the viscosity drops sharply,! presumably because
bridging fluorids linkages ave broken,

! I

~Be-F-Be 1 F~

oY

——

|
2[wBle—F] , (D

and the degree of polymerization decreases. The
stoichiometric end point for this process occurs
at the composition ZLil*-BelF,. Thereafter, the
principal beryllium species in the melt presumably
is BeF 27,

This plausible qualitative picture of LiF-BeF
melts has been treated more quantitatively in the
following fashion. It was assumed that all of the
many Be-F complexes formed would retain the
same three elements of structure — I-BeF42“ tetra-
hedra, bridging F7 ions, and teminal F7 ions.
For a given complex anion Be F (P-28)= it can
then be shown that there are 4a — b bridging ¥
ions and 2b ~ 4a teminal F7 ions. Hence, for
the formation reaction,

—

aBeF *" = Be F (P77, (4a - BF™, (D)

 

s, Cantor and W. T. Ward, Reactor Chem. Div. Ann.
Progr. Rept. Dec. 31, 1965, ORNL-3913, p. 27.

15

the number of bridging ¥~ ions increases by 4a — b
while twice that number of terminal F~ ions dis-
appear. It seems reasonable, therefore, to write
as an approximation for the free energy of reaction
the equation

AG, =(4a - B4, (3)
in which {at a given temperature) A is a constant
representing the free energy increase associated
with the formations of 1 mofe of bridging ¥ bonds
from 2 moles of terminal F~ bonds. The develop-
ment of the model from this approximation is sum-
marized in Table 2.1, and equation numbers given
below are from that table. In the equilibrium con-
stant expression (Eg. 4), the anion activities are
represented by adaptations (Egqs. 5-7) of Flory’s
equation? for polymer solutions, in which the vol-
ume fraction (9’5a,b) of Beanj(bwza)b is calculated
in terms of the number of ¥ ions the species
containg (Eq. 8). These modified Flory equations
also contain a heat of mixing parameter (8) and
an average value of b (b, Eq. 7) in the mixture.
From Eqs. 4-9, the volume fraction of a given
polymeric ion (¢, ,) may be cslculated (Eq. 10)
in terms of a, b, the volume fraction of BeF 2~
(961'4), the volume fraction of F™ ((760’_1) and the
two adjustable parameters A and B. Introducing
this expression for an,b info two material balance
equations (11 and 12) one obtaing iwo equations
which may be solved for the unknowns ¢1,4
téo,l' These, in turn, were used to calculate the
activity of BeF by a relationship (Eq. 13) de-
rived from the proportionality

and

{14)

and Egs. 6 and 7.

 

QP. J. Flory, Principles of Polymer Chemistry, p. 513,

Cornell Univ. Press, Ithaca, N.Y., 1953.
16

Table 2.1. Derivation of BeF, Activity in LiF from Polymer Model

 

The formation constant for Bean(b'fiza)“ is given by

 

K, - OBeaFpte) 7 , (4)
a, b (31361?4)6
in which
1 2
In(age, r,) = In(h, ;)= (b~ 1)+ b1 - (1 -y )+ BBSL |, (5)
/ ! 2
ln(aF)zln(g)o!l)+ lw_g_ (1“—¢>0'1)+B(1—q60'1) (6)
1 2
1“(aBeF4) = ln(¢1’4) ~3+411 - ? 1 - ¢0'1) + 43650’1 , (7
C1?5.51',1) =b Na,b//zég(b Na,b) ! (8)
B-EEGBN, V/ZEW, ) - 1LEES. /h, 9)
ab ! ab ’ ab !
gives the volume fraction of Bean(b“ 2a)-
4a—>b
e )6 823§60,1
d)a,b B ( fus Lfi (1tsta/RT) ' (10)
e 0 1 €
This is introduced into the material balance expressions
1-386, ;. (1)
ab !
XL b0 s)
XBeFf_"’—bfg?m—fifi , (12)
1 — —ch
ab (b ja’b>
to obtain ¢, , and ¢ |, which are introduced into
, 2
In agey, =Inle, /o5 120, -= A -, )+2Bdy (246, ) (13)

to give the activity of BeF ,.

 
The calculation of a by this model thus in-
volves three double summzzitions,
( ]) ) -

LL @, LL (b bo, ) ,and LY

ab
These were extended to include all values of a
and, for each value of a, values of b beginning at
b - 2a + 2, until each series converged. Using
the CDC 1604 computer, the adjustable parameters
A and B were varied by a least-squares procedure
until the calculated values of apep, were the most
consistent with values?

ments of the equflibnum

4 calculated from measure-

HgO(g) + BeF (dy = BeO(S) + ZHF(g) , (15)
PZ
Kool HE (16)
F : .
H20 BeF 2
While the present calculations of agep, were

lengthy, the model at its present stage is a simple
one which fits the data well enough to suggest
that it offers a reasonable representation of the

structure of LiF-BeF, mixtures. In this repre-

. . . .+
sentation, the mixtures contain Li' as the only
cation; the anions are 7, Be’F,;z‘", and numerous
(b--2aj~

which include
chains of varying length and cross-linked struc-

polymeric anions Be F,

tures containing rings in various numbers, and
. . /‘F‘\ -~ .
in sizes down to Be\ ~Be . The distri-

bution of Be?’ and F- dmung all these possible

structures depends primarily upon the F/Be?"

ratio in a given mixture and upon the relative
|

!
stability of wBe—FwBiem plus F~

I
two — Be — F groups.
i

compared to

PHASE EQUILIBRIUM STUDIES
IN THE U0,-Z<0, SYSTEM

K. A. Romberger C. ¥. Baes, Jr.
H. H. Stone

The previously described study of the UO, —?rO
system in which a molten-salt flux was u‘;ed to
achieve equilibrium between the oxide phases has
been completed during the past year. The com-
position assigned to the tetragonal solid solutions
at the eutectoid temperature (1110°C) has been

17

increased from the value 1 mole % uo, reported
previously to 2.7 mole % on the basis of recent
measurements which indicate that the previous
result had not corresponded to equilibrium. In
addition, analyses have been obtained on a mix-
ture of cubic UO, and monoclinic ZrO which had
been equlhbrated with a molten salt at ~600°C
for 60 days® glvmg ~0.3 mole % ZrO, in UO,
and ~0.07 mole % UG, in ZrO . These results
indicate a lower rate of eysolutmn of the two
phases with decreasing temperature than previously
estimated.

The revised phase boundaries in Fig. 2.1 reflect
these changes. In addition, the boundares below
1150°C reflect the distribution behavior expected
for dilute solid solutions in the three two-phase
regions C =& M, M & T, and C = T. In partic-
ular, it was assumed that the distribution coeffi-
cient (P, a ratio of mole fractions) for one com-
ponent between two solid solutions at equilibrium
would obey the relationship

log D =a+ b/T. (17)

In the case of the M = T equilibrium, the esti-
mated dependence of Xz.0 (T)/Xz.6,(M) on tem-
perature indicates the enthalpy of the M — T transi-
tion in pure ZrO, (1170°C) to be 1.7 1 keal/mole.
This compares favorably with = measurpd value
of ~1.4 keal.” Just above 1150°C (the upper
limit of the present measurements) it is clear that
values of D wmust deviate markedly trom those
predicted by Eq. (17) (corresponding to the sharply
bending dashed curves in Fig. 2.1) if the C = T
phase boundaries are to be consistent with the
publisted higher temperature data,  Such pro-
nounced deviations from ideal behavior in solid
solutions that are still quite dilute seem unusual.
Consequently, additional measurements between
1150 and 1500°C by the methods used here would
be of considerable interest.

 

3A. L. Mathews and C. F. Baes, Jr., ORNL-TM-1129
(May 1965).
‘AL I.. Mathews, Ph.D. thesis, “*Oxtde Chemistry and

Thermodynamics of Molten Lithium Fluoride—Beryllium
Fluoride by Equilibration with Gaseous Water—Hydrogen

Fluoride Mixtures,’ University of Mississippi, Oxford,
june 1965,

K. A. Romberges ef al., Reactor Chem. Div. Ann.
Progr. Rept. Dec. 31, 19635, ORNL~-3913, p. 8.

6J. E. Eorgan et al., Reactor Chem. Div. Ann. Progr.

Rept. Jan. 31, 1964, ORNL.-3594, p. 45.

7}. P, Coughlin and K. G. King, J.
2262 {1950).

72,

Am. Chem. Soc.
ORNL-DWG 66-41779

 

 

 

 

 

 

 

 

 

 

 

 

 

3000 ———— e
o500 | L &:7 |
~L+C—]
‘ /
2000 e
- /
Q .
; < T
wt ‘
r
= 1500
<
x
L
a
=
o
1000
500
0
0O 0.2 C.4 C.6 08 1.0
UO2 mole fraction ZrO2
Fig. 2.1. Revised U02-2r02 Phase Piagram. L,
liquid; C, face-centered cubic; T, face-centered tetrag-

onal; M, monoclinic.

THE OXIDE CHEMISTRY OF ThF ,-UF, MELTS

B. F. Hitch C. E. L. Bamberger

C. F. Baes, ]Jr.

The precipitation of protactinium and uranium
from LiF—BeF2-’1“hF4 mixtures by addition either
of BeQ or ThO_ was demonstrated several years
ago by Shaffer et al.® 1'% as a possible method
for processing an MSBR blanket salt. The purpose
of the present investigation is (1) to verify that
the oxide solid phase formed at equilibrium with
UFq-’I‘hE‘4—containing melts is the expected solid

81, H. Shaffer et al., Nucl. Sci. Eag. 18(2), 177 (1964).

9_]. H. Shaffer, G. M. Watson, and W. R. Grimes, Re-
actor Chem. Div. Ann. Progr. Rept. Jan. 31, 1960,
ORNL-2931, pp. 84—-90.

10_]. H. Shaffer et al,, Reactor Chem. Div. Ann. Progr.
Rept. Jan. 31, 1961, ORNL-3127, pp.- 8-11.

 

18

solution of uo, and ThC, and (2) to determine
the distribution behavior of Th** and U*” between
the oxide and the fluoride solutions,

U*T(£) + Th*'(0) = U* (o) + Th*'(f) (18)
(here f and o denote the fluoride and the oxide
phases).

The experimental technique is similar to that
used in the U0,-Z:0O, phase study (see preceding
section); Th()2 and UO, were contacted with 2LiF-
Bel®, containing U¥, and ThF, within nickel
capsules under a hydrogen atmosphere in a rocking
furmmace. The equilibrated oxide solids were al-
lowed to settle before the samples were frozen.
Good phase separation was obtained provided suf-
ficient quantity of the fluoride phase had been
added originally.

A (U-Th)O, solid solution has been found in
every sample examined thus far; the lattice param-
eter determined by x-ray diffraction'! was con-
sistent with the composition calculated for such
an oxide phase. The equilibrium quotient for the
metathesis reaction shown above was determined
by analysis of the fluoride phase for the small
amount of uranium which it contained. 'The results
obtained thus far give

X

X
0= X_Hfiu(_f_) ~ 1000 to 2000

u(f) XTh(O)

(19)

at 600°C.
oxide phase, Xu(o)’ was varied from 0.2 to 0.9
while the mole fraction of thorium in the fluoride
phase was varied from 0.01 to 0.07.

It thus appears that the Ut present is strongly
extracted from the fluoride phase by the oxide
solid scolution formed at equilibrium. This con-
firms the effective precipitation of U*" by oxide
first reported by Shaffer et al.® In addition, the
formation of a single oxide solution phase offers
a much more flexible and effective oxide separation
method for breeder blankets than would be the
case if only the separate oxides ThO, and UO,
were formed.

The mole fraction of uranium in the

llThese x-ray examinations were performed by G. D,
Brunton and D. R. Sears of Reactor Chemistry Division,
The mole fraction of ThO, in the oxide solid sclutions
was calculated from the lattice parameter using the
equation given by L. Q. Gilpatrick and C. H. Secoy,
Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965,
ORNL-3789, p. 241.
We plan to continue these measurements in order
to extend the range of oxide composition and ThF
concentration in the melts and to determine the
temperature coefficient of the distribution quotient.

THE OXIDE CHEMISTRY OF LiF.BeF,-ZrF,
MIXTURES

B. F. Hitch C. F. Baes, ]Jr.

Measurements of the solubility of BeO in LiF-
BeF , melts and of Z10, in 2LiF—BeF2 + ZtF | melts
(simulating mixtures of MSRE flush salt and fuel
salt) have been completed. As described previ-
ously,??1? the procedure consisted in withdrawing
from an oxide-saturated melt a filtered sample
which was then sparged with an HF-H, mixture
to remove the dissolved oxide as water. The total
amount of water liberated was determined by Karl
Fischer titration. The principal difficulty en-
countered with the method was in filtering the
samples. Particularly in the case of BeO-saturated
melts, the oxide crystals evidently were sometimes
small enough to pass through or to plug the filter.
The BeO solubility measurements were found to
be reproducible only after sevefal days of equili-
bration at temperatures above 600°C. Considerably
less difficulty was encountered with melts satu-
rated with Z:O -

The results of these measurements are repre-
sented adequately (> £20%) by the following ex-
pressions; the concentration unit employed is the
mole fraction:

Xizni/(nt,ilr +nBeF2+anF4) . (20
The solubility of BeO in LiF-BeF , with Xgep, =
0.3 to 0.5 and T = 750 to 1000°K, is given by

— 2625/T .

F, (21)

log Xo ,_ = —0.901 4 1.547’){’}3e
The solubility product of Z:O, in 2LiF-BeF, +
ZiF , with X 7,5, = 0.001 to 0.05 and in the same
temperature range, is given by

0o, (5, ) 02, o )

 

12C. F. Baes, Jr., and B. F. Hitch, Reactor Chem.
Div, Ann. Progr. Rept. Dec. 31, 1965, ORNL-3913, p.20.

UMSRP Semiann. Progr. Rept. Feb. 28, 1966, ORNL-
3936, p. 133,

where
log a = —1.530 — 2970/T ,

log b= 1.195 - 2055/T .

The results are reasonably consistent with pre-
vious, less direct estimates!* based upon measure-
ments of the following equilibria:

H, (0 + 2F~(d) = 0%7(d) + 2HF(g) , (23)
H,0(g) + BeF (d) = BeO(s) + 2HF(g) , (24)
2}120(g) + ZrF4(d) = ZIOQ(S) + 4HF(8) . (25)

The tolerance of MSRE flush salt for oxide should
be determined by the solubility of BeO (Fig. 2.2).
When the flush salt becomes contaminated with
enough fuel salt (~1.6% by weight) to produce a
Z1F | concentration of 0.0008 mole fraction, Z:0,
should appear as the least soluble oxide. The
oxide tolerance should then decrease with increas-
ing fuel salt content because of the mass action
effect of the increasing concentration of ZrF .
This oxide tolerance should pass through a mini-
mum at X, ™ 0.01 (Fig. 2.2), and then it should
increase because of the effect of the medium upon
Z10, solubility. The oxide tolerance of such fuel-
salt—flush-salt mixtures is given by

a
X TTR mpmmemsmssssnses oo

4+

4+') 3/2 .

Zr

(26)

The solubility of ZrO, was also measured in a
salt mixture whose composition (65.5% LiF-29.4%
Ber-—S.l% ZrF4) simulated more closely the
MSRE fuel composition than did the 2LiF-BeF , +
ZrF , mixtures. The results (Fig. 2.2) differ little
from the values given by the equation immediately
above for the case where X, 44 = 0.05.

The effect of salt composition on the Zr0, solu-
bility product may be caused, at least in part, by
specific chemical effects, that is, complex forma-
tion. For example, the form of the expression for
the wvariation in the Z10, solubility product is
consistent with — though it does not prove the
existence of - the following complex-forming
reaction,

2z0* "+ 0% = Z21,0°7. (27)

 

VAMSRP Semiann. Progr. Rept. Feb. 28, 1965, ORNL-
3812, p. 120.
ORNL-DWG 67-769

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

#(°C)
TOO 600 500
1000 — . O
_ L o T ]
N Mg L -
7y
500 o L L]
\ © "N&M
\% r “.
— Ay e
&, ) [ i I
& % @)
= M ’8“; 6
‘ 1y S
5 200 ’ e N | &Sq o~
= | 5 (3 7y
= 2 \
Zz
wl ‘
‘ - 1
o N ~~
g Lo N&%© -
é — L .,..,\\9‘4(/\ -
50 e —— Y - s
- . (2 __ ..
ool 1L L \
1.0 1.4 1.2 1.3
1000/7 (o)
Fig. 2.2, Estimated Oxide Toleronce in MSRE Salt
Mixtures. (1) Flush salt saturated with BeO, (2) flush

salt—fuel salt mixture of minimum oxide tolerance, and

(3) fuel salt,

The activity coefficient of Ze*t, 0%, and any
oxide complex of Zr*" will vary with the melt
composition, however, and no quantitative inter-
pretation of the present results in terms of com-
plex formation is warranted until mote can be
learned about such activity coefficient variations.!?®

CONSTANT-VOLUME HEAT CAPACITIES
OF MOLTEN SALTS

Stanley Cantor

From the magnitude of the constant-volume heat
capacity (C ) of molten salts one can infer what
kinds of motion are executed by the ions. An
attempt has been made, therefore, to derive ac-
curate values of C_ and to relate these values

 

e, wm Baes, Jr., Reactor Chem. Div. Ann. Progr.
Rept. Dec. 31, 1965, ORNIL.-3913, p. 23.

20

to ionic motions. In this study, C_ was calcu-

lated from the relationship

C

_ p
v 1+(a2(12MT/Cp), (28
where C_is the heat capacity at constant pressure
of 1 gram-formula weight, a is the volume expan-
sivity, p is the velocity of sound in the molten
salt, M is the gram-fommula weight, and T is the
absolute temperature. The volume expansivity is

obtained from density-temperature data by

1 <ap>
p \OT o

where p is density and p is pressure.

For systematic study, it is necessary to compare
C, of the salts at corresponding temperatures.
An obvious corresponding temperature is the melt-
ing point. Other corresponding temperatures were
defined empirically by the equation

a =

T=T,+ (J‘(TB -T,), (29)
wherte T, and T, are the normal melting and boil-
ing points respectively, and €/ is a constant that
may vary between 0 and 1.

Values of C at T, and at some other corre-
sponding temperature are given in Tables 2.2 and
2.3. An obvious limitation of this treatment
that, since virtually all available data were ob-

15

tained at atmospheric pressure, the values cal-
culated for a single salt at the various tempera-
tures (as in Table 2.2) do not refer to precisely
the same volume. A method for calculating the
effect of temperature on a truly constant-volume
heat capacity is described briefly in the following
section.

The ‘‘experimental” C ’s listed in Tables 2.2
and 2.3 exceed in almost all cases thosze calcu-
lated on the basis of contributions due to (1)
harmonic oscillation, (2) molecular rotation, (3)
intraionic vibrations. For the halides, it may be
shown that for contributions 1 through 3, the
highest calculated C would be based on the as-
sumption that each ion executes harmonic oscil-
lations in three coordinates. For instance, for
an alkaline-earth halide, if we assume only har-
monic oscillation, then C_ for 1 gram-formula weight

would be 9R cal/deg. If we assume a model for
21

Table 2.2, Constant-Yolume Heat Capacities of Molten Halides

 

Cv (cal/deg)

 

 

Halide References Calculated
T,, (melting point) &= 0.1 va (cal/deg)

LiF c, d, e 12.5, 12.4, 6R (=11.92)

LiCl £ g b 12.7, 12.5,

LiBe £ g h 12.9, 12.8,

NaF i, d,j 12.3, 12.3

NaCl f, g h 12.7, 12.6,

NaBr f, g h 12.6 12.5,

Nal c, 8 h 12.5, 12.3,

KF ¢, d, h 1l 1.

KCl ¢, & h 12.6, 12.4,

KBr £ 8 h 1.7, 115,

K1 £ 8 h 12.2, 12.0,

CsCl f g h 13.2, 13.1,

CsBr t g h 13.2, 13.1,

MgCl, ¢, k h 21.2, 211, GR (=17.48)

CaCl, 1, k & 19.5, 19.3,

CaBr, 1 k ok 24.5, 24.4,

Cal,, 1, k A 20.9, 20.8,

8rCl, 1, k, h 21.7, 21.4,

SrBr, 1, k, h 23.1, 22.7,

SiI, !, k, h 21.9 21.5

BaCl, 1, k, R 19.9, 19.7,

BaBr2 l, I, k 18.73 18.5‘5

Bal, I, &k, h 20.7 20.3

CdCl, m, g, h 24. 23., (at @ = 0.5)

CdEr, m, k, h 18., 18. , (at € = 0.5)

Cdl, m &, h 22., 21., (at f=0.5

HeCl, c, k, h 17., {at T, + 8)

HgBrz c, k, h 17'9 16.7 (at TB)

Hel e, k, A 17.4 17. (at T ) .

 

 

See Fqg. (29).

PBased on harmonic osciilations otily (see text).

°JANAF Thermochemiéal Tables.

9. Blanc ef al., Compt. Rend. 254, 2532 and 255, 3131 (1962); 258, 491 (1964).

<

f

D. G. Hill, S. Cantor, and W. T. Ward, J. Inorg. Nucl. Chem. (in press).

A. 5. Dworkin, Oak Ridge National Laboratory,_ personal communication.
81, O’M. Bockris and N. E. Richards, Proc. Roy. Soc. 241A, 44 (1957},

hG. J. Janz et al. (eds.), Molten Salt Dafa, Tech. Bull. Series, Rensselaer Polytechnic Institute, 1Troy, N.Y.,

July 1964.

K. K. Kelley, U.S. Bur. Mines Bulil. 584 {1960).

1. b. BEdwards et al., J. Electrochem. Soc. 100, 508 (1953),

k1. O'M. Bockris et al., Rev. Chim., Acad. Rep. Populaire Roumaine 7(1), 59 (1962).

’a. S. Dworkin and M. A. Bredig, J. Phys. Chem. 67, 697 (1963),

M.. E. Topol et al., J. Phys. Chem. 64, 1339 (1960).
Table 2.3. Constant-Volume Heat Capacities of Nitrates
and Sulfates at the Melting Temperature

 

CV (cal/deg)

 

 

 

Compound References @ ——— il —_—

Experimental Calculated?
LiNO3 b e, d 24_0 21.47
NaNO3 e, ¢ d 26.5 22.12
KNO3 b, c, f 23.2 22.44
AgN03 g ¢, f 25.0 21.37
LiQSO4 h, i, f 42.5 37.21
Na SO4 b, 1, f 39.6 37.27

 

“Harmonic oscillations (6R, nitrates; 9R, sulfates)
plus rotation of the nitrate or sulfate ion (1.5R) plus
vibrational contribution of nitrate or sulfate.

Pk. K. Kelley, U.S. Bur. Mines Bull. 584 {19560).

°R. Higgs and T. A. Litovitz, J. Acoust. Soc. Am. 32,
1108 (1960).

9G. P. Smith et al., J. Chem. Eng. Data 6, 493 (1961).
®G. J. Janz et al., J. Chem. Eng. Data 9, 133 (19564).

fG. J. Janz ef al. (eds.), Molten Salt Data, Tech. Bull.
Series, Rensselaer, Polytechnic Institute, Troy, N.Y.,
July 1964,

€G. J. Janz et al., J. Phys. Chem. 67, 2848 (1963).

N K. Voskresenskaya et al., Izv. Sektora Fiz.-Khim.
Analiza, Inst. Obshch, Neorgan. Khim., Akad. Nauk
SSSR 25, 150 (1954).

iM. Blanc et al., Compt. Rend. 254, 2532 and 255,
3131 (1962); 258, 491 (1964),

a liquid alkaline-earth halide in which the cations
are associated, for example, Mx]* and X, then
CV would be less than 9R; for the example chosen,
assuming that the vibration of the ion [MX]" is
fully excited, C_ equals 8R.
CV’S for mercuric salts are less than 9R, it is

Since experimental

quite probable the same associated species exist
in these salts.

Experimental C ’'s of nitrates and sulfates are
especially interesting because these salts con-
tain bona-fide complex ions, that is, nitrate or
the measured vibra-
tional spectral® provide the means for calculating
vibrational contributions of nitrate or sulfate to
the heat capacity of each salt. The rotational

sulfate ions. Furthermore,

 

1GK. Nakamoto, Infrared Spectra of Inorganic and
Coordination Compounds, pp. 92, 107, Wiley, New York,
1963.

22

contribution for these complex ions may be safely
assumed as 3/2R. Contributions due to harmonic
oscillation for the nitrates were assumed equal
to 6R, and for the sulfates, 9R. Note in Table 2.3
that the experimental C  for these salts always
exceeded the value calculated based on the con-
tributions just noted.

Two general patterns are clear from this study:
(1) experimental C, exceeds that calculated on
the basis of simple classical and/or quantum con-
tributions; (2) C, decreases with increasing vol-
umes. Pattern 1 may be partly explained by cor-
recting the notion that these ions execute harmonic
oscillations; almost certainly these ions may be
better represented as anharmonic oscillators whose
potential energy contribution exceeds the 3/21(T per
ion that is associated with harmonic motion (see
following section); this excess cannot presently
be calculated with very much accuracy. Pattem 2
probably reflects the decrease in average potential
energy per ion that occurs when the volume of the
liquid increases. In other words, the liquid ex-
hibits moie gaslike behavior as volume increases;
the kinetic energy contribution per ion probably
remains at efsz, but the potential energy contri-
bution slowly decreases with volume.

TEMPERATURE COEFFICIENT OF C
FOR MOLTEM SALTS

Stanley Cantor

As noted in the previous section, experimental
values of C_ for molten salts exceeded values
calculated from the expected contributions (i.e.,
degrees of freedom). To determine how the inter-
nal energy is changing with temperature alone,
it is necessary to evaluate C  maintaining the
volume constant with changing temperature. The
method of evaluation used here is similar to that
published by Harrison and Moelwyn-Hughes. !’

The variation of C  with volume at constant
temperature is given by

(ac) <62P>
) =T )
v/ ar?/,

1.J’D. Harrison and E. A. Moelwyn-Hughes, Proc. Hoy.
Soc. 239A, 230 (1957).

(30)
Integrating between the molar volume at a reference
pressure (VO) and any volume, V, one obtains

C -C va<agp> v, (3D
= 0—'— —-:-—---;J— ' .
v v Lo NITH

where CVO is what one actually obtaing from the

equations of the preceding section.
Moelwyn-Hughes *® has shown that, for liquids

which obey a simple potential function of the form

& = DR™" ~ BR™™ (32)

{(where D and B are constants, m and n are inte-
gers, and R is the interatomic distance), the con-
stant ¢ in a Tait equation '® of the form

1(1/f
dAB,) (33)
dP
is given by
C = 3 (m + n + 6) 34

([3T is the isothermal compressibility).

Owens’ *? data on nitrates afford the only experi-
mental tests of whether these equations are valid
for molten salts, His work suggests that the for-
mulas are obeyed at constant temperature but that
¢ varies somewhat with temperature. For NaNO .,
for example, ¢ = 4.5 at 400°C and 5.0 at 500°C,

Integration of the Tait equation between a stand-
ard reference pressute (P%) and pressure P yields

(Ve/¥° —1

0
e ) l)' = F ’
CBO
- T

(35)
where the superscripts refer to the reference con-
ditions. Successive differentiation of this with
respect to T at constant volume and substitution
in Eq. (31) yields

2
N

Fortunately, this complicated expression can be
simplified by use of the relation

B2 = AePT

T

(37)

where A and b are constants, which holds remark-
ably well for all molten salts and which is dis-
cussed more fully in the next section,

Differentiating Bgr with respect to temperature
and substituting in Eq. (36) yields

TV®
(c = DB

G

This equation was then used to generate (with the
aid of a computer) many values ot C_ for 34 salts.
For the cases of the nitrates, experimental values
of ¢ were used. For all other salts, values of ¢
were estimated from Eqs. (32) and (35) by setting
m = 1 and allowing n to vary between 5 and: 14.

T(V® — Vb2

v 0 CB;

 

da®  B?
X {c(ao)z- 20%h 4 —

 

! c

— 1} . (38)

 

 

For all salts, CV increased with temperature when
compression was nccessary ‘to maintain constant
volume. When pressure was decreased, C_  de-
creased with temperature. Some typical results are
shown in Tables 2.4 and 2.5. ‘

These increases of C_ may be correlated with a
decrease in the elastic forces holding the ions
together. Such a decrease in elastic forces for
the crystalline state usually means that the aver-
age potential energy exceeds the average kinetic
energy,2 ! Because the temperature range examined
was in the vicinity of the melting point rather

 

]

wv-vh{ 2 /dB® 1 d*B?2
C =g |~ =gz |~ | * 27 7
Yoy B2, (B \ dT (o, dT

-

 

 

dTl

2
TV° 20 dBY da® 2 dR° " 1 4*BY%°
X |'_(%___~5~} {c(a,o) 2 _1, HA_[.-T + * (pT> _—— BT l (36)

 

18,
(1951).

Y handbook of Chemistty and Physics, 44th ed.,
p- 2212, Chem. Rubber Publishing Co., Cleveland, Ohio,
1962-63.

208, B, Owens, J. Chem. Phys. 44, 3918 (1966).

A. Moelwyn-Hughes, J. Pavs. Chem. 55, 1246

o e bt e
BS dT " dT < (BY)

el dT* |

 

than the critical point, the liquid is probably more
solid-like in character; it would then follow that
these molten salts similatly possess an average

 

2 1E. Fermi, Molecules, Crystal and Quantum Statistics,
p. 156, W. A. Benjamin, Inc., New York, 1966.
Table 2.4. C_ of LiF at 14.662 cms

(V% at 1204°K)

 

 

 

 

 

 

 

T (CK) C =4 C=6 C,o
1121 (mp) 12.23 12.36 12.54
1204 12.45 12.45 12.45
1287 12.68 12.53 12.38
1370 12.89 12.59 12.32
Toble 2.5. C_ of NaNO, at 44.61 cms
(v? at 580°K)
T (°K) C=4.5 C - 5.0 C,o
580 (mp) 26.53 26.53 26.53
650 26.84 26.71 26.22
723 27.10 26.90 25,94
800 27.54 27.07 25.71

potential energy greater than the average kinetic
energy. Assuming that the kinetic energy com-
ponent of C_ is 3‘/2k per ion (for simple ions), then
it is understandable why the total C_ per ion ex-
ceeds 3k.

TEMPERATURE COEFFICIENT OF
COMPRESSIBILITY FOR MOLTEN
SALTS

Stanley Cantor

The temperature variation of isothermal com-
pressibility at 1 atm was found to follow the simple
equation (see previous section)

RO = AePT (37)

Individual values of BOT at temperature T were ob-
tained from the expression

 

(39)

(all symbols were defined in the previous two sec-

tions). The sources of data for substitution in

24

Eq. (39) are the same as those for calculating C_
(see Tables 2.2 and 2.3).

The magnitudes of the constants A and b are
given in Table 2.6.
coirelated with the position of the ions in the
periodic table. The 2nomalous values for MgCl,

Both constants are roughly

reflect the observation that sonic velocity in this

medium does not vary with temperature; this ob-
servation is most likely

Siace the constants A and b are rather restricted

erroneocu s,

in magnitude, it is an easy matter to empirically
For
instance, one might predict, by a rough interpola-
tion between KCl and CsCl, that for RbCI, A -
65%10" 12 and b =177 x 1077,

The reasons that simple Eq. (37) successfully
comrelates (3 with T arc not as yet known. The

estimate [R?% for salts as yet unmeasured.
T y

Table 2.6. Constants for Equation of Isothermal

Compressibility at 1 atm vs Temperature

 

 

 

‘Bl;fi(cmz/dyne) = AebT(O <)

Salt A b Salt A b

10712 x107? x107'% x 1073
LiF 2.31 1.35 CdCl2 11.6 1.07
LiClt 5.66 1.38 Cd}f_’»r2 13.5 1.49
L.iBr 6.81 1.33 CdI2 17.9 1.32
NaF 2.29 1.51 HgBr2 21.1 1.78
Na(Cl 4,92 l1.61 Hg12 10.9 3.34
NaBr 6.64 1.54
Nal 8.63 1.59 LiNO3 8.99 1.43
KF 3.83 1.53 N&lNO3 6.20 1.89
KC1 5.88 1.74 KNO3 6.49 2.08
KBr 7.20 1,72 AgNO, 5.23 1.53
KI 8.13 1.82
CsCli 6.92 1.82 LiZSO4 3.36 1.00
CsBr 7.39 1.94 N32504 4,16 1.02
Mg(Cl 44.41 0.239
CaCl 4.42 1.07
CaBr, 6.06 0.569
Ca12 8.30 1.13
SrCl1 4.14 1.00
SrGr 4_.80 1.06
SrI2 7.51 1.03
f3aCl 3.70 1.08
BaDBr 3.94 1.22
BaI2 5.88 1.13

 

 

 
narrow ranges for A and b are perhaps more under-
standable. Compressibility at low pressure most
likely represents squeezing against the volume not
occupied by the ions, that is, the “‘free” volume.
Since the forces exerted by the ions on this free
volume should be virtually independent of the kind
of ions, it follows that increases in free volume
which occur with temperature should be roughly
the same for all ionic liguids; hence the tempera-
ture dependence of squeezing against the free
volume would also be essentially independent of
the ions involved.

ViSCOSITY AND DENSITY IN THE LiF-BeF,
SYSTEM

C. T. Moynihan %2 Stanley Cantor

Densities and viscosities of selected composi-
tions in the LiF-Bel  system were measured to
determine if temperature coefficients of viscosily
are correlated with coefficients of volume expan-
sion.

The viscosity of pure BeF, was measured over
the temperature range 573 to 985°C, using Brook-
field LVT and HBF 5X viscometers. In this range
the viscosity, 7, varied from 10 to 10° poises and
was about 10% greater than previously reported. 23
The plot of log n vs 1/T(°K) showed only slight
curvature. A least-squares fit of the data to an
equation quadratic in 1/7T yielded

log  (poises) = —8.119

1.1494 » 10*  6.39 x 10°
+ + (40)

T (°K) T?

 

(standard error in log # = 0.013). This equation
yields an activation energy for viscous flow of
57.3 kcal/mole at 985°C and 59.6 kcal/mole at
575°C. These are relatively small activation en-
ergy changes; thus the temperature dependence
of viscosity for BeF, is basically Arthenius over
the range covered in this investigation (10 to 10°
poises).

For ,’LiF—BeF2
confimed earlier results?® which showed marked

mixtures, viscosity measurements

 

22Chemistry Department, California State College at
Los Angeles; summer participant, 1966,

235, Cantor and W. T. Ward, Reactor Chem. Div. Ann.
Progr. Rept. Dec. 31, 1965, ORNL-3913, pp. 27-_28.

Table 2.7. Densities, Molar Yolumes, and
Exponsivities of LiF-E’n’eF2 Melts at 800°C

 

 

Mole Fraction Density Molar Volume
BeF , (g/cms) (Cms) Expansivity
0.000 1.831° 14.17° 2.6 x10™*
0.502 1.894 19.28 2.3
0.749 1.902 21.93 1.3
0.892 1.915 23.35 0.8
1.000 1.921% 24.47° 0.4°

 

aExtrapolafed below freezing point of salt.

b s P
Extrapolated from composition-volume or composition-
expansivity curves.

decreases of viscosity and its temperature coef-
ficient with LiF concentration.

The densities of three BeF Z—LiF mixtures (see
Table 2.7) were detemined by the Archimedean
method by measuring the apparent weight loss of
a platinum sinker upon immersion in the melt.
Volume expansion coefficients (i.e., expansivities)
derived from the density data decreased with de-
creasing LiF concentration. Molar volumes, also
derived from these data, appeared to be additive
at 800°C.

Density measurements of pure BeF
tempted by adapting the Archimedean method to
the technique of zero velocity extrapolation.?*

were at-

In none of the four attempts to measure the BeFQ
density was it possible to eliminate the few small
bubbles that adhered to the sinker. The buoyant
effect of the bubbles leads to low values of the
apparent weight of the sinker and hence the high
values of the density. If a surface tension of
200 dynes/cm is assumed for BeF , the best of
our density results yielded a value of 1.96 +0.01
g/cm” for BeF  at 850°C. This result must be
considered only as an upper limit to the real BeF
density, but it may be compared to 1.95 % 0.01
g/cm?® at 800°C reported by MacKenzie?® and to
1.968 g/cm® measured for the BeF  glass at room
temperature. These results suggest that the ex-
pansivity of liquid BeF2 is quite small.

 

247,. Shartsis and S. Spinner, f. Res. Natl, Boar. Std
46, 176 (1951).

25y D. MacKenzie, J. Chem. Phys. 32, 1150 (1960).
The results thus far obtained indicate that the
temperaiure coefficient of viscosity decreases with
increasing volume expansion coefiicient and in
particular (1) the Arthenius behavior of pure Bel" ,
like that of SiO, is associated with the tempera-
ture-independence of ‘‘free’’ volume (i.e., volume
vnoccupied by the ions),?® and (2) LiF, when
dissolved in BeF , not only breaks up the network
of linkages between beryllium and fluorine, but
also increases the temperature dependence of the
thereby decreasing the activation

free volume,

energy of viscous flow.

VAPOR PRESSURES OF MOLTEN FLUORIDE
MIXTURES

Stanley Cantor W. T. Ward
C. E. Roberts

Transpiraticn Studies in Support of the
Yecuum Distillation Process

To determine the equilibrium vapor separation of
rare-earth fission products from MSR carrier salts,
a series of vapor pressures have been measured
by the transpiration (i.e., gas-entrainment) method.
The melis were composed of 87.5-11.9-0.6 mole %
LiF-BeF -LaF .
Bel" | are approximately those expected in the still
pot of the vacuum distillation process. The lan-
thanum concentration is many times greater than

The concentrations of LiFF and

what would be permitted as total rare-earth concen-
tration in the still; this high concentration of lan-
thanum in the melt was required in order to give
lanthanum concentrations in the vapor that were
high enough to analyze.

Measurements were carried ouf in the temperature
interval 1000 to 1062°C; dry argon, the entraining
gas, flowed over each melt at the rate of about
30 em®/min. Salt vapor, condensing in a nickel
tube, was analyzed by spectrochemical and neutron
activation methods. The latter method gave higher,
more consistent, and probably more reliable lan-
thanum analyses. The most consistent results have
been obtained at the two temperatures shown below.

QGP. E. Macede and T. A. Litovitz, J. Chem. Phvs,
42, 1 (1965).

20

 

Decontamination Factor®
Temperature for Lanthanum
1000°C 910
1028°C 1150
“Defined as (mole fraction of lanthanum in lig-

uid)/(mole fraction of lanthanum in vapor).

At six other temperatures, transpiration pressure
measurements have yielded much higher (up to
7300) decontamination factors; however, these de-
terminations either were based on insufficient
sample or else duplicate analyses gave widely
It did appear that the higher the
temperature the higher the decontamination factors.

Although much more study is required before the

vacuum distillation process is shown to be prac-

different results.

tical, it seems that decontamination factors close
to 1000 can probably be demonstrated.

Vapor Pressures of 73-27 Mole % LiF-UF

The manometric pressure of this mixture, which
is the composition of the MSRE fuel concentrate,
was measured by the Rodebush-Dixon method?7 in
the temperature range 1090 to 1291°C. The results
fit the equation

log p(mm) = 7.744 — 10,040/T(°K) . (41)

Transgpiration studies have also been cairied out
to determmine the composition of the vapor. The
results to date indicate that at 1050°C, the mole
ratio of LiF to UF  in the vapor equals 3.3.

POTENTIOMETRIC MEASUREMENTS IN
MOLTEN FLUORIDES

A. R. Nichols, Jr. 28 K. A. Romberger
C. F. Baes, ]Jr.

Continuing the program of potentiometric mea-
surements in molten fluorides, it is planned tfo

. H. Rodebush and A. L. Dixon, Phys. Rev.
851 (1925).

28Visiting scientist, Sonoma State College, Rohnert
Park, Calif.

 

26,
explore the chemistry and thermodynamics of a
variety of substances which can occur in a molten-
either in container materials or as
fission products. Previously, the development of
an H_ HF/F~ electrode and a Beo/Be2+ electrode
as suitable reference electrode half-cells for use
in 2LiF-BelF, melts has been described. 29

Presently investigations are being made of nio-
bium in molten ZLiF«BeF . Several attempts to
measure the voltage of the cell

salt reactor,

Be’|BeF ,,LiF |{BeF ,LiF,NbF, [Nb®  (42)
(in which the electrode compartments were of
graphite and, later, copper) have been only par-
tially successful because the voltage of this cell
invariably decreased with time, It is infecred, but
has not yet been proved, that the cell voltage
decreased because a small amount of beryllium
metal dissolved in the 2LiF-BeF  solvent and
intemally shorted the cell. However, before these
decreases became marked, reasonably stable
voliages were observed, in one case for as long
as three days. |

Measurements of the cell voltagge were made asg a
function of increasing NbFX concentration. (NbF ,
was added to the melt by anodic dissolution of nio-
bium metal.) The voltage first increased and then
remained constant, an indication that the melt
had become saturated with NbFX. Based upon the
number of faradays which were required to reach
saturation and the amount of melt within the nio-
bium compartment, it is estimated that at 597°C
the solubility was ~3 x 1077 equivalent of the
niobium salt per kilogram of solvent. The cell
voltage at saturation was 0.96 1 0.04 v. This value
can be combined with the free energy of formation
of dissolved BeF , (-214.6 keal /mole)3? to yield
the free energy of formation ot NbF ,

1 1
SAG e AT

-

BeF 2((1) + FE

=851 +0.9kcal (43)

 

295, Dirian, K. A. Romberger, and C. F. Baes, Jr.,
Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965,
ORNIL.-3789, pp. 76—-79.

39A. L. Mathews and C. F. Baes, Jr., ORNL-TM-1129,
pp. 7475 (May 7, 1965),

27

(F is the Faraday constant, 23.06 kcal per equiv-
alent). This preliminary result suggests a sur-
prising stability of the lower fluoride. Compared to
the HZ,HF/F“‘ reference couple, the NbFX/NbO
couple would have the potential

1 1
~NbF  + e-::;::-}_... Nbo
x X Ty

+F7, E? (897°C) ~ ~0.85v. (44)
This indicates that Nb° should be more electro-
positive than chromium in QLiF-Ber. This result
is difficult to reconcile with the evident corrosion
resistance of Nb toward I\IaI*"~ZrF4-UF4 melts ob-
served in early loop tests,”! but it is reasonably
consistent with potentials reported recently by
Senderoff and Mellors. 3%

The beryllium electrode has now been replaced
by an H_,HF/F~ half-cell. The latter, while less
convenient to use than the Be?/Be?" half-cell, is
expected to yield voltages which are more stable
and reproducible. No results are yet available from
this new cell.

The value of X, that is, the oxidation state of
the niobium in the presence of the metal and the
fluoride salt, appears to be +2. This is a tentative
conclusion based upon the results of a transpira-
tion experiment in which a known amount of gas-
ecous HF was used to partially oxidize niobium
metal in contact with the melt. The value of the
oxidation number was then determined from the
amount of metal which was consumed. No apparent
evolution of NbFS occurred, which is consistent
with a stable lower valence state.

APPEARANCE POTENTIALS OF LITHIUM
FLUORIDE AND LITHIUM BERYLLIUM
FLUORIDE IONS

R. A. Strehlow J. D. Redman

A study was made of appearance potentials of
ions fomed by electron impact from LiF and
Li, BeF , vapor. This work was undertaken in order

 

31z, A. Kovachevich, B. L. Long, and D. F. Stone-

burner, Results of Niobium Thermal Convection Loop
Tests, ORNL-CF-57-1-161.

323, senderoff and G. W. Mellors, J. Electrochem, Soc,
113(1), 66 (1966).
to assist in the interpretation of sublimation heats
determined mass spectrometrically. A suiprising
amount of structure was found in the ionization
efficiency curves, and it is this aspect of the work
which is emphasized here.

The study of sublimation heats with a mass
spectrometer requires either a knowledge of frag-
meittation patterns of polymeric vapor molecules
or the assumption that a given ion, for example,
Li’
This assumption, called the specificity rule, has
been found not to apply to lithium halides.®?® The
fragmentation patterns of (LiF)2 have been the
33,34 In addition,

: +
or Li F =, has only one neutral precursor.

subject of study. for lithium-
beryllium fluoride species, the neced to postulate
structures such as (I) to account for the mass
spectrometric observations®® led to the expectation
that a detailed study of appearance potentials
might help to clarify the phenomena which have

been observed.

The ionization efficiency (I[.E.} curves obtained
in this study possess structure to an unexpected
degree (considering that the vapor density for some
of the species corresponded to less than 102
torr). Selected ILE. curves are shown in Fig. 2.3.

The data were obtained using the retarding po-
tential difference method and a Bendix time-of-
flight mass spectrometer. determina-
tions were made for a reference gas with appropriate
appearance potentials and for two salt ions, The
reference gas was admitted (at a pressure of 1 x
10~ 7 torr) to permit voltage calibration as well
as to detect possible instrumental vagaries. The
simultaneous monitoring of two salt vapor peaks

Simultaneous

33]. Berkowitz, II. A. Tasman, and W. A. Chupka,
J. Chem. Phys. 36, 217079 (1962).
34P. A. Akishin, I.. N. Gorokhov, and I.. N. Siderov,

Russ. J. Phys. Chem. 33, 64849 (1659).

35A. Blichler and J. L. Stauffer, Symp. on Thermo-
dynamics with Emphasis on Nuclear Materials and
Atemic Transport in Solids, Vienna, 22..27 July 1965,

eliminated temperature changes of the furnace as
a serious source of error.

As has often been obseived for other gases, a
current of singly charged ions appears at some
onset potential and then increases linearly with
electron encergy until a subseguent appearance
potential (AP} due to an added process of ion
generation. These AP’s f{or fragment ions mav

arise in several ways: ion pair formation, ion-
neutral reaction, rearrangement, fragmentation of
the neutral moiety, formation of excited states
in the neutral or ion, and others. Unambiguous
assigiiment of a process to an AP other than onset
is not often possible. For the species studied
here most of these possibilities, however, may be
eliminated. Table 2.8 lists the APP’s found for
various of the ions from LiE and Li ,BeF . The
range of values zand the number of determinations
The
existence of API{Li*) appears to be slight, but

are shown to give an indication of precision.

real. There sesms to be little question of the ex-
istence of structure for the other species.

The observed onset appearance potential for Li*
of 11.21 v with the ionization potential (Li*) of
5.36 v leads to a value of D(LiF) =5.84 1 0.10 ev,

~F

L

The onset is therefore probably char-

which is in agreement with the literature value
5.95 cv.3°
acterized by formation of Li® and F° in ground
states from LiF in its ground state. The small dif-
ferences between the onsets for Li”, Liz}?‘+, aidd
Li3F2+ are believed to be due principally to the
successively greater values of the bond strengths,
D(I.,izF — ) and D(List——F). This belief is cor-
roborated by a consideration of the appearance

a7

potentials for negative ions. Using the value

of 2.90 for the electron affinitv of ¥% and con-
sidering the process
Li,F s Li T +F~, AP =355v, (45)

one obtains D(L.i,F—F) = &.45 ev. This value is
about 0.5 ev greater than ILiF) and does not
conflict with the value of

[AP (Li F*) — AP (Li")] = 0.14 ev,
since the ionization potential of the neutral Li ,F

may be less than that for J.i. Similar reasoning

SGL- Brewer and E. Brackett, Chem. Rev. 61, 425 (1961).
3711, Ebinghaus, Z. Naturforsch, 19a, 727..32 (1964).
7Y POSITIVE 10N CURRENT (arbitrary units)

29

ORNL-DWG 67-770

 

A

 

o

« B > »

Art1x40™7 torr /
~7

KrT 410 torr REFERENCE GASESS® ...

CoHg 110 torr Cp Hg

i
Li,FY
LigFp '

. +
Li BeF,

BeF,*

| \ 4

 

/1

FROM LiF

 

FROM Li,BeF,

 

 

 

 

 
  
     
 
 
 
 

11

 

 

 

 

 

 

 

 

 

 

o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o
It pos

7 4

 

 

 

 

 

 

 

 

 

Fig. 2.3. lonization Efficiency Curves.

[VE 1ON CURRENT f{arbitrary unifs}

e
T
1
30

Table 2.8. Appezarance Fotentials of lons from L.iF and LiEBeF4 Vapor

T = 800 to 920°C

Appearance Potential

 

 

 

 

 

Sample Ion Quantity? e e
I 11 11 v
LiF Lit AP 11.21 11.73 13.59
w £0.10 +0.24 — 0.13 £6.24 - 0.36
n 5 5 5
LiF" AP 11.35 11.84 12.78 13.51
w £0.16 +0.40 — 0.13 £0.42 10,17
n 11 11 9 6
Li¥, AP 11.61 12.12 13.69
w £0.10 +0.30 — 0.40 +0.23 — 0.19
n 3 3 3
Li, BeF Li" AD (11.7) (12.2) (14.1)
n 1 1 1
Lif' AP 11.90 12.44 13.72
w +0.22 - 0.12 +0.19 - 0.14 £0.15
i 3 3 3
LiBeF AP 12.79 13.46 14.52 (14.9)
w £0.02 £0.02 £0.29
n 2 2 2 1
Ber " AD 15.38 (16.31) 17,22
w 10,02 10,07 10.10 — 0.08
n 3 3 3
Reference gases
C i, CzH; AP 11,60 (Li-F ions)
Kr Kr' AP 14.00 (LiBe¥ ")
Ar At AP 15.75 (BeF )

 

a - . . . .
AY = appearance potential in volts; w —=range; n = number of determinations.
applies to D(Li,F, —F). From other data, 3% es-
timates of some other bond strengths and electron
affinities for some of the pertinent species may

be made. These include:
D(LiF —~Li) 3.0 ev
D(LiF , — Ld) 8.0
D(LAF —F) 1.6
E (LiF — LiF) 3.35
EALD) 0.29
—0.25

EA(LA,F)

The value for EA(Li F) indicates the general de-
gree of uncertainty in these values.

With the derived value of D(LiF2 —Li) = B.00 ev,
one would anticipate the possibility of producing
Li* from the dimer at an AP of 8.00 + 5.36 =
13.36 v. Our value of AP (Li*) = 13.6 + 0.3 indi-
cates that this process is responsible for APHI(Lf ).
The second appearance potential, APH(Li*), is,
it real, not readily explainable since the lowest
excitation level for F° is 14.4 ev and for Li',
much more. Invoking possible precursor excitation
is not a happy explanation in view of the 0.5v
difference between API(LiJ’) and API{(L1+) and the
agreement of our value of D{LiF) with the literature.

The onset potential for the mixed species LiBelF *
can be used to estimate a D(LiBeF, —F) = 7.9 v,
which is near D(BeF) = 8.02.%° This indicates that
the fluoride atoms are all associated with the
beryllium in the vapor species LiBeF . Consid-
eration of the energies and bond strengths and the
near agreement between API(Li2F+) from LiF with

that from Li BeF, make it seem unlikely that
Li BeF is a precursor of Li2F+.

For BeF 2+ an additional set of possibilities must
be considered as causes of structure in the LE.
curve, These include autoionization and meta-
stable ion formation. Since all of the ions except
possibly Li* present marked similarity of structure,
it seems most plausible that even for the species
from Li BeF the structure observed is attributable
to ionic excitation, either vibrational or electronic.

The energies determined can be used to predict
possible appearance potentials beyond the limited
range used in these studies. For examples,

 

(LiF) , ~——> LiF* + LiF , 14.6v, (46)
—3> LiFf 4+ Li+F, 206v, 47)

LiF ——3» Lit + F", 25t0 28 v,  (48)
3> Li%+ F", 23 v. (49)

The ion Li _‘zBeF;r was also observed but in too
small amounts for AP determinations. During the
melting of one sample of Li ZBeF4 many bursts
of CO2 were observed; these bursts total about
2 x 107 % atm cc g~ !, but we believe this did not
cause any irregularity in the Knudsen cell opera-
tion.

Some initial determinations were made of slopes
of log (I'"T) vs 1/T for LiF species in order to
provide some comparison with the literature values
before undertaking the more difficult task of study-
ing the LiF-BeF , System vaporization. The results
of numerous determinations are shown in Table 2.9
without further discussion of details.

Table 2.9, Second Law Apparent Af*s for lons from LiF

 

 

Average
This Work a b ¢ d
.+ *
I.i 65.7 £ 0.7 67.4
Fr 64.1 +1.3
R
LiF 62.3 + 0.6 64.8 £2 66.5 1 62.4 1.5 (62.7)
. +
Li,F 67.7 1.3 68.3 +2 71.6 t2 65.3 t 1.7 (70.4)
+
Li F, 70.4 £ 0.7 73.9 3 74.9 £1.0

 

®A. Bichler, CPIA Publ. No. 44 (February 1964).

Pp. 1. Hildenbrand et al., J. Chem. Phys. 40, 288290 (1961).
“p. A, Akishin, L. N. Gorokhov, and L. N. Siderov, Russ. J. Phys. Chem. 33, 648—_49 (1959).
da. C. P. Pugh and R. F. Barrow, Trans. Faraday Scc. 54, 6§71 {(1958). Note: Values are torsion-effusion AH's

(sublimation heats for monomer and dimer).
3. Separations Chemistry

REMOVAL OF JODIDE FROM LiF-BeF, MELTS

C. E. L. Bamberger C. F. Baes, ]r.
The removal of iodide from LiF-Be¥F  mixtures
by HF sparging, presumably by the reaction

HF(g) + I™(d) « F~(d) + HI(g) , (1

was previously described!’? as a promising method
for removing 6.7-hr '3°I from an MSBR fuel at a
rate greater than the rate of decay of this nuclide
to !3%Xe.

One difficulty with the results rteported pre-
viously was that poor recoveries (typically 80%)
of iodide were obtained by HIF sparging. In con-
tinued studies the cause of this has been traced
to small particles of salt entrained in the gaseous
mixture of HF, HI, and H, emerging froimm the re-
action vessel. Evidently these particles caused
condensation of the HF and HI with water vapor
in the NaOH bubbler used to trap the HI. The
condensed droplets of acidic solution, readily
visible as a fog in the gas phase of the trap,
evidently did not react completely with the NaOH
solution, thus causing low recoveries. Introduc-
tion of a filter of sintered Teflon or gold in the
effluent gas stream ahead of the trap has resulted
in iodide recoveries greater than 95%.

It was ifound previously that the fraction of
iodide the melt ((I7]/[1719 de-
creased logarithmically with the number of moles
of HF passed per kilogram of melt (a  _/w),

remaining in

In (I~ VI71%) = - Q(a,, . /#) . (2)

 

n. F. Freasier, C. F. Baes, Jr., and H. H. Stone,
MSR Program Semiann. Progr., Rept. Aug. 31, 1965,
ORNIL.-3872, p. 127,

2B. F. ¥reasier, C. . Baes, Jr., and k. H. Stone,
Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1965,
ORNL-3913, p. 28.

32

and lrradiation Behaovior

With the assumption that reaction (1) was the
only reaction involved, the negative slope (Q/2.3)
of a plot of log (I71/[I71% vs n, _/w was
equated to the equilibrium quotient “QI of re-
action (1),

Q B PH I
B

HIEF

 

kg/mole . (3)

This interpretation of the previous data was sup-
ported by the observation that, within the scatter
of the measurements, @ was independent of the
flow rate and the paitial pressure of HF. In the
subsequent measurements, however, a fuller study
of the effect of the HF pressure at different tem-
peratures and as a function of melt composition
has shown a pronounced effect. It has been
found (Fig. 3.1) that Q varies approximately in-

versely with PHF as follows:

1
B‘iaerPHF.

ORMNL-DWG GT-774

 

008 T |
-‘ 7=520°C | /
007 ’~ ©a 33mole% BeF, ot - Fai -
| * 42mole % Bef, ‘
0o€ ! .——© 50mole % Bef, . _

   
    

<
o
o

1/22 (molas/kg)

 

0 005 040 045 o 0.20
PARTIAL. PRESSURE OF WHF (atm)
Fig. 3.1. VYariation of Q for lodide Removal {(£q. 2)

with the HF Sparging Pressure in LiF-BeF2 Melts,
While the cause of this effect is still being in-
vestigated, it may be noted that if, in addition
to reaction (1), there is an appreciable solu-
bility of HI in the melt,

HI(g) = HKd); O, = [HIV/P,, , )

the above expression can be accounted for with
a=1/Q and b = Q,. Sparging with HI as well
as HF is presently being used to determine
whether or not equilibrium conditions are being
attained during the measurements and to determine
the solubility or other possible reactions of HI
in these melts.

Whether the dependence of @ on PHF is due to
a rate effect or to the occurrence of other equi-
libria in addition to reaction (1), it seems evident
that, in the application of this treatment to the
processing of an MSBR fuel, better HF utilization
(higher Q values) will be obtained at lower HF
partial pressures, The limiting (maximum) value
of Q) obtained as P approaches zero is plotted
vs the mole fraction of BeF, in Fig. 3.2.

ORNL-DWG 67-772

 

 

LIMITING VALUE OF @ (kg /mole)

 

 

 

 

 

 

 

 

 

mole % Ber

Fig. 3.2. The Dependence of the Limiting Value of O

upon the BeF2 Concentration in LiF-BeF2 Melts.

REMOVAL OF RARE EARTHS FROM MOLTEN
FLUORIDES BY SIMULTANEGUS
PRECIPITATION WITH UF,

F. A Doss H.F.McDuffie J. H. Shaffer

The relatively low solubility of UF, in fluoride
mixtures of interest to the MSR program?® and the
known similarities of the crystal structure of rare~
earth trifluorides with UF34 provide a basis for
studies of the precipitation of solid solutions of
these compounds from fluoride melts.  Since
fission product rare earths represent a major
portion of the poison fraction in the fuel of a
molten-salt nuclear reactor, this study may be
applicable toward the development of suitable
reprocessing methods for rare~earth removal.
Initial experiments conducted in this program con-
sidered the reduction of UF, contained in a re-
actor fuel mixture to UF, and the simultaneous
precipitation of rare-earth trifluvorides with UF,
as the temperature of the fuel mixture was re-
duced. A second series of experiments is in
progress to examine the precipitation of rare
earths from a simulated fuel solvent upon addi-
tion of solid UF ,.

If all UF, contained in the current MSRE fuel
mixture, LiF-BeF -ZrF -UF, (65.0-29.1-5.0-0.9
mole % respectively), were reduced to UF,, the
solution would be saturated with UF, at approxi-
mately 725°C. By lowering the melt temperature
to 550°C, approximately 83.5% of the uranium
would be precipitated from solution. Results of
preliminary experiments designed to investigate
this reprocessing method demonstrated that LaF ,
CeF,, and NdF, could be precipitated with UF ,.
Europium and samarium were probably reduced to
their divalent states by the in situ reduction of
uranium with added zirconium metal and showed
little or no loss from solution during the pre-
cipitation of UF,. Subsequent experiments with
excess reducing agent showed that cerium removal
could be related to the U®" concentration in solu-
tion by the equation

 

InN, . =kln NU3+_ + const, (5)

3Reactor Chem. Div. Ann. Progr. Rept. Jan. 31,
1964, ORNI1.-3591, p. 50,

4Reactor Chem. Div. Ann. Progr. Rept. fan. 31,

1965, ORNL.-3789, p. 16.
ORNL-DWG 66 -11460

  

 

 

. 7 It,MPI—_RATUfiE RANGE: 850-5320 °C. EXPT Ce-2]
O ! i i —
= 6 —_— e—— S by Y — - »-
= [ Ol
é) l.l')(‘2 5 | _ . XA‘ . V/.:!;{’Q. |
T /'/"
Z 5 T ST T
5 E 3 e I _ — —
3 :
L ar ; i
= 5 : i ! i :
5 £
=2 Z o2l o INITIAL U CONCENTRATION _|
I[}:J } ~ ; o 11 wt %
© -1.5 . __‘—.__%x_._xh5\WI O/u__._‘...._._...‘_‘
1, 2 3 4 5 6 8 10 15 20 25
URANIUM FOUND IN SOLUTION (mole fraction x 103)
Fig. 3.3. Simultaneous Precipitation of CBF3 and

UFS from Simulated MSRE Fuel Mixture.

are respective mole frac-

tions of rare earth and trivalent uranium. As
illustrated by Fig. 3.3, a value of about 0.55
was obtained for k in Eq. (5) for the simultaneocus
precipitation of CeF .. Further
would be needed to verify this experimental re-
lationship for other rare earths of interest to the

where NRE and N 3+

investigation

prograim,

A more recent experimental program has been
concerned with the retention of
fluorides on a bed of solid UF,
reprocessing technique. In the first experiment
UF, was added in 30-g increments to approxi-
mately 2.2 kg of LiF-BeF, (66-34 mole %) that
initially contained 10~* mole fraction of C,el:"3
with about 1 me of '**Ce as a radiotracer. Fil-
tered samples of the salt mixture were taken
approximately 48 hr after each addition of UF,
and analyzed radiochemically for cerium. The
results illustrated a somewhat linear decrease
in cerium concentration as UF_  was added and
corresponded to a solid phase which contained
about 1 mole % CeF .

tained in a separate experiment

rare-garth  tri-
as an alternate

Similar results were ob-
with NdF_,
except that the solid phase corresponded to about
0.2 mole % NdF , in UF .

EXTRACTION OF RARE EARTHS
FROM MOLTEN FLUORIDES
INTO MOLTEN METALS

J. H. Shaffer F. F. Blankenship
W. P. Teichert W. K. R. Finnell
D. M. Moulton W. R. Grimes

This experimental program has been oriented

toward

the development of a liquid-liquid ex-

34

traction process for removing rare-earth fission
products from the fuel of a two-tegion molten-salt
breeder reactor.
for the reference desipn MSBR, uranium will be
removed by fluorination. Thus, for purposes of
this the barren fuel solvent has
been simulated by dissolving selected rare-earth
fluorides into a mixture containing 66 mole %
LiF and 34 mole % Bel",. When this mixture is
contacted with a molten bismuth-lithium mixture,
rare earths are reduced to the metallic state and
dissolved in the molten metal phase. ‘The program
further envisions a similar back-extraction process

In processing schemes proposed

investigation,

for concentrating rare-earth fission products in a
second salt mixture for disposal or further utili-
zation.  Experimenis conducted thus far have
examined the distribution of rare carths between
the two liquid phases as functions of the lithium
conceiltration in the metal phase. Studies of the

equilibrium

2Li% + BeF = 2LiF + Be® (6)
in the extraction systein are currently in progress
to ascertain activity cecefficients of lithium and
raie earths in bismuth and to study effects of salt
composition on rare-earth distribution coefficieuts.

Fluoride starting materials were prepared in
nicke! equipment by treatment with HF-H, mix-
tures at 600°C to remove oxide impurities and at
700°C with H, alone to reduce concentrations of
structural metal difluorides in the fluoride melts.
Selected rare-earth fluorides were added prior to
this treatment in quantities sufficient to attain
concentrations of about 10~ * mole fraction in the
salt mixture. Bismuth was further purified by
treatment with H, at 600°C in the 204L stainless
steel, low-carbon-steel-lined extraction vessel,
Following this treatment the prepared salt mixture
was transferred as a liquid to the extraction
vessel, Each experiment typically contained
2.35 kg of bismuth and about 2 kg of the salt
Lithiuin, for incremental additions to
the experiment, was freshily cut and tared under
mineral oil, affixed to a small-diameter steel rod,
rinsed in benzene, and dried in the flowing inert
atmosphere of the loading port prior to its in-
sertion into the molten bismuth. This loading pott
extended near the bottom of the extraction vessel
to avoid contact of lithium with the salt phase
prior its dissolution into the
phase. Filtered samples of each phase were taken
under assumed equilibrium conditions after each

mixture.

to molten metal
addition of lithium. Radiochemical analyses of
each phase for rate-earth gamma activity and
spectrographic analyses of the metal phase for
rare-earth and lithium concentrations provided
data for calculating the distribution of rare earth
in the system and its dependence on the lithium
concentration of the metal phase. A summary of
these results, illustrated in Fig. 3.4, shows that
a mixture containing 0.02 mole fraction of lithium
metal sufficed for removing essentially all cerium,
lanthanum, and neodymium and substantial quan-
tities of samarium and europium from the barren
fuel solvent under separate but comparative con-
ditions. In all experiments rare earths that were
reduced from solution in the salt phase were
found as dissolved components of the metal
phase.

The reduction of rare-earth fluorides by lithium
is expected to proceed by the reaction

mLi® + (REY" == (RE)® + mLi", (7)

where m is the effective valence of the rare-earth
cation, If unit activities prevail for all metal
species in the salt phase and for all ionic species
in the metal phase, then the activity of lithium
dissolved in the metal phase can be expressed

QRNL--DWGE 66-4783

 

 
 
 
 
 
  

 

500
11 O SAMARIUM
@ & CERIUM
<[
X 400 - B LANTHANUM -
. ® NEODYMIUM
': A ELROPIUM

 

 

 

FRACTION OF RARE EARTH IN METAL PHASE

100

 

o

 

 

MOLE FRACTION OF RARE EARTH IN SAL

MGL

 

 

 

 

O _ _—
o i 2 3 4 5
LITHIUM FOUND IN METAL PHASE {mole frociion x '102)

 

Fig. 3.4. Extraction of Rare Earths from LiF-Ber
(66-34 Mole %) into Bismuth by the Addition of Lithium
Metal at 600°C.

35

as a function of other activities in the system as

g O)metal (‘4
)

(A LiF .fsnalt
Am — R
Li® Ka (A

 

(8)

RE"salt

By assuming that the activity of LiF and the
activity coefficients of Li% RE®, and RE™* re-
main constant, the dependence of rare-earth dis-
tribution on the lithium concentration can be
expressed as

L

where D is the ratio of the mole fraction of rare
earth in the metal phase to the mole fraction of
rare earth in the salt phase and

- m >
K, (}/1 io)metal redsan

 

— (10)
(yREO)metaI‘(ALiF)

A plot of the experimental data according to the
logarithmic form of Eq. (9) is shown as Fig. 3.5.
Values for m and K _ calculated from the slopes
and intercepts of this plot are as follows:

Rare Earth m KQ

Lanthanum 2.7 2.5 x 107
Cerium 2.3 3.8 x 10°
Neodymium 2.5 2.5 % 10°
Samarium 1.6 1.8 < 10*
Europium 1.9 5.9.% 103

Although the apparent fractional exponents for
the reductions are as yet unexplained, the results

ORNL—DWG BE—~ 2425

   

LITHIUM FOUND IN METAL PHASE
(mole fraction % 10%)

 

O Lo e e e
oo Lol LB L L L
Q.01 oA 1 10 W00 1000
MOLE FRACTION OF RARE FARTH IN METAL
MOLE FRACTION OF RARE EARTH IN SALT
Fig. 3.5. Effect of Lithium Concentration in Metal

Phase on the Distribution of Rare Earths Between LiF-
BeF2 (66-34 Mole %) and Bismuth at 600°C
are in rough agreement with the occurrence of
lanthanum, cerium, and neodymium as trivalent
ions in the salt mixture; samarium and europium
are probably reduced to their divalent states
prior to their extraction into the metal phase.

In earlier experiments the extraction of rare
earths from a salt phase into molten bismuth was
achieved by the addition of beryllium metal to
the system.®
sulted

This reduction process also re-
in a measurable increase of the lithium

concentration of the molten metal phase. Accord-
ingly, further study of the reaction
2LiF + Be® = 2Li% + BeF (11)

in the two-phase extraction system was initiated
by experimental
employed for the

those
The
intention of these experiments was to measure
the activity coefficient of lithium in bismuth by
bringing it to equilibrium with metallic beryllium.
that the stoichiometric
amount of lithium did not appear in the metal
phase.

to
rare~earth extractions.

procedures similar

It was found, however,

In experiments where lithium metal was
added to the system, the lithium loss was pro-
portional to the square of the mole fraction of
lithium in bismuth. Such behavior suggests the
presence of a reduced divalent species at less
than unit activity, for which the most obvious
choice is Be? dissolved in the melt. When Be®
was added to the melt, the lithium loss was pro-
portional to the first power of XLi(Bi)’ which
is consistent with the formation of neither Be%(d)
nor Be +(d). For a third set of experiments, where
salt was added to bismuth containing lithium,
the loss was independent of the lithium concen-
tration, indicating the presence of some easily
reduced impurity in the melt. No simple mecha-
nism bas been devised to explain all three of
these reactions.

It is known that certain properties of this melt
[e.g., heats of solution of HF and solubilities of
PuF , and (RE)FS] show extreme values at the
ratio 2Li:Be. The salt composition used in the
described experiments started at about this con-
centration and went to opposite sides of it. One
can conceive, therefore, that further experimenta-
tion may reveal solvent effects which are as yet
unexplained.

SReactor Chem. Div. Ann. Progr. Rept. Dec. 31,
1965, ORNL~-3913, p. 40.

36

In the beryllium-addition experiments at 600°C
a limiting mole fraction of lithium in bismuth was
reached. As this was well below the solubility,
it was assumed that the lithium was in equilibrium
with metallic beryllium. From this it was possible
to calculate an activity coefficient for lithium of
9.8 x 107° to 1.3 x 10~ * (two experiments) for
the mole fraction of lithium referred to a standard
state of unit activity (i.e., pure lithium). A sim-
ilar analysis gave the activity coefficient of lith-
ium in lead as 1.5 x 10>,

REMOVAL OF PROTACTINIUM FROM MOLTEN
FLUORIDES BY REDUCTION PROCESSES

J. H. Shaffer W. P. Teichert
D. M. Moulton F. F. Blankenship
W. K. R. Finnell W. R. Grimes

The removal of protactinium from solution in
LiF-BeF -ThF, (73-2-25 mole %) has been
demonstrated by adding thorium metal that was
either put directly in the salt mixture or initially
dissolved in molten lead or bismuth that was in
contact with the salt.® More recent studies have
examined methods by which this reduction re-
action might be used for reprocessing the fertile
blanket of a two-region molten-salt breeder re-
actor. The results of several batch-type laboratory
experiments led to the design and operation of
a small pump-loop experiment which has demon-
strated, in principle, the removal of protactinium
from the fluoride mixture by a liquid-liquid ex-
traction technigue.

In static batch-type experiments,
fractions of 233Pa
on adding thorium
components of the
of the
used in these experiments indicated that most of
the precipitated protactiniuin had deposited on the
vessel walls that were in contact with the salt
phase. Although this behavior may have resulted
from nonwetting characterigtics of the two liquid
phases, an experiment was conducted to examine

only minor
remmoved from the salt phase,
metal, were found as soluble
metal phase. Subsequent ex-

aminations low-carbon-stcel containers

the absorption of ?°%Pa on iron surfaces in the
absence of a molten metal phase. As a result of
adding thorium metal, ?3°Pa was found uniformly

l—)Reactor Chem. Div. Ann.
1965, ORNL-3913, p. 42.

Progr. Rept. Dec. 31,
distributed on steel wool that had been immersed
in a blanket salt mixture. When this salt mixture
was drained from the vessel and filtered through
sintered nickel, essentially no ?3°Pa activity
could be found in the salt mixture or on the filter.

In other experiments in which no salt phase was
used, solutions of ?3°Pa in molten lead or bis-
muth, obtained by the addition of irradiated
thorium metal, were not stable in either metal
solvent.  However, a much larger fraction of
233pa activity was retained in bismuth than in
lead during 48-hr contact periods. Subsequent
examination of the low-carbon-steel vessels used
in these experiments showed a distribution of
233pa on the container walls which resembled
sedimentary deposition of insoluble materials
rather than surface absorption. In view of earlier
results, tentative conclusions assumed that 233Pa
was preferentially absorbed on insoluble par-
ticles that were initially present in the molten
metals or formed by reactions with added thorium.

Since the anticipated function of the molten
metal phase in the extraction process is that of
an intermediate carrier for protactinium, the rate
at which ?33Pa can be extracted from the blanket
salt and concentrated in a second salt mixture by
back extraction with HF need only depend on the
mass transfer rate of protactinium dissolved in a
recirculating molten metal stream. Thus a pump-
loop experiment, shown schematically in Fig. 3.6,
was tried in an endeavor to achieve the transport
of 233pa in bismuth while maintaining its con-
centration or that of its carrier at relatively low
values, Thorium was introduced into the system
by contacting the liquid metal with thorium chips
just prior to its reentry into the extraction vessel.
At low bismuth flow rates protactinium could be
reduced at the surfaces of free-falling droplets.
For the recirculating molten metal
stream was pumped through a bed of steel wool
to provide for the collection of protactinium, pre-
sumably by absorption, and to provide coarse fil~
tration of the bismuth in the event that **3Pa
was being carried by suspended solid particles.
Sutrface areas of steel wool columns used in the
experiment were at least tenfold greater than
those of other iron surfaces exposed to the molten
metal elsewhere in the loop. The extraction
vessel was also provided with a niobium sleeve
to isolate the salt phase from the iron surfaces
of the loop.

simplicity

37

The pump-loop experiment was operated dis-
continuously for approximately 60 hr over a period
of about six weeks and was terminated because
of pump failure. Material balance calculations on
the system at the conclusion of the experiment
showed that approximately 96% of the 2**Pa had
been removed from the salt mixture. At least
43% of the 2?33Pa originally in the system had
been pumped as a solution or a suspension with
molten bismuth and deposited on rather small
volumes of steel wool. Since only 4% of the
233pa remained in solution in the bismuth, ap-
proximately 49% of the #33°Pa was lost as solids
in the system. The collection of 2%7Pa on the
columns was, in fact, better identified with a
filtration process even though some surface ab-
sorption was apparent. Spectrographic analyses
of high-melting metallic plugs taken from the
system associated relatively high concentrations
of thorium with iron and chromium. These ob-
servations suggest that a more inert containment
material will be needed before a satisfactory

ORNL -DWG 66 -11484

    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

HELIUM
SUPPLY
EXHAUST
V-2
....... E
VARIABLE Wl H O
SPEED = . v e
MoToR 1! EXAOST 5
O - vos BT al| O
g
g q -
0
T] fl( ?’%
THORIUM 2
ADDITION-T & o
H = He o4 =
L q o8 =
[ ] Yx i M=
= W B
h ul # V-7 :-,LIJZ A
I s 1
~__H g5 14
F L =0 AT M9
/ v-g EZO 4 -
H 11—2 4 =
J - o
- : M5
f z Q
3 o
] Ha 3 fi i &
M / ;
azzzfis/z (i
M—
Fig. 3.6, Schematic Diagram of 233p, Extraction
Pump Loop.
38

demonstration of the liquid-liquid extraction process
can be achieved.

REMOVAL OF PROTACTINIUM FROM MOLTEN
FLUORIDES BY OXIDE PRECIPITATION

J. H. Shaffer W. K. R. Finnell
W. P. Teichert F. F. Blankenship
W. K. Grimes

In a previous experiment protactininm was re-
moved from solution in a solvent mixture of Lil-
BeF (66-34 mole %) which also contained ZrF
(0.5 mole per kg of salt) by the addition of Z:0,
at 600°C.7
data according to the equation

An interpretation of the experimental

! 1 b W (12)
—_— — 4 — W
r Zro ’ )
Fpa salt 2
where D = (Pa>oxide/(Pa)salt’ Fo.~= fraction of

protactinium in the salt, and W = weight of the
designated phase, showed that the distribution
of protactinium between the two phases remained
over the
range of the experiment.

constant protactinium concentration

These results could be
explained as the formation of labile oxide solid

25

® SURFACE AREA ZrO, == 80 m?/g
© SURFACE AREA ZrO, =50 m%jy

4 SURFACE AREA Zr0,==1.32 mZ/g
20 ¢

 

WEIGHT OF SALT MIXTURE: 3.55 kg
W

.(_ . ".Zr’OZ )

Fro Wealt

WHERE Fpo'—g—FR‘f\CflON OF Po IN L

W= WEIGHT OF DESIGNATED PHASE
— CONC OF Pg IN SQLID PHASE
CONC OF Pa IN LIGUID PHASE

4+D(
233

 

TION OF 33P0 IN SO_UT:ON

r

 

RECIPROCAL FRA!

o

3 B -
.,.,«M:mf%—’/

et A T

10 20 30

-~
e

s T

solutions or as surface absarption of ?°3Pa on the
solid ZrO,. Further studies of this oxide pre-
cipitation method were conducted in the same
fluoride solvent with ZrQ, powders having varied
surface areas.

Zirconium dioxide used in the original experi-
ment was purchased commercially and had a sur-
face area of about 19.6 m?/g. Material having
higher surface areas was prepared from Zr(OH)4
by 8 Sufficient .’I/IrO2 for this ex-
perimental series was fired at 600, 700, and
1000°C in separate batches that yielded average
surface areas of 80, 50, and 1.32 m?/g respec-
tively. About 3.55 kg of a salt mixture having a
nominal composition of LiF-BeF -ZrF  (64.8-
33.6-1.6 mole %) with about 1 mc of 233Pa as
irradiated ThO, was prepared in nickel by con-
ventional HF-H, treatment at 600°C and H,
sparging at 700°C for further purification and dis-
solution of protactinium as its fluoride salt. Se-
lected ZrO, was added to the salt mixture in

dehydration.

10-g increments; the mixture was then sparged

 

"Reactor Chem. Div. Anm.
19635, ORNI.-3913, p. 41.

SZr()2 was prepared by H. H. Stone, Reactor Chem-
istry Division.

 

Progr. Rept. Dec. 31,

ORNL -DWG 66-14462

 

IQUID

 
  

40

Zr0, ADDED {g)

Fig. 1.7, Effect of Surface Area of ZI‘OZ on the Remova
600°C.

| of 233Pg from LiF-BeF,-ZrF , (64.:8-33.6-1.6 Mole %) at
with helium at a rate of about 1 liter/min during
24-hr equilibration periods. Filtered samples of
the salt mixture were taken after each equilibra~
tion period and analyzed radiochemically for
233pa by counting its 310-kv photopeak on a
single-channel gamma spectrometer. At the con-
clusion of the experiment the mixture was hydro-
fluotinated to convert added ZrO_ to its fluoride
salt and to restore 233Pa activity in the molten-
salt phase. This experimental procedure was
repeated with the same salt mixture for all three
lots of Zr0,,.

In each experiment the addition of ZrO, to the
fluoride mixture resulted in the loss of protace
tinium from solution. However, as shown by Fig.
3.7, a plot of the reciprocal fraction of 233Pa
in solution vs Zr0O_ added, yielded, according to
Eq. (12), distribution coefficients for ??3Pa be-
tween the two phases which varied continuously
as the precipitation reaction approached comple-
tion.  Although these experimental results are
contrary to those obtained previously, with re-
spect to the constancy of the 233Pa distribution
coefficients, they indicate that 2?33Pa removal
from the salt mixture is probably not singularly
dependent on the surface area of the added oxide
particles nor on solid solution formation.

PROTACTINIUM STUDIES IN THE HIGH-ALPHA
MOLTEN-SALT LABORATORY

C.J. Barton  H. H. Stone

The High-Alpha Molten-Salt Laboratory was
briefly described in the previous report,? and
results of the first experiments performed in this
facility were given. Attention has been focused
on development of methods of removing protac-
tinium at realistic concentrations (25 ppm) from
breeder blankets, but a few experiments have been
performed in an effort to obtain a better under-
standing of the chemistry of protactinium in
molten fluoride systems. Because of the variety
of experimental methods that have been applied
to the protactinium removal problem, only a brief
summary is presented here, with emphasis on the
experiments that gave the most promising resulis.
Some information from these experiments has been

 

previously reported, 10— 12
c. Barton, Reacfor Chem. Div. Ann. Progr.
Rept. Dec. 31, 1965, ORNL=3013, p. 44.

Protactinium Recovery Experiments

We found that protactinium dissolved to the ex-
tent of 20 to 30 ppm in molten LiF-ThF, (73-27
mole %) could be readily reduced by solid thorium
or by thorium dissolved in lead. In the latter
case, only a small fraction of the reduced protace
tinium was found in the molten metal phase. Re-
duction experiments with solid thorium in three
different container materials (nickel, copper, and
graphite) showed that more than half the reduced
protactinium temained suspended in the molten
fluoride mixture. We believe that the reduced prot-
actinium is attached to small particles of a struc-
tural metal such as iron or nickel which are large
enough to be removed by the sintered copper filter
material through which the samples are drawn
but small enough
reasonable
molten salt,

Partial reduction of protactinium was effected by
electrolysis with various electrode arrangements,
but wvery little protactinium was found in the
bismuth layer that underlay the molten-salt mixture
in most of the electrolysis experiments.
of these experiments,
from the fluoride mixture to bismuth or to some
other electrode material that could be readily
separated from the salt mixiure, was not realized.

Efforts to collect tracer quantities of reduced
233pa on steel wool have been reported.!? A
series of three experiments of this type were re-
cently performed with ?31Pa concentrations in
an LiF-ThF, (73-27 mole %) mixture in the range
24 to 81 ppm.
ratio of milligrams of 23'Pa to prams of steel
wool. These ratios were 1.1, 3.1, and 6.5 for the
three experiments. Detailed results are given
only for one experiment (**'Pa to Fe ratio 6.5),
but conclusions are based on findings of all
three experiments, which gave similar results.

A weighed quantity of I_.iF-ThF4, previously
purified, was placed in a welded nickel reaction

to remain suspended for a
length of time in the high-density

The aim
to transier protactinium

The principal variable was the

 

106, J. Barton and H. H. Stone, Removal of Protac-
tinium from Molten Fluoride Breeder Blanket Mixtures,
ORNL~TM-1543 (June 1, 1966).

He, J. Barion, MSR Program Semiann. Progr. Rept.
Feb. 28, 1966, ORNI1.-3936, pp. 148--52.

12¢, J. Barton, MSR Program Semiann. Progr. Rept.
Aug. 31, 1966, ORNL-4037, pp. 156-58.

13}. H. Shaffer et al., MSR Program Semiann. Progr.
Rept. Aug. 31, 1966, ORNL-4037, pp. 148--56.
vessel, irradiated ThF4 containing a known
amount of 2%*Pa and 231Pa was added to the
mixture, and it was treated first with a mixture

of HF and H, and then with H, Four
grams of steel wool (grade 00, 0.068 m?/g sur-
face area) was placed in a low-carbon-steel liner

alone.

inside another nickel vessel. 'The contents of
the vessel were then treated with purified hy-
drogen at 800°C for several hours to remove as
much as possible of the oxide surface contami-
nation of the steel wool and The two
vessels were then connected together at room
temperature and heated to about 650°C, and the
salt was transferred to the steel-lined vessel.
After two separate exposures of the salt to a
solid Table
3.1, the salt was transferred back to its original
The
steel-lined vessel was cut up, and samples were
submitted for analysis,

liner.

thorium surface, as indicated in

container and allowed to cool in helium.

40

data in ‘T'able 3.1 show that 99% of the
was precipitated in a {form that

The
protactinium
would not pass through a sintered copper filtet
after a fairly short exposure to solid thorium, but
nearly 7% was in the unfiltered salt that was
transferred back to the nickel vessel after ex-
posure to thorium. About 62 g of salt was associ-
ated with the steel wool in the steel liner in the
form of a hard ball. Partial separation of the salt
from steel wool was effected by use of a magnet
after crushing the ball, and the iromn-rich fraction
had the higher protactinium conceantration. The
sinall amount of protactinium found on the vessel
wall is especially notable. The last column in
Table 3.1 shows that a reduction in the iron con-
centration occurred concurrently with the reduction
in protactinium concentration. It may be signifi-
cant that the ratio of precipitated iron to precipi-
tated protactiniuin was in this

miuch smaller

experiment than in the other two experiments,

Table 3.1. Precipitation of Protactinium from Molten LiF-ThFd (73-27 Mole %) by Thorium Reduction

in the Presence of Steel Woo!

 

 

 

PPa Total Total
Sample Concentration 23 pg, Iron
(mg/g) (mg) (mg)
Salt after HF-H, treatment 0.0634 20.3 <15
Salt just before transfer (.081 26.1 116
Salt 35 min after transfer 0.079 24.9 85
Salt after 50 min thorium exposure 0.0026 0.69 22
Salt after 45 min thorium exposure 0.0009 0.27 18
Nonmagnetic fraction of material 0.20 11.5 994
in steel liner
Magnetic fraction of material in 0.628 10.2 2750
steel liner
Unfiltered salt after transfer to 0.0076 1.75 <15
nickel vessel
Steel liner wall 0.0006
Stainless steel dip leg 0.53
Filings from thorium rod 0.29
All salt samples 1.35
Total protactinium recovered 25.5
where the retention of protactinium by the steel
wool was more efficient. Coprecipitation of
metallic protactinium iron (and possibly
nickel) would help to account for the manner in
which protactinium settled out on, and adhered to,
the steel wool surface.

On the basis of presently available information,
thorium reduction of protactinium from molten
breeder blanket mixtures in the presence of steel
is believed to be a promising recovery
method warranting further investigation.

and

wool

GRAPHITE-MOLTEN-SALT IRRADIATION
TO HIGH FiSSION DOSE

H. C. Savage E. L. Compere
J. M. Baker M. J. Kelly
E. G. Bohlmann

Irradiation of the first molten-sall convection
loop experiment in ORR beam hole HN-1 was
terminated on August 8, 1966, after development
of 1.1 x 10'® fissions/cm?® (0.27% 235U burnup)
in the ‘LiF-BeF -ZiF ~UF  (65.16-28.57-4.90-
1.36 mole %) fuel. Average fuel power densities
up to 105 w per cubic centimeter of salt were
attained in the fuel channels of the core of MSRE-
grade graphite.

Successful operation of the major heating,
cooling, temperature-control, and sampling sys-
tems was demonstrated; however, leaks developed
in two of the four cooling systems. The experi-
ment was terminated after radioactivity, resulting
from fuel leakage from a break in the sample line
near the loop, was detected in the secondary
containment.

Irradiation of a second loop, modified to elimi-
nate causes of failures encountered in the first,
will begin in January 1967. Operation at an
average core fuel power density of 200 w/cm?
for a period of the order of a year will be sought.

Objectives and Description

The loop is designed to irradiate a representa-
tive molten-salt fuel circulating at typical tem-
perature differences in contact with graphite and
Hastelloy N at desired core power densities of
200 w/cm?, with provisions for gas removal and
salt sampling. In particular, it is desired to

41

study the interaction of fission products with
graphite, metal, fuel, and gas phases and the
stability of the fuel salt at high levels of burn-
up. L1417

The core of the first loop consisted of a 2«in.-
diam by 6-in.-long cylinder of graphite obtained
from MSRE stock. Through the core, eight vertical
1/4—ir1. holes for salt flow were bored, arranged
octagonally with centers S/8 in. from the graphite
center line, A horizontal gas separation tank
connected the top of the core to a return line to
the core bottom, completing the loop. The tank,
lines, and the core shell were fabricated of Has-
telloy N. The heaters and the cooling tubes in the
core and return line were embedded in sprayed-on
nickel, as was the 12-ft sample tube leading from
the loop to the sample station in the external
equipment chamber.

Operations. — The loop was operated with MSRE
solvent salt for 187 hr at Y-12, and several salt
samples were taken. [t was inserted in beam
hole HN-1 of the ORR on June 9. 1966, and
operated 1100 hr with solvent salt; during this
period calibration and testing of equipment and
performance were conducted. The loop was in-
serted to the position nearest the reactor lattice
on July 21, and water injection into the air
streams to the tubular core coolers and the jacket
around the gas separation tank was tested. One
of the two core coolers leaked and was plugged
off. Water injection was discontinued until after
uranium addition.

On July 27, after sampling, eutectic 7LiF-‘»UF4
(93% enriched) fuel salt was added to develop a
uranium inventory concentration of 1.36 mole %.
At this time a capillary tube in the sample re-
moval system broke, precluding further sampling.
An associated inleakage of air impelled solvent
salt to a cold spot in the gas sample line, thereby
plugging it.

During subsequent operation fission heat was
determined.  During this period water was re-
leased into the loop container from what proved

 

145K Program Semiann. Progr. Rept. Aug. 31, 1965,
ORNL~3872, pp. 106--10.

ISMSR Program Semiann. Progr. Rept. Feb, 28, 1964,
ORN1.-3936, pp. 152--54,

16 peactor Chem. Div. Ann. Progr. Rept. Deec. 31,
1965, ORNL-3913, pp. 34-35.

17 Reactor Chem. Div. Ann. Progr. Rept. Jan. 31,

1965, ORNL-3789, pp. 4548, Fig. 2.4.
to be a leak in the cooling jacket around the gas
After a short reacter shutdown
on August 8 to permit removal of the water accu=
mulated as a result of the leak, the irradiation
was release of sub-
stantial

separation tank.

That evening,
into the loop container

The loop temperature
was lowered to freeze the salt, and the loop
was retracted to 2% flux. [t was removed to the
hot cells for disassembly and examination on
August 11, 1966.

Chemicol Analysis of Salt. — Samples of solvent
salt taken prior to irradiation and after 1100 hr
in pile and of irradiated fuel salt obtained after
dismantling were analyzed chemically and radio-
chemically. A sample of salt found between the
metal core shell and the graphite was also
analyzed. Results are given in Tabhle 3.2 and
are discussed below.

resumed.
radioactivity
indicated fuel leakage.

Corrasion. — The level of corrosion products,
particularly chromium and nickel, in the salt
increased in the successive samples. This was

possibly due to uptake of moisture by the solvent
salt prior to loading, with consequent corrosion
of the Hastelloy N. This appears to have occurred
in the addition tank, since a sample taken directly
from the addition tank without entering the loop
showed similar levels of corrosion products.

Producis. - Fission products were
counted in a fuel sample after 110 days’ cooling;
concentrations are given below as a percentage
of the amount produced, calculated on the basis
of observed fission heat (4.8 x 10!7 fissions/g).

Cerium-144 and -141 (77, 64%) and zirconium-89
(65%) were somewhat below the calculated pro-
duction. Cesium-137 (41%) and strontium-89 (42%),
with noble-gas precursors of ~3 min half-life,
could have thereby been lost to the gas space
or graphite voids. Tellurium-127 (10%) was
largely removed from the salt. Ruthenium-103
and -106, which were expected to deposit on
Hastelloy N surfaces, were not detected (<0.03%)
in the salt.

Nuclear Heat and Meuvtron Flux, — Nuclear heat
was measured at various loop insertion positions
by comparing electrical heat requirements under
similar conditions with the reactor at zero and
full power. Reactor gamma heat fully inserted
was 2900 w (with unfueled salt). With fuel con-
% wuranium (93% enriched),
fission heat in the fully inserted position was

S800 w.

Fission

taining 1.306 mole

The coiresponding overall average fis-

42

sion heat density was 80 w per cubic centimeter
of salt at 650°C, and in the graphite core the
density
cubic centimeter of fuel salt.

average fission heat was 105 w per

The overall effective thermal-neutron flux in the
salt was estimated independently from nuclear
heat, from activation of solvent salt zirconium,
from cobalt monitors in the loop exterior, and by
neutron transport calculation. The results agreed
well, ranging between 0.9 and 1.2 x 10'® neutrons

2 gec™ 1,

cin

HoteCall Examination of Components. - After
separation from other parts of the package, the
loop proper was kept for about three months in
a furnace at 300°C to prevent fluorine evolution
by fission product radiolysis of the salt. At this
it was removed for detailed examination.

The type 304 stainless steel tubular core cooler
was found to have broken entirely loose without
ductility at its outlet end as it left the core near
a tack weld to the corc shell. Intergranular cracks
originated on the outer circumference of the coiled
tube.

The cooling jacket on the gas separation tank
leaked at a weld. The fuel leak resulted from
a nonductile break in the Hastelloy N sample line
tubing near the attachment to the core bottom.
The sprayed nickel was also cracked in this
regioii.

Fuel salt in the form of a scale a few mils
thick was found on the interior of the core shell,

time

between it and the closely fitting graphite core.
The analysis shown in Table 3.2 appeaws to be a
mixture of fuel salt and Hastelloy N (probably
metal debris from cutup operation).

Hot-cell metallurgical examination of the in-
terior surfaces of the Hastelloy N comprising the
core bottom and core shell wall showed no evi-
dence of any interaction with salt or carbon, or
other change.

Evaluation of System Performance

Heaters. — The molten-salt loop package used
21 continuous or intermittent heaters, all 1/g-in.—
OD, Iaconel-sheathed, MgO-insulated, with Ni-
chrome V elements designed for 870° continuous
operation.

Coolers. — The heat removal rate of the loop
coolers

No failures cccurred,

was entirely adequate to remove the

8.8 kw of fission and gamma heat, even after
Table 3.2.

Loading and Samples from in-Pile Molten-Salt Convection L.oop

Analyvses as mg/g or mole %

 

 

Li Be Zr U F Cr Fe Ni Mo
Composition as Loaded
Solvent salt
Composition as manufactured, mole %b (64.78) {30.06) (5.16)
me/g 114,58 68.4 118.9 698.2
Average production analysis, mg/g 108.6 72.5 119.5 699.4 ~0.020 ~0.020 ~0.100
Fuel, eutectic, mole % (72.48) (27.54)
mg/g 48.5 619.6 331.9
Fueled loop mixture {calcd), mole % {65.16) {(28.57) {4.90) {1.36)
mg/g 106.5 62 104.5 74.8 654.0
Hastelloy N, representative analysis,® mg/g 70.400 46.000 696,200 161.000
Aralyses of Loop Samples, mg/g
Hours Radiation Sample
Molten Hours No.
120 0 1 (solvent salt) 98.0 66.5 119.5 683 ¢.310 0.275 0.137 No data
166 0 3 (solvent salt) 897.5 56.1 122.0 673 0.3558 0.285 0.455 No data
1260 208 6 (solvent salt) 116.5 69.1 120,0 704 0.6709 0.09z2 0.540 No data
After shutdown
1578 329 9 (fueled mixture) 113.0 58.5 99.4 71.5 0.780 0.258 0.555 <0.015
Graphite-Inor annulus specimen S»1 105.0 56.1 101.5 71.7 3.250 3.8060 25.300 4.740

 

92,067 29%y,

bj. H. Shaffer, MSR Program Semiannual Progr. Rept. Feb. 28, 1965, ORNL-3%12, pp. 15052,

“Heat SP=19 for comparison.

SIC:' activation gave a chromium concentration of 0.990 mg per gram of salt.

v
the loss described earlier of one of the two
cooling coils around the loop core section. The
air plus water-injection technique appears ade-
quate and responsive. The use of water injection
was not necessarily the cause of failure of the
two cooling units, but only made the failures
evident.

Temperature Control. — The response of the
heating and cooling systems to rapid changes in
the nuclear heat could only be tested under full
fission conditions in pile. Since this was re-
garded as important, reactor setback tests were
conducted. Temperature-control system response
was adequate to maintain the salt molten during
a reactor setback with resultant loss of 8.8 kw
of nuclear heat, and to return the loop to normal
operating condition during a rapid (11-min) return
to full power.

Sampling ond Addition. -- Sampling and addition
systems and procedures were adequate to permit
addition and removal of molten salt while operating
the loop in pile and to transport shielded samples
under an inert-gas atmosphere to the analytical
laboratory. A broken capillary connecting tube
prevented additional sampling.

Salt Circulation. — Convective salt circulation,
at rates of 5 to 10 cm?®/min, was achieved by
causing the return line to operate at temperatures
below the core temperature. Flow stoppages
These were attributed
to bubble formation resulting from different solu-

occurred from time to time,

bility of argon cover gas at the varied tempera-
tures around the loop. Salt flow was reestablished

by evacuation and readdition of cover gas. Loss
of flow had no adverse effect on loop operation.
Second In-Fiie lrradiation Assembly. — A second

in-pile molten-salt convection loop, essentially
identical to the first convection loop experi-
ment,!® has been constructed, and it is antici-
pated that in-pile irradiation will begin early in
1967. Problems encountered
vection loop experiment and subsequent post-
irradiation hot-cell examination, described above,
have led to modifications to the second loop
which are designed to eliminate these problems.

The coolant tubes, embedded in nickel spray
Y=in.~00) x
0.035-in,~wall Incone!l tubing instead of the z-m.-

OD x 0.035-in.~wall 304 stainless steel used on

in the first con-

around the core section, are now of

44

the first loop. The stainless steel tubing should
have been entirely adequate for the service, but
Inconel is the preferred material for exposure to
the high-temperature steam (™~400°C) genecrated
when air-water mixtures are used as coolant.
Since the rupture of one of the core coolant
tubes occurred adjacent to a point where the
tube was tack welded to the core wall, the tack
weld was eliminated in favor of a mechanical
strap attachment. An expansion loop to relieve
stresses has been included in each of the coolant
outlet lines. A mockup of the modified cooling
coil was operated at temperature with air-water
mixtures for more thaa 400 hr, including 120
thermal shock cycles (600 - 350°C), with no
sign of difficulty. Thermal! cycling occurs during
a reactor setback and startup, and it is estimated
that no more than about 20 such thermal cycles
will occur during a vear of operation.

The two failures which occurred in the capillary
tubing (0.100 in. OD x 0.050 in. ID) used in the
salt transfer system appear to have resulted from
excessive mechanical stress. Consequently, the
wall thickness of this line has been increased to
0.050 in., and additional mechanical support has
been added such that there is now no part of the
salt sample line which is unsuppoited —~ as was
the case in the first loop assembly.

The Y -in.-thick stainless steel cooling jacket
surrounding the reservoir tank has been replaced
by an Incone! tube wrapped around the outside
of the tank and attached by means of sprayed-on
nickel metal, as is done on the core section and
cold leg. Also, provisions for use of an air-water
mixture as coolant have been added, since it was
found that air alone did not provide sufficient
cooling in the first experiment.

Continuous salt circulation by thermal cone-
vection was not maintained in the first experi-
ment. It was concluded that loss of circulation
was caused by gas accumulation in the top of
the core section. Accordingly, the salt flow
channels at the top and bottom of the eight ' -in.
holes for salt tflow in the graphite core were re-
designed to provide better flow conditions!’ at
the inlets and exits of the vertical holes. [Fur-
ther, the top and bottom of the core section,
horizontally oriented on the first loop, were in-
clined at 5°to minimize trapping of gas.
4. Direct Support for MSRE

EXTENT OF UF ; REDUCTION DURING
MSRE FUEL PREPARATION

B. F. Hitch C. F. Baes, Jr.
Uranium was added to the barren fuel salt of
the MSRE as a binary mixture of 27 mole % UF,
in ’LiF. ‘This fuel concentrate had first been
purified by the usual sparging with an HF-H,
mixture to remove oxide, followed by sparging with
hydrogen alone to complete the reduction of struc-
tural metal fluorides such as NiF, and Fef . !'?
During this final reduction step, a small portion
of the UF, should also have been reduced, the
amount depending upon the duration of the treat-
ment and the equilibrium constant for the reaction

UF (d) + LH, (g) = UF (d)+ HF(g).

The exact amount of UF | thus introduced into the
MSRE fuel has become a matter of special interest
with continued operation of the MSRE, owing to
evidence that significant amounts of some fis-
sion products are far more oxidized (see following
section) than would seem compatible with the
presence of significant amounts of UF, in the
MSRE fuel. Consequently, the data collected by
Shaffer et al.® during the purification of the fuel
salt concentrate at the production facility recently
have been examined in detail in an attempt to de-
termine the equilibrium quotient for the above re-
action,

P_ X

HF UJF
Q- =2, (1)

PL/2 X

}{2 UF4

 

1]. H. Shaffer ef al., Reactor Chem. Div. Ann. Progr.
Rept, fan. 31, 1965, ORNL-3789, pp. 99-109.

2,]. H. Shaffer, MSR Program Semiann. Progr. Rept.
July 31, 1964, ORNL-3708, pp. 288~303.

3Unpublished data, supplied by J. H. Shaffer.

45

and to determine the extent of UF_ reduction in
the LiF-UF , mixture.

For small amounts of reduction, the UF,/UF,
ratio may be related to Q and the volume, V, of
H, passed per mole of UF ,(n,) by *

(HU%/HU)2 = 2@1:';1/22 (V/n, RT) + (nng/nU)z,
(2)

provided equilibrium conditions are maintained
duting sparging. The last term on the right is

 

the initial nUFS/nU ratio. Replacing n FE’/nU
p1/2
by QPHZ /P
1 2 V 1
P2 ’:Qpl/z(' RT> +(_P'6—””z“' )
HF u, \Mu HF

In accord with this equation, plots of 1/PZ _ vs
V, based on data collected at 700°C during the
purification of the various batches of fuel con-
centrate, were found to be linear. All plots could
be fitted reasonably well with lines of slopes
corresponding to Q ~ 0.9 x 10-% atm!/2. By
measuring V from the intercept of each plot at

 

4Cmnbination of

 

 

dn . = qv
3 RT
and
n
UF, Pup
Q= —
n.. — Pl/2
U~ flyp, tu

3 2

to eliminate P followed by integration gives

 

HF’
v -1
=——~175[r +la(l -0+,
e, RT QPH2
where I = nUFs/nU. For small values of r this sim-

plified to Eq. (2) of this repoxt.
1/‘0311? = 0, the average amount of uranium re-
duction at the end of the hydrogen treatment was
estimated to be 0.16%.

In an attempt to confirm this estimate of @ and
the amount of reduced uranium present initially
in the MSRE fuel, an 11.4-kg portion of unused
fuel concentrate was studied further in the lab-
oratory. Hydrogen sparging was initiated at 510°C.
At this relatively low temperature, no sigpnificant
reduction of U*" to U3% should occur; however,
HF evolution was detected immediately and con-
tinued at a significant leve! until 250 liters of H,
had been passed and 0.0019 mole of HF per mole
of uranium had been evolved. This indicated that
inadvertent exposure of the salt to oxidizing im-
purities such as water or oxygen had occurred
during prior storage, during transfer of the sam-
ple to the reaction vessel, or in later handling.
Since the HF at this temperature in the amounts
seen should have quickly oxidized the UF ; pres-
ent, it was not possible to confirm the amount
of UF, initially present in the fuel concentrate.
In two subsequent H, sparging runs at 700°C,

however, data were obtained which permitted im-
2

proved estimates of Q from plots of 1/ vs V:
Temperaturs Hy Flow Q
o | -1 1/2
("C) (! min kg™ ) {atm )
Run 1 707 53 1.74% 107°
Run 2 705 35 1.85% 107°

The resulting values of Q are about twice those
estimated from the salt production data. It is
not reasonable to attribute this discrepancy en-
tirely to the differences in temperature, since,
judging from Long’s measurements of the tempera-
ture dependence of Q in LiF-BeF, melts,® more
than a 30°C difference would be required. It seems
more likely that the discrepancy is due partly to
nonequilibrinin sparging conditions in the produc-
tion treatment. The present value of Q = 1.8 x 10~¢
atm !/ 2 determined for the fuel concentrate is some-
what lower than the value ~4 x 107% atm!/?
which may be estimated for the MSRE fuel salt
at 700°C from Long’s measurements. This in-
dicates that UF_ is not as easily reduced in the
fuel concentrate as in the fuel salt.

Even though equilibrium conditions might not
have prevailed during purification of the fue! con-

 

G, l.ong., Reactor Chem. Div. Ann.

Jan. 31, 1965, ORNL-3789, pp. 68~72.

Progr. Rept,.

46

centrate, 0.16% reduction of UF , remains a valid
estimate, since, in effect, it is based upon the
integrated amount of HF evolved by reduction,
which, in turn, is related by material balance to
the amount of UF ; formed.

CHEMICAL BEHAVIOR OF FLUORIDES
DURING MSRE OPERATION

R. E. Thoma

The Molten-Salt Reactor Experiment operated
during six separate periods in 1966; virtually all
of the operating time accumulated after mid-May
was at the maximum possible power of about 7.5
Mw. The reactor accumulated approximately 11,200
Mwirr during the year.

During periods of reactor operation, samples of
the reactor salts were removed routinely and were
analyzed for major constituents, corrosion piroducts,
and (less frequently) oxide contamination. Stand-
ard samples of fuel are drawa three times per
week: the LiF-Bel”
two weeks.

, coolant salt is sampled every

Current chemical analyses suggest no percep-
tible composition changes for the salts since they
were first introduced into the reactor some 20
months ago. ©

While analyses for ZrF , and for UF, agree quite
well with the material balance on quantities charged
to the reactor tanks, "LiF and
Bel ) have never done so; analyses for LiF have
shown higher and for BeF, have shown lower
values than the book value since startup. Table 4.1

the values for

shows a comparison of current analysis with the
original inventory value. While the discrepancy
in LiF and BeF , concentration remains a puzzle,
there is nothing in the analysis (or in the be-
havior of the reactor) to suggest that any changes
have occurted. The burnup of uraninm totaling
some 0.3 kg out of 230 kg in the systemn should
be perceptible (and does not seem to be) within
the experimental scatter. A chronological sum-
marv of all MSRE fuel salt analyses is shown in
Fig. 4.1; periods of reactor operation are indicated
by the shaded areas of the figure.

in MSRE fuel is

64 ppm at present; the entire operation seems to

‘The chromium concentration

°r. E. Thoma, MSR Program Semiann. Progr. Rept.
Aug. 31, 1966, ORNL-4037, pp. 134--39.
have increased the chromium concentration only
26 ppm. This increase corresponds to removal
of about 130 g of chromium from the metal of the
fuel circuit. If this were removed uniformly it
would represent removal of chromium to a depth
of about 0.1 mil. Analyses for iron and nickel
in the system are relatively high (120 and 50 ppm
respectively) and do not seem to represent dis-
solved Fe?! and Ni?* While there is consider-

Table 4.1, Current and Original Composition
of MSRE Fuel Mixture

 

Original Value” Current Analysis

Constituent

 

(mole %) (mole %)
TLiF 63.40 £ 0.49 64.88
BeF 30.63 +0.55 20.26
ZiF 5.14 +0.12 5.04
UF, 0.821 1 0.008 0.82

 

¥From amounts of materials charged to system.

 

!
Qo -
LITHIUM a :
lwt %} . [
1000
TGO
BEGYL) UM
530
fw %)
500
1506 1
ZIRCONIUM i
Moo |-
{wt 2e)
10,30
4700 §-
URANIUM
4650 -
{wi%)
4,600 |-
&80
CHROMIUM =
50 1-
(ppm)
ap -
130
IRON 00
(ppm)
PP 30 |
150
MICKE | 190
(npm)
GO
RUN

 

 

 

Fig.- 4.1. Summary of MSRE Fuel Salt Analyses.

47

able scatter in these analyses, there seems to
be no indication of corrosion of the Hastelloy N
by the salt.

The fuel mixture in the MSRE contained (see
preceding section) considerably less UF . than
the quantity intended; 1.5% of the added uranium
was to be as U?’. Furthemmore, the fission proc-
ess should prove oxidizing to UF, in the melt”
(or to chromium in the Hastelloy N). The extent
to which the fission process should prove oxi-
dizing depends on several variables including (1)
the extent to which Kr and Xe are swept from the
reactor, (2) the redox potential of the fuel-metal
system, and (3) the extent to which evolution of
“unstable’ species (such as MoF , or RuF ) oc-
curs through nonequilibrium behavior. It
very likely, however, that fission of about 0.3 kg
of uranium (perhaps aided by a small amount of
inadvertent oxidation within the MSRE) can have
used up most of the UF, added. An attempt to
determine UI"; concentration in the MSRE after
about 11,000 Mwhr of operation (by H,-HF equi-
[ibrium, a method similar to that employed in the
preceding section) showed less than 0.10% of the
uranium to be in the trivalent state.?

Accordingly, the lack of corrosion in the MSRE
seems to be somewhat surprising.
tionalized by the assumption (1) that the Hastelloy
N has been depleted in Cr (and Fe) at the surface
so that Mo and Ni only are under attack, with Cr
(and Fe) reacting only at the slow rate at which
it is furnished to the surface by diffusion, or (2)
that the noble-metal fission products (see section
on Fission Products on Metal and Graphite from
MSRE Corte) are forming an adherent and protective
plate on the reactor metal,

Though neither of the analyses nor the reactor
behavior suggests appreciable corrosion, plans are
under way, and techniques are being studied, for
reducing about 1% of the MSRE UF , to UF ; within
the reactor. Such a reduction (which would surely
take the MSRE fuel to near its intended UF, con-
centration) should remove all apprehension about
possible corrosion and should, we believe, allevi-
ate some of the problems of volatile fission product
fluorides (see subsequent sections).

seems

It can be ra-

 

7 . .
¥. R. Grimes, internal memorandum.

8"Hydroge_n Reduction of MSRE Fuel,’ intralaboratory
correspondence from A. 5. Mever to W. R. Grimes, Jan. 3,
1967.
Routine determinations of oxide (by study of
salt—~H,O-HF equilibria) continue to show low
values (about 50 ppm) for 02—, There is no reason
to believe that contamination of the fuel has been
significant in operations to the preseunt.

MSRE maintenance operations have necessitated
flushing the interior of the drained rcactor circuit
on four occasions. The salt used for this oper-
ation consisted originally of an 7LiF-BeF2 (66.0-
34.0 mole %) mixture. Analysis of this salt before
and after each use shows that 215 ppm of uranium
is added to the flush salt in each flushing oper-
ation, corresponding to the removal of 22.7 kg of
fuel-salt residue (about 0.5% of the charge) from
the reactor circuit.

The MSRE coolant salt has circulated within
Current
apalysis of this salt indicates no corrosion or
leakage in the coolant salt circuit. On one oc-

the reactor for approximately 6400 hr.

 

 

 

casion, coolant salt was inadvertently partially
Table 4.2.
Sample No. FP56-19
Sample date 5-26
Accumulated Mwhr 2800
Operating time, days® 2.5
Fission Yield
Isotope Half-Life ()
lgr 9.67 hr 5.81
925y 2.6 hr 5.3
BIsr 51 days 4.79
41ce 33 days 6.0
143ce 33 br 5.7
Mo 66 hr 6.06
1930 39.7 days 3.0
105Ru 4.45 hr 0.9
1321e 77 hr 4.7
1311 8.05 days 3.1
133 20.8 hr 6.9
1351 6.7 hr 6.1
2390 2.33 days

frozen in the radiator. No damage was sustained
by the radiator either as the salt froze or thawed.
It is believed that the remarkably low volume
change which the coolant salt undergoes in freeze-
thaw cycles (less thaa 5%) is a consequence of
the large free space found in the Li ,Bel", crystal
structure. ®

FiSSION PRODUCTS IN MSRE FUEL

S. S. Kirslis F. . Blankenship

It has been possible to analyze samples of the
MSRE fuel for the 12 fission product isotopes
shown in Table 4.2 and for ?3°Np and the 2.44 x

10* year ?3°Pu produced in the reactor fuel.

9_]. H. Burns and E. K. Gordon, Reactor Chem. Div.
Ann. Progr. Rept. Jan. 31, 1955, ORNL-378%, p. 30.

Fission Products in MSRE ¥Fuel Somples During Full Power Operation

 

FP7-7 FP7-12
6-27 7-13
5100 7200
13.3 11.9

 

(x 1011y (x 1011y (x 1014
1.20 1.16 1.32
1.19 0.97 1.49
0.223 0.266 0.396 ’

0.61 0.688
1.45 1.5 1.32
3.51 0.951 0.315
0.070 0.024 0.071

0.35 0.376
0.421 0.515 0.381
0.42 0.50 0.536
1.31 1.35 1.45
1.50 1.17 1.11

5.35 10.8

 

“Continuous operating time since shutdown of more than 12 hr or since appreciable change in power.

bCalculated as of sampling time.
Typical results obtained for these materials are
shown in the table.

The strontium and cerium isotopes are of special
interest as fission monitors since they have con-
venient half-lives and stable, nonvolatile fluorides
which would be expected to remain almost com-
pletely in the circulating fuel. The concentration
of these monitors, however, is in only fair agree-
ment with calculations based on power level of
the reactor from heat balance; fission power based
on 2'8r is 75% of nominal power, while that based
on '%3Ce shows 88% of normal reactor power.

Molybdenum and ruthenium are typical of a class
of metals expected to deposit, at least in part,
as elements. These analyses of the salt show
that these materials are present in less than the
expected concentration; if calculations of total
yield are based on ?!Sr, about 60% of the “*Mo
and about 30% of the '"%Ru are accounted for in
the salt. It is not possible to decide whether
these isotopes are present as colloidal particles
or are soluble chemical species.

Isotopes of tellurium and iodine are of interest
as xenon precursors and as elements which might
show appreciable volatility from the melt. Only
about 30% of the !%?Te appears in the salt, but
the expected quantities (90 to 100%) of the iodine
isotopes were found in the salt samples.

Analyses of the MSRE fuel samples do not,
therefore, seem surprising except for the low con-
centration of 13?Te. Examination of graphite and
metal samples and, especially, of specimens from
the vapor phase as described in subsequent sec-
tions do show several surprises.

FISSION PRODUCTS IN MSRE EXIT GAS

Equilibrium Pressures of Noble-Metal Fluorides
Under MSRE Conditions

C. F. Baes, ]Jr.

As the following sections of this document de-
scribe in brief, volatile species of Mo, Te, Ru,
and (probably) Nb have been found in the helium
cover gas of the MSRE. In addition, sizable
fractions of these elements appear (presumably
as metal) on the metallic surfaces of the reactor.
Their unexpected behavior prompted a review of
the thermodynamic data on the volatile fluorides
of these elements and an assessment of their

49

equilibrium pressures under
MSRE conditions.

The formation free energies for NbF ., MoF,
and UF , may be calculated with relatively good
accuracy because of recent measurements at Air-
gonne of the heats of formation of these com-
pounds by fluorine bomb calorimetry. 1% The
entropies and heat capacity data also are avail-
able. 13 While the people at Argonne have meas-
ured RuF_,'* no entropy or heat capacity data
seem to be available:

some hypothetical

AHL;DB ASfigs Reference
MoF (#) -372.35 £ 0.22 ~72.13 11
UF () —510.77 £0.45 —~67.01 12
NbF (s) ~433.5 10.15 -91.56 10
RuF ((s) -213.41 £ 0.35 14

From these wvalues and the available heat ca-

pacity data the following expressions for AGE were
derived. In the case of RuF , Glassner’s'’ ear-
lier estimate was corrected to be consistent with
the above AHf measurement:
AG!H (NBF _, g) = —416.70 + 54.40(T/1000) ,
AGH (RuF , g) = —200 + 25 (7/1000) ,
AG! (MoF _, ) = ~370.99 + 69.7(7/1000) ,
AGH(UF,, @ = ~509.94 + 65.15(7/1000) .

The following values of AG! have been reported
previously for UF., and UF, in '.ZLiF-BeI:?‘g:16

AGH(UF ,, d) = ~336.73 + 40.54(7/1000)
AGT(UF ,, d) = —444.61 + 58.13(T/1000) .

From these free-energy values the following equi-
librium constants have been calculated for the

 

IOE. Greenberg, C. A. Natke, and W. N. Hubbard,
J. Phys. Chem. 69, 2089 (1965),

Ly, L. Settle, H. M. Feder, and W. N. Hubbard, J.
Phys. Chem. 65, 1337 (1961).

12y L. Settle, H. M. Feder, and W. N. Hubbard, J.
Phys. Chem. 67, 1892 (1963).

13%. K. Kelly, U.S. Bur. Mines Bull. 584, 1960.

144, A. Porte, E. Greenberg, and W. N. Hubbard, J.
Phys. Chem. 69, 2308 (1265).

134, Glassner, The Thermochemical Properties of
Oxides, Fluorides, and Chlorides ta 25007K, ANL-5750
(1958).

16C. F. Baes, ]Jr., Reactor Chem. Div. Ann. Progr.
Rept, Dec. 31, 1965, ORNL.-3913, p. 22.
formation of the volatile fluorides by reaction with
UF ,(d) in the MSRY from the equation

Tellurium hexafluoride has not been included
in this listing, but this compound seems certain

3 to be less stable than any shown here. No data
log K = a+ B(10°/T) :
Reaction K a b
[ ; e - 5 5
Nbis) + 5UT4(d) = NuFS(g) + SUT3(d) NbF XUFS/XUF4 7.33 —76.82
R s oy 5 5
Ru(s) + SUF ((d) &= RuF (&) + SUF ,(d) PRuFS XUF3/XUF4 13.76 —74.17
—= 6 6
Mo(s) + 6UF ,(d) == MoF (&) + 6UF4(d) Pyor, Xur,/Xur, 7.83 ~60.38
3UF ,(d) & UF 4(8) + 2UF ,(d) P, x% /x3 6.15 —-32.88

In Fig. 4.2, calculated equilibrium partial pres-
sures of the gases are plotted vs the UF,/UF,
ratio in the melt. As the oxidizing power of the
melt is increased, NbF . is expected to appear
first, followed by MoF ., and then RukF .. Uranium
hexafluoride has a lower dependence on oxidizing
power because its reduction product is UF , rather
than the metal. It was assumed in the case of
NbF ., MoF , and Ruk  that the reduction product
was the metal. The UK should not be formed
in significant amounts until the melt is oxidizing
enough to produce RuF,. If any stable inter-
mediate fluorides of Nb, Mo, and Ru are formed
in the melt, the result would be correspondingly
lowered equilibrium gas pressures and lowered
power dependences on the UF /UF ; ratio.

 

ORNL-DWG 67-773

   
 

PRESSURE {atm)

 

108 10 02 ot
XUFq/XUFa

Fig. 4.2. Eguilibrium Pressures of Yolatile Fluorides
as Functien of UFA/UF3 Ratio in MSRE Fuel.

UF6 UF3 UF4

which would permit inclusion of the fluorides of
technetium seem to be available.

Anolysis for Fission Products in MSRE Exit Gas

S. S. Kirslis F. . Blankenship

The only gas-liquid interface in the MSRE (ex-
cept for the contact between liquid and the gas-
filled pores of the moderator graphite) exists in
the pump bowl. There a salt flow of about 60 gpm
(5% of the total system flow) contacts a helium
cover gas which flows through the bowl at 4
liters/min. Provisions for direct sampling of this
exit gas are planned but have not yet been in-
stalled in the MSRE.

Samples of the liquid fuel are obtained by low-
ering a sampler, on a stainless steel cable, through
this cover gas and into the liquid. It has been
possible, accordingly, to detect chemically active
fission product species in this cover gas by ra-
diochemical analysis of the stainless steel cable
and its accessories which contact only the gas
phase and by analysis of special getter materials
which are attached to the cable. Coils of silver
wire and specimens of Hastelloy N have generally
been used as getters for this puipose.

No quantitative measure of the isotopes present
in the gas phase is possible, since no good es-
timate can be made of the gas volume sampled.
The quantity of material deposited on the wire
specimen does not correlate’ well with contact
time (in the range 1 to 10 min) or with the gstter
materials studied.

The quantity of material deposited, however,
is relatively large. Table 4.3 indicates relative
Table 4.3. Qualitative Indication of Fission Product
in MSRE Exit Gas

 

 

 

Amount?

Isotope On Ni On Ag  On Hastelloy From
Liquid®

9o 8 2 1 4
132p¢ 14 6 7 9
105gy 10 3 3 5
106, 6 2 1 1
1354 0 0 0
133 2 1 2 2
131 1.5 0.9 0.5 0.8

 

The unit of quantity is that amount of the isotope
in 1 g of salt.

POn stainless steel cable.

amounts found in typical tests. Volatile species
of Mo, Te, and Ru must certainly be presumed
to exist in the gas phase. The iodine isotopes
show perceptibly different behavior. Jodine-135,
whose tellurium precursor has a short half-life,
does not appear, while '3[ and 133 both of
which have tellurium precursors of appreciable
half-life, are found. These findings — along with
the fact that these iodine isotopes are present
in the salt at near their expected concentration —
suggest that any iodine in the vapor phase comes
as a result of volatilization of the tellurium pre-
CUursors.

Attempts to detect deposition of uranium (from
evolution of UFfi) on the wites have so far been
unsuccessful. This fact — along with the failure
to find many of the fission products which have
no volatile compounds — rules out the possibility
that salt spray is responsible for these obser-
vations.

It seems most unlikely that these data can be
reconciled as equilibrium behavior of the volatile
fluorides. Tt is possible that the MSRE metal is
plated with a noble-metal alloy whose thickness
is several hundred angstroms, and it is conceiv-
able that the UF /UF, ratio is near 10*. The
compound NbF_ (not tested for in the gas phase)
could show an appreciable pressure under these
circumstances, The other possibilities such as

MoF ., TeF , and RuF would require much higher

51

UF /UF, ratios, and it seems most unlikely that
any single redox potential can yield the relative
abundance observed for these isotopes.

The following speculation may be televant: As
the fission products, which originate in highly
electron-deficient states, thermalize and acquire
electrons in the melt, they pass through these
“unstable’” but volatile valence conditions. If
the plated reactor metal is sufficiently unreactive
and if (as it seems to be at present) the MSRE
fuel is quite deficient in UF ,, it is conceivable
that some fraction of these materials might ap-
pear in the gas phase and enter the MSRE graphite
or leave the system in the exit gas. If this is
true, then a considerable increase in UF, con-
centration in MSRE fuel might well markedly de-
crease the fraction in the vapor phase. It is
cleat that additional study will be required before
the situation becomes clear.

FISSION PRODUCTS ON METAL
AND GRAPHITE FROM MSRE CORE

S. S. Kirslis F. F. Blankenship

An assembly of MSRE graphite and Hastelloy N
specimens was exposed on the central stringer
within the MSRE core during its initial operation.
This assembly was removed during the July 17
shutdown after 7800 Mwhr of reactor operation,
and many specimens have been carefully examined.

No evidence of alteration of the graphite was
found under examination by visual, x~radiographic,
and metallographic examination. Autoradiographs
showed that penetration of radioactive materials
into the pgraphite was not uniform and disclosed
a thin (perhaps 1- to 2-mil) layer of highly radio-
active materials on or pear the exposed graphite
surfaces. Examination of the metal specimen
showed no evidence of corrosion or other danger.

Rectangular bars of graphite from the top (out-
let), middle, and bottom (inlet) region of this
central stringer were milled in the hot cell to
temove six successive layers from each surface.
The removed layers were then analyzed for sev-
eral fission product isotopes, 17

The results of analysis of the outer layer from
the graphite specimen are shown in Table 4.4,

 

17The initial sampling was carried out by J. G. Morgan,
M. F., Osbome, and H. E. Robertson. ‘Their help and
that of the Hot-Cell Operation Group is gratefully ac-
knowledged.
52

Tabiz 4.4. Fission Froduct Depesition on Surfoce® of MSRE Graphite

 

Graphite T.ocation

 

 

 

 

 

 

Top Middle Bottom
Isotope -
Percent Percent Percent

dpm /e of Total® dpm /cm * of Total® dpm /cm * of Total?

(x 10%) (x 10%) (x 10%)
?%Mo 39.7 13.4 51.4 17.2 34.2 11.5
13271¢ 32.2 13.8 32.6 13.6 27.8 12.0
1030y 8.3 11.4 7.5 10.3 4.8 6.3
*SNb 4.6 12 22.8 59.2 24.0 62.4
1311 0.21 0.16 0.42 0.33 0.33 0.25
957¢ 0.38 0.33 0.31 0.27 0.17 0.15 )
144ce 0.016 0.052 0.083 0.27 0.044 0.14 -
89 3.52 3.24 3.58 3.30 2.99 2.74 :
140, 3.56 1.38 4.76 1.85 2.93 1.14 ]
141ce 0.32 0.19 1.03 0.63 0.38 0.36
137 6.6x 107* 0.07 2.3x 1073 0.25 2.0% 1073 0.212

 

 

#Average of values in 7- to 10-mil cuts from each of threc exposed graphite faces.

bPercent of total in reactor deposited on graphite if each em? of the 2 x 10% cm? of moderator had the same con-

centration as the specimen.

It is clear that, with the assumption of uniform
deposition on or in all the moderator graphite,
appreciable fractions of Mo, Te, and Ru and a
large fraction of the Nb are associated with the
graphite. No analyses for Tc have been obtained.
The concentrations of these noble metals would
be sufficient to exert significant poisoning in a
breeder reactor.

The behavior of '#°Ba, 8%Sr, '41Ce, '**(Ce, and
137Cg, all of which have xenon or krypton precur-
sors, can be accounted for in terms of laws of dif-
fusion and half-lives of the precursors. Figure 4.3
shows the change in concentration of the fission
product isotope with depth in the graphite. Those
isotopes (such as !'49Ba) which penetrated the
graphite as noble gases show straight lines on the
logarithmic plot; they seem to have remained at
the point where the noble gas decayed. As ex-
pected, the gradient for '*%Ba with a 16-sec '*%Xe
precursor is much steeper than that for #9Sr, which
has a 3.2-min 89Kr precursor. All the others shown

show a much steeper concentration dependence.
Generally the concentration drops a factor of 100
from the top 6 to 10 mils to the second layer.

It is possible that carbide formation is respon-
sible for the deposition of Nb and possibly for -
that of Mo, but it seems quite uanlikely for Ru and
Te; the icdine probably got in as its tellurium pre-
cursor. Since these materials have been shown
to appear in the exit gas as volatile species, it
seems likely that they entered the graphite by
the same mechanism. The possibility that the
strongly oxidizing fluorides such as MoF
present raised the question as to whether Uk’
was accumulating in the graphite. An average
of 0.23 ug/cm? was found in the surface of the

were

graphite; much less was present in interior sam-
ples. This amount of uranium, equivalent to less
than 1 g in the core, was considered to be neg-
ligible.

Table 4.5 shows the extent to which various
fission product isotopes are deposited on the
13

GRAM OF GRAPHITE

DISINTEGRATIONS PER MINUTE PER

Fig. 4.3,

53

ORNL-CWG &7-774

1 20 30 40
DISTANCE FROM SURFACE OF GRAPHITE (mils)

 

o0

Concentration Profile of Fission Products

in MSRE Core Graphite After 8000 Mwhr,

Hastelloy N specimens in the core. A large frac-
tion of the molybdenum and tellurium and a sub-
stantial fraction of the ruthenium seem to be so
deposited. It seems possible that the '3 was
carried into the specimen as its tellurium pre-
cursor. The values for ?5Zr seem surprisingly
high, since those for the !'?iCe and '*%Ce with
noble-gas precursors probably reflect the amount
expected by direct recoil at the moment of fission.

If the Nb and Tc are assumed to behave like
the Mo, Te, and Ru, it may be noted that the MSRE
could have been uniformly plated during its oper-
ation with several hundred angstroms of relatively
noble metals.

XENON DIFFUSION AND FORMATION
OF CESIUM CARBIDE IN AN MSBR

C. F. Baes, ]Jr. R. B. Evans Il

Compared to the MSRE, a full-scale molten-salt
breeder reactor is expected to have approximately
50-fold greater neutron flux and 25-fold greater
flow velocity through the cote. Calculations have
been made!® in order to consider the extent of

 

18- F. Baes, Jr., and R. B. Evans 1II, MSR Program
Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, pp.
158-65.

Table 4.5. Deposition of Fission Products on Hastelloy N in MSRE Core

 

Hastelloy L.ocation

 

 

 

 

 

Top Middie Botfom
Isotope
2 Percent 2 Percent 5 Percent

dpm /cm of Total? dpm/cm of Total® dpm /em of Total?

(x 10%) (x 10%) (x 10%
29Mo 212 42.8 276 55.6 204 41.2
1320 508 131 341 88 427 110
103p4 35.5 20.3 25.5 21 23.2 19.1
1317 8.2 3.8 4.0 1.8 5.2 2.4
957, 1.8 1.0 1.8 1.0 2.6 1.3
1410, 0.05 0.02 0.22 0.07 0.15 0.06
1440, 0.01 0.02 0.09 0.18 0.35 0.07

 

TPercent of total present in reactor which would deposit on the 1.2 x 10% em? of Hastelloy N if deposition on all
surfaces was the same as on the specimen.
54

xenon diffusion and the behavior of daughter ce-
sium born in the moderator graphite of an MSBR
operating under such conditions. A one-dimen-
sional steady-state diffusion model was assumed
in which the moderator was represented as a slab
of graphite infinite in two dimensions, with a
thickness of 1 cm, immersed in the fuel salt.
It was further assumed that all cesium bomn in
the graphite was in the elemental (gasecus) form.
The parameters varied were: (1) the diffusion
coefficients D of xenon and cesium in the graphite
(assumed to be equal), (2) the {ilm coefficient H
associated with the salt-graphite interface, and
(3) the rate at which gas is stripped from the
fuel, r\ST.

Over the range chosen for these parameters,
the rate step in the diffusion of xenon into the
graphite was found to be at the salt-graphite in-
terface and was dependent on the value of X
(Fig. 4.4). As a consequence, a decrease in D
did not materially decrease the inward diffusion
of xenon;, however, it did decrease the rate at
which gaseous cesium diffused to the graphite
surface, where
taneously with the fuel salt:

it was assumed to react instan-

% { N + 1 +
Cs® L U4 =2U3T st

Thus, somewhat paradoxically, the maximum par-
tial pressure of Cs? (at the center of the slab)
was found to increase as I} was decreased.

Under all combinations of [, D, and /\ST values
chosen, the calculated cesium partial pressure
at steady state was high enough to cause the

formatien of lamellar cesium carbides (Fig. 4.4):
Cs?%g) + nC(s) = CsC_(s) .

However, the calculated accumulation rate of Cs
was s0 low that the amounts of CsC_ which could
be formed did not appear to he significant.

Finally, these calculations indicate that in the
absence of iodine removal (i.e., 6.7-hr 135I), Xe
poisoning in a full-scale MSBR will be controlled
primarily by the film coefficient H (Fig. 4.4) and
will be difficult to reduce tc an acceptable value
by gas stripping alone. It could be reduced more
effectively either by iodine removal or by some
means which effectively reduces the film coef-
ticient.

 

 

 

 

 

     

 

 

 

c
*O ,,,,,,,,, T .H. ——
1 (cm/sec)(s“_@.@fifl“""’
0,02 e
? o /
o
G
’.,
Q
=
b
g
n 10 |
o
&
@
x
0y
o
102 e -
o
(em?/sec)
1076
o
1075
E
s’
W
o
>
&
& 1o %
<
o
=
2
w)
L
(]
1o
ESTIMATED PRESSURE AT WHICH CsC, 1S FORMED
e e rmaama s e e e e
10 ¥ ‘
0.1 00 0.001
P\ST.(sec")
Fig. 4.4. Calculated Steady-State Pressures of Ce-
. . 13
siuin at Center of a 1-cm Graphite $lab oand *Xe

Poisoning os o Function of the Gas Stripping Rate
(’\ST)" the Diffusion Ceefficient (D), and the Film Co-
efficient (H).
the fraction of the maximum possible value {(a poison
fraction of 0.05). The flux is 7 x 1014 2

sec” ' and the graphite porosity is 0.05.

The *°Xe poisoning is represented &s

neuirons cm
Part 1l

Aqueous Reactors

 
5. Corrosion and Chemical Behavior in
Reactor Environments

NASA TUNGSTEN REACTOR RADIATION
CHEMISTRY STUDIES

G. H. Jenks H. C. Savage
E. G. Bohlmann

" Poison conticl solutions of CdSO, are being
considered by NASA Lewis Research Center for
possible use in the NASA Tungsten Water-Moderated
Reactor (TWMR). Information regarding the effects
of irradiation on the stability of these solutions
toward loss of Cd was needed in evaluations of
this poison control system, 1+?

We have conducted experimental investigations?
of the stability of CdSO, solution under electron
irradiation using the following experimental con-
ditions:

Solution compesition, 0.02 and 0.067 M CdSO4 in water

Temperature, 60 to 120°C
Radiation intensity, 73 and 145 w per cms of solution
Container, Zircaloy-2 with titanium filter
Agitation, static solution

3 . 2 3
Surface-area-to-volume ratio, 61 em™/cm

The container was in the form of a loop of 26-
wil-ID tubing with the titanium filter (3 p) at one
end, The solution was exposed within the tubing
for a period of time and then expelied through the
filter; the expelled solution was analyzed for Cd.

 

la. H Jenks, H. . Savage, and E. (. Bohlmann,
Reactor Cheni. Dive Ann. Progr. Rept. Dec, 31, 1965,
ORNIL.-3913, p. 58,

2G. . Jenks, E. G. Bohlmann, and J. C. Griess,
An Evaluation of the Chemical Problems Associated
with the Aqueous Systems in the Tungsten Water
Moderated Reactor, Addenda, 1 and 2, ORNL-TM-978,

NASA-CR-54214 (March 1965).

. M. Jeanks, H. C. Savage, and E. G. Bohlmann,
NASA Tungsten Reacior Radiation Chemistry Siudies,
Final Report, ORNL-TM-1630, NASA-CR-72070 (October

1966).

57

Small amcounts of Cd were lost from the solution
during 30-min irradiations at each tested combina-
tion of the above set of conditions. With 0.02 ¥
CdSO, solutions, the loss at 120°C and 145 w/cm3
was 5.0 + 3.4% error at 80% confidence. The loss
at 77°C was 3.3 + 2.8% and that at 77°C and 73
w/cm?® was 2.0 £ 2.7%. One experiment with 0.02
M CdSO, and H,S50,, pH 2, indicated negligible
loss.

With 0.06 M CdSO,, the loss at 60°C and 145
w/cm?® was 1.5 + 1.0%. At 120°C, the best indica-
tion was about 4% loss. The results of experi-
ments with 5- and 50-min irradiations of 0.067 M
CdSO4 at 60°C and 145 w/em? indicated that the
amount of Cd lost was greatest at the longer time.

Experimental information on recovery of the
separated Cd after irradiation indicated that the
rates of redissolution are slow.

Considerations of these results and of theory
suggest that Cd metal is formed under irradiation
and that this separates as relatively insoluble
material by agglomeration or by plating on solid
surfaces.,  Additional experimental investigation
of effects of agitation and of surface-area~to-
volume ratics would be required to predict the ef-
fects of radiation on stability in a reactor in
which these parameters differ from those in our
experiments,

Design and development work was done on a
system which could be used to study effects of
electron irradiation on stability in a dynamic
system.”  The planned dynamic experiment was
to be conducted with a small, high-speed (35,000
rpm) ceantrifugal pump with which solution was to
be circulated through a 26-mil-ID tube forming

 

*G. H. Jenks, H. C. Savage, and E. G. Bohlmanu,
NASA Tungsten Reactor Radiation Chemistry Studies,
Phase I, Experiment Design, QRNL-TM-1403, NASA-
CR-54887 (March 1966).
a loop in front of the cover plate of the pump.
The entire solution inventory was to be irradiated
The puipose of the tube was to
provide a channel in which film conditions could
be made comparable to those in the TWMR. The
results of component tests showed that the pro-
posed design was feasible. Detailed design
drawings of the equipment were reported,®

Work on this program was discontinued prior to
construction of the dynamic system because of
a lack of funds.

continuously.

CORROSION OF ZIRCALOY.2 BY DILUTE
HYDROGEN PEROXIDE AT 280°C

R. J. Davis T. H. Mauney

R. J. Hart

Heavy-particle bombardment accelerates the
corrosion of Zircaloy-2 in oxygenated aqueous
media® 8 but probably does not accelerate cot-
rosion in hydrogenated agqueous media.® Ionizing
radiations (beta or gamma) do not accelerate
Zircaloy-2 corrosion in agueous media,®!°

The above observations, along with interpreta-
tion'! of recent corrosion data'? and recent cal-
culations of the concentration of radiolytically
formed species in aqueous solutions,'® led to the
following hypothesis., Hydrogen peroxide is re-
sponsible for the acceleration of corrosion by
heavy-particle irradiation. Beta-gamma irradiation
produces peroxide concentrations too low for
notable corrosion acceleration.  Heavy-particle

bombardment also results in low peroxide concen-

 

SG. H. Jenks, H. C. Savage, and E. G. Bohlmann,
internal memorandum, 1966.

6G. H. Jenks, pp. 232—45 in Fluid Fuel Reactors,
ed. by J. A. Lane, H. G. MacPherson, and Frank
Maslan, Addison-Wesley, Reading, Mass., 1958.

“G. H, Jenks, pp. 41-57 in ASTM Spec. Tech. Pub.
No. 368, ASTM, Philadelphia, 1963.

8G. H. Jenks, R. J. Davis et al., HRP Quart. Progr.
Rept. July 31, 1958, ORNL-2561, pp. 234.-36; July 31,
1957, ORNI.-2379, pp. 115-21.

9B. O. Heston and M. D. Silverman, ORNL-CF-56-2-2
(February 1956).

105, J. Harrop, N. J. M. Wilkins, and J. N. Wanklyn,
AERFE-R-4779 (1664),

11B. Cox, private communication.
12w, A. Bums, BNWL-88, p. 23 (August 1965).

3q u1 Jenks, Effects of Reactor Operationn on HFIR

Coolant, ORNL-3848 (October 1965),

58

trations in aquecus solutions with excess hydro-
gen, but relatively high concentrations are formed
with excess oxygen. A peroxide concentration
of 107% M was estimated’? for a fast-neutron
(energy above 1 Mev) flux of 10?% neutrons cw =7
sec” L.

t'* was run in which Zircaloy-2

An experimen
specimens were cxposed to 107> M H O, at 280°C
for 297 hr.

tained by a continuous feed of 1072 M peroxide.

The peroxide concentration was main-

Control specimens were exposed to oxygenated
water in the same experimental setup.

The specimens and controls all gained weight
at average rates of 7 to 8 pg cm ™% day™ . There
This
rate of increase in weight is about a factor of 10
less than that known to occur as a result of a
fast-neutron flux of 10'? neutrons cm™? sec™!
on a system of Zircaloy-2 in oxygenated water
at 280°C.

It follows that the acceleration of corrosion of
Zircaloy-2 in oxygenated aqueous media by heavy-
particle bombardment is not due, solely at least,

was no significant effect due to peroxide.

to the hydrogen peroxide generated in the aqueous
environment,

AMODIC FILM GROWTH OMN ZIRCONIUM AT
ELEVATED TEMPERATURES

A. L. Bacarella H. S. Gadiyar'®
A. L. Sutton

In our previous report'® we postulated that the
current (f) for anodic film growth on zirconium in
oxygenated, dilute H,50, at temperatures from
174 to 284°C is an exponential function of the
tield strength, Vz/X’ across the oxide filin. This
relation is given by

(qay*V,
1=1 exp ——_< (1
o P X )
where { is the anodic current (amp), 1, is the cur-

0
rent at zero field strength (amp), (ga)* is the prod-

uct of the charge of the mobile ion (e) and the

 

Y4p  I. Davis, T. H. Mauney, and J. R. Hart, J.
Electrochem. Soc. 113, 1222 (1966).

1SAiien Guest from the Indian Atomic Energy Estab-
lishment, Bombay, India.

I6A, 1.. Bacarella and A. L. Sutton, Keactor Chem.
Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789,
pp. 135—38.
activation distance {(cm), V2 is the poteatial dif-
ference across the oxide layer (v), X is the film
thickness (cm), and k7 is the thermal energy
equivalent (ev), Calculation of the magnitude of
the *‘activation dipole,”” (ga)*, showed that the
activatinon distance {a) increased from 13.5 A at
25°C to 37 A at 284°C. These large values were
explained’®~ 1% according to a model which con-
siders dielectric polarization of the oxide and
predicts that the effective field causing ion migra-
tion is greater than the average applied field.

 

Using the Mossotti-Lorentz field, the apparent
activation dipole is
o (e+2)
(ga)* = 3 (qa) »

where ¢ is the dielectric constanl of the oxide
(dimensionless) and (ga) is the ‘“‘true’’ activation
dipole.

We showed further that at large film thickness
{small field strength), current flow of mobile charge
against the field becomes significant
and that the net {ilm growth current may be ex-
pressed by a hyperbolic sine fuaction of the field
strength:

carriers

 

_ R CONE —(gaf)V,
P KD oo o @K =
o [T T P TTRIX
(ga)*V
= 24 sioh el 2
0 kTX @

More recent measurements in dilute K, 50, solu-
tion, particularly measurements of the Tafel slope
at constant film thickness, (2 log i/(?VZ)X, showed
the need for two additional modifications to the
In K,80, solution, it was found that the
deviations from expected behevior at large film
thickness were greater than could be accounted
for by the current flow of mobile charge carriers
against the field. Also, measurements of polariza-
tion curves and Tafel slopes at constant film
thickness were not in
with Eg. (2). Differentiation of Eq. (2) shows

model,

 

17A. I.. Bacarella and A. 1.. Sutteon, J. Electrochem.
Techaol. 4, 117 (1966),

Ly, J. Dignam, J. Electrochem. Sec. 112, 722--29
(LOES).

%R J. Maurer, J. Chom. Phys. 9, 579 (1941).

59

satisfactory agreement

that the Tafel slope is given by Eg. (3), where

B = (qa)*/kT:
() (%) ©

The experimental results from Egs. (2) and (3)
were found to be greatly improved if a constant
150 A were added to all film thicknesses (X).
Furthermore, with the use of this correction, de-
viations observed at very small film thicknes-
sest® 17 were also accounted for.

Use of an additional constant in Eqs. (2) and
(3) can be justified by a more general derivation
of the relation for anodic film growth. Equation
(2) was derived on the assumption that the fraction
of the total potential difference, from metal to

B
T eeeeens h
2.3X

dlog 1
v,

BV,
X

2B(i, /1)
-— X

2.3X

 

 

solution, which affects ion transport is that por-
tion which exists across the oxide film, The
potential differences at the metal-oxide and oxide-
solution interfaces were considered to be con-
stant and independent of 7,
treatment,

In a more general
we take into account the possibility
that charge transport across each interface may
also affect the potential distribution. We have
therefore extended the formalism of the so-called
dual-barrier model®® 7?2 to a triple-barrier problem

and derived the rate expression of Eq. (4):

i = kfa/ilf%fi)kzk;6&/()("r6a)k§a/(x+fia)

24
ex e nnne
SR X 4+ bz

In Eq. (4), V=V +V, +V, is the total potential
difference between metal and solution phases; it
equals the sum of V. (potential difference between
metal and oxide phases), Vz (potential difference
across oxide layer), and ¥, (potential difference
between oxide and solution phases), The param-
eters k , k,, and &k, are the corresponding rate
constants of the charge transfer processes at
each barrier. The [ractional exponents are ob-
tzsined from the assumptions that the electrochemi-

(4)

cal transfer coefficient at the wetal-oxide barrier

 

WA L. Bacarella and A. L. Sutton, J. Electrochem.
Sce. 112, 546 (10865).

Mg . Meyer, J. Blectrachem. Soc, 110, 167 (1963).

226 A, Posey, G M. Cartledge, ond R. P. Jaffee,
J. Electrochem Soc, 106, 582 (1959).

23'. T, MacDonald and B. E. Conway,

. FProc, Roy.
Soc. A269, 419 (1962).
60

is given by ga = (2) (0.5) = 1, while that for the
= 2a/X and that for the oxide-
Differentiation of

oxide barrier is qa,
solution barrier is ga, = 0.5.
Eq. (4) shows that the Tafel slope is given by

Eq. (5):
d log 1 B .
e (5)
av /. 23(X+6a)

where B = 2a/kT. Equations {4) and (5) were
found to be very satisfactory in accounting for the
data in both H,SO, and K,SO, media, where to a
first approximation the activation distance a =
2e + 2)/3.

In a great majority of the experiments performed,
the potential of the zirconium electrode was main-
tained constant at 10.0 v vs a Pt reference elec-
trode. This potential is about +1.0 v more noble
than the open-circuit corrosion potential, E  (cf.
Fig. 5.1). The Tafel slopes were determined
over a range of about 0.3 to 0.4 v fiom this poten-
Over this potential range the Tafel slopes,

 

tial.

5x10° %

1079

CURRENT (amp?
™

44 12 40 -08 -0 -04 -02 O 0.2

(0 log i/dV), ., were apparently linear, and Fgs.
(4) and (5) described the data satisfactorily. Re-
sults of a much more extensive anodic polariza~
tion measurement (covering a 9-v range) are pre-
sented in Fig, 5.1. We find that the Tafel slope
is not constant but decreases somewhat with
increasing V.

The foregoing observation may be rationalized
on the basis of Dignam’s'®~2*% theory, which
postulates a field-dependent transfer coefficient
in the oxide phase. Incorporation of this theory
in our triple-barrier model leads to Eq. (6), ex-
pressed in logarithmic form:

21 Qav,/edX)] | KN v
= — nn —-- —_—
"X ed1 - aV,jed0l | \& ) kT

n

 

+lnk,. (6)

 

 

04 08 08 1.0 20 30 40 50 80

24M. J. Dignam, Can. J. Chem. 42, 1155 (1964).

ORNL-DWG 67-775

POTENTIAL [volts vs Pt {ref) ]

Fig. 5.1. Polarization Curve for Zirconium in Oxygenated 0.05 m K950, Solution at 203°C and Constant Film

Thickness 700 A,
In Eq. (6) the potential ¥V, (the potential difference
across the oxide layer) appears implicitly in the
rate expression, and ¢ is the activation energy of
the anodic reaction. The parameter ¢ is not an
independent constant in Dignam’s theory; %% it is
a function of the quantity 2aV,/¢$X. Values of ¢
may be computed on the assumption that the inter-
action potential between the mobile ion and its
immediate surroundings can be represented by a
Morse function for small displacements. The
validity of Eq. (6), which was applied to data
shown in Fig. 5.1, is indicated by the results
shown in Fig. S5.2. Here the activation energy
¢ = 1.35 ev was determined by us in previous ex-
periments, and V, = 2.3 v was assumed to be a
good first approximation.” We find that a value of
a = 26.5 A may be estimated from plots of leg 1 vs
1/(X + 6a) for large X. This value of the activa-
tion distance seems to be quite reasonable. We
conclude that this model is capable of describing
the anodic film growth process in zirconium over
a wide range of electrode potentials and film thick-
nesses with satisfactory accuracy by use of reason-
able values of pliysical parameters.

GRNL-DWG 657-7 76

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

. - | -
0 I y
9
3}

"

o 7

%

i

Ll

E o

a

¢ B

-

< 4

=

=

kzl

5

[o 8 .,

2 .‘.‘..vi
., / ______________ B
1
G i e R P —]
O 20 40 80 30 iC0 51/120 140 160 180 200 220
v g3 (1. 22k
sl
P A LA 12%10™8
53(-3s)

Fig. 5.2. Polarization Dota of Fig. 5.1 Plotted Ac-
cording to Eq. (6).

o1

AC IMPEDANCE OF OXIDE FILMS IN AQUEOUS
SOLUTIONS AT ELEVATED TEMPERATURES

G. H. Jenks R. J. Davis
A. L., Bacarella H. 8. Gadiyar!®

We are in the process of developing equipment,
methods, and techniques which will permit us to
measure the ac impedance of corrosion films on
zirconium and its alloys in aqueous solutions
at elevated temperatures. The immediate objec-
tive of these measurements is to determine whether
pores or fissures which admit aqueous solutions
occur in these films during corrosion at tempera-
tures up to about 300°C. If these determinations
can be made satisfactorily, we will then attempt
to conduct similar experiments in-pile. It is ex-
pected that valuable information on the electrical
properties of the high-temperature corrosion films
will be obtained also during the course of the
me asurements.

In the method employed, the specimen is im-
mersed in an electrolyte which, together with the
metal container, comprises one electrode for im-
pedance measurements of the oxide film on the
specimen. The metal of the specimen is the second
electrode., Measurements are made of the capaci-
tance and resistance of the oxide over a range of
frequencies: and at two or more electrolyte concen-
trations, The results are examined for behavior
expected to result from penetration of the electro-
lyte into fissures within the oxide.

The principles upon which this method of pore
detection is based were discussed by Young.??®
Wanklyn and co-workers?% 27 have used ac im-
pedance measurements at room temperature in
studies of the protective and electrical properties
of oxide films formed on zirconium alloys during
high-temperature corrosion.

Our measurements to date have covered a tem-
perature range of 25 to 200°C and have been made
on a single zirconium specimen bearing a 1500-A
film formed anodically (0.0 v vs Pt at 55°C) in
oxygenated 0.05 m H, S0, at 220 to 230°C in the
Ti electrochemical cell.?®  The electrolyte for

 

257, Young, Anodic Oxide Films, pp. 150-70, Aca-
demic, New York, [961.

2

61.
(1966).

271. N. Wanklyn and D. R. Silvester, J. Electrocheam.
Soc, 105, 647 (1958).

34 L.
(1961).

N. Wanklyn, Electrochem. Technol. 4(3-4), 81

Bacarella, J. Elecirochem. Soc. 108, 331
one series (25, 50, 100, and 198°C) of impedance
measurements was oxygenated 0.05 m H,SO,.
For a subsequent series {25, 50, 100, and 150°C),
the electrolyte was oxygenated 0.05 m X, SO,.

The impedance cell was made up of a small
Zircaloy-2 autoclave containing an insulated gold
cup 1/4 in. in inside diameter x 1/2 in. long, which
held the electwlyte. The 0.2-cm rod-shaped speci-
men was positioned along the axis of the cup and
proiruded into the gas-steam phase. The bottom
end of the specimen was covered and sealed with
Teflon, so that only the axial surfaces were in
with the electrolyte.  FElectrical leads
from the specimen, gold cup, and from a small
platinum reference electrode which dipped into the
electrolyte through the gas-steam
phase and out of the autoclave through a Teflon
In operation, gas (O,-He) pressurization
was used to prevent gas-bubble formation within
the electrolyte of the gold cup. This was ac-
complished by adjusting the pressure upward as
the temperature was raised, so that the pressure

contact

were passed

seal.

o
~

of dissolved gas never exceeded the overpressure.
A General Radio 1615-A bridge with three terminal
connections was employed in impedance measure-
ments,

The results of series capacitance measurements
on the single specimen over a range of tempera-
tures and freqnencies are illustrated in Fig. 53.
The slopes of these straight lines are listed in
Table 5.1. The results of the corresponding
series resistance measurements for the film fell
near straight lines in plots of R_ vs 1/f. These
lines passed through the origin after correction
for the electrolyte resistance. The slopes of
the lines in these data plots are also listed in
Table 5.1.

Tentative evaluations of these data with regard
to overall consistency and with regard to evidence
of penetration of electrolyte were based on the
theoretical expressions for K_ and C_ discussed
by Young, *°

479 « 101 1d

S €A log, , [p(d)/p(O)]

1/C

) € p(d)
x |log,, f+log, 181001

79 x 101 1d 1
R o
S edlog, [p(d)/pO)] f

(7)

(&

62

Young derived these equations by emwploying a
mode! in which the resistivity, p, varies exponen-
tially with distance through the oxide, The re-
sistivities at the opposite surfaces of the oxide
of thickness d are p(0) and p(d) respectively; € is
the dielectric constant and A is the area. These
equations reproduced Young’s experimental re-
sults for anodic films on Nb with respect to straignt-
line formation in plots of 1/C_ vs log f and of R_
vs 1/f and with respect to the rclationship between
the slopes of the 1/C_ and R plots.?® The fre-
quency dependencies of C_ and R_ observed by
otherz for anedic and corrosion films on Zr have
beern ascribed to conductivity gradients thiough
the oxide.?”

Values of p(d) and p(0) determined from our
capacitance data employing Eq. (V) are listed in
Table 5.1. Plots of the p(d) values are shown in
Fig. 5.4. The ratios of the slopes of the 1/C_ and

(x 108 T
M2 - s

08 | b bk
C O

 

100

 

  

 

00°C 1
, |
| / !
80 ‘

 

 
        

76

72 b -

68

64 | —nr -

  

  

80 |- - ST g bty -t [
[ :
Cd

 

 

i

[ ! 1
cd

Lo | i
ot

.

ol
i
5

56 , ) .
of Q.2 05 2 {0 20 50
FREQUENEY fkc)
Fig. 5.3.  Variation of 1/C_ with Frequency and
Temperature.
63

Table 5.1. Electrical Properties of Zirconium Oxide Corrosion Film at Temperatures Up to 198°C

 

 

 

 

a M2b A
7 c Cc
Temperature Miz /.{_I (ohms, cm”, D O
13) : ’ (cm £ ) cycles) (ohm-~cm) (ohm-cm)
°c S °s)
H,S0, K,S0, i, 50, K, 50, H,SO, K,SO, H, S0, K,S0,
% 10° % 10° « 10% % 10%
25 4.3 4.9 5.02 5.94 2.3 x 10°° 2.0x 10 3.7x10°  3.1x10°
50 4.1 5.7 4.98 6.42 2.2 % 1020 2.7 % 10" 4.5 x10° 2.7 x 10°
100 5.1 7.3 5.78 7.52 5.2 % 1016 3.8 « 101  6.5x%10° 2.7 x10°
150 8.7 9.15 6.1 % 101! 2.4 % 10°
198 7.8 8.70 2.9 % 1012 2.9 % 10°

 

fSlope of line in plot of l/CS Vs loglo f.
bSlope of line in plot of total series resistance vs 1/f.

®Dielectric constant at 20 ke employed in calculations.

ORNL—DWG 67-778
48

 

R_ plots varied between 8.3 and 9.7. A value of
9.2 is predicted by Egs. (7) and (8).

Qur tentative conclusions are that the observed
frequency, temperature, and electrolyte dependence
of the impedance components can be reasonably
ascribed to resistivity gradients through the oxide
and to changes, with temperature and with electro-
lyte, of the resistivity of the layers of oxide in
contact with the electrolyte. There is no evidence
for fissures which admit electrolyte into the oxide.
Additional work is needed to confirm these con-
clusions and to permit more complete explanations
of the impedance behavior,

 

46 1~ @ 005 M H,50, ELECTROLY
© 0.05 M K,50, ELECTROLYTE

 

     
 

 

 

 

 

 

 

 

We expect to continue these studies. Some
modifications of the cell will probably be neces-
sary to permit measurements with smaller speci-

/ men areas. Also, it may be necessary>to modify
the method of pressurization to permit better con-

{é‘"‘SLOPE CORRESPONDS trol of pressure on the electrolyte.
TO ACTIVATION ENERGY

OF 30 kcal /mole FOR

 

 

34

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CONDUCTANCE
................ CORROSION SUPPORT FOR REACTOR
PROJECTS
“““““ J. C. Griess, Jr. j. L. English
P. D, Neumann
26
24 23 25 27 29 34 33 35
000/ /0y Experimental programs conducted for the purpose

of selecting and estimating the corrosion damage
Fig. 5.4, Variation with Temperature of Resistivity to engineering materials in the High Flux Isotope
of Quter Surface of Zirconium Film, Reactor (HFIR) and the Argonne Advanced Research
Reactor (AARR) were completed during the past
year, and the results were presented in two re-
ports.29'3° Summaries of the pertinent results
are given below. Results concerning the corrosion
of the aluminum-clad HFIR fuel elements have
been reported previously®! and are not included
here.

The HFIR, which is presently operating at the
design power of 100 Mw, uses water adjusted to a
pH of 5.0 with nitric acid as coolant; all tests
were conducted in this environment at 100°C.
Aluminum alloys (6061, 1100, and X-8001) cor-
roded at rates of 0.2 mil/year or less at all flow
Contacting aluminum with
aluminum, nickel, or stainless steel
resulted in pitting of the aluminum in contact
areas only. The pits were of appreciable depth,
but they were randoinly spread and should not be
of major consequence in the HFIR system. At
velocities of 13 to 81 fps, beryllium corroded at
a constant rate of 2 mils/year and showed only =
slight tendency to pit. Corrosion damage to the
beryllium reflector in the HFIR should be negligibly
small.  Both nickel and electroless nickel de-
posits corroded excessively in the acidified water
and are not usable in the HFIR. ¥Electrolyzed
coatings (a proprietary process for coating ma-
terials with chromium) on stainless steel exhibited
excellent corrosion resistance and good adherence
Similar deposits flaked and
Hardened types

rates up to 81 fps.
beryllium,

to stainless steel.
peeled from aluminum surfaces.
416 and 420 stainless steel were shown to be un-
suitable; the former alloy blistered, and the latter
underwent stress-corrosion cracking.

The results of this testing program and the ex-
perience gained during operation of other water-

 

297, L. English and J. C. Griess, Dynamic Corosion
Studies for the High Flux Isotopc Keactor, ORNL-TM-
1030 (September 1966).

30]. C. Griess and J. L. English, Materials Com-
patibility and Corrosion Studies for the Argonne Ad-
vanced Research Reactor, ORNIL.-4034 (November 1966).

317, ¢. Griess, H. C. Savage, and J. L. English,
Effect of Heat Flux on the Corrosion of Aluminum by
Water, Part IV, ORNL-3541 (February 1964).

64

cooled production and research reactors were fully
used in the design of the HFIR. Assuming ade-
quate control of the water chemistry, the HEFIR
should be free of major corrosion problems.

The primary areas of concern in the AARR were
the beryllium reflector, the 5061-T6 aluminum
beam tubes, and the stainless steel fuel element
cladding. All tests were conducted in deicnized
water at 93°C (200°F). The results indicated
that the corrosion rate of the bLeryllium reflector
in the AARR will be between 1 and 3 mils/year
with a minimum of localized attack.

The coriosion of 6061-T6 aluminum
results in the formation of an insulating layer of
corrosion products on the surface. Since heat
generated in the beam tubes must be transferred
across this insulating laver, the coriosion rate
of aluminum wmust be minimized to prevent exces-

The
initial

in water

sive temperatures in the beam-tube walls.
AARR beam-tube cooling system its
design would allow surface temperatures up to
260°F; under these conditions excessive film
formation would occur and would cause increases
in temperature of as much as 66°C (150°F) during
75 days of operation. Tests showed that the cool-
ing system would have to be capable of keeping
the temperature at the aluminum-water interface
at 200°F or less if frequent replacement of beam
tubes was to be avoided.

1i1

Type 304 stainless steel did not develop ap-
preciable deposits on water-cocled surfaces when
exposed under thermal conditions similar to and
even more severc than those anticipated in the
AARR fuel elements. However, after a 1000-hr
exposure at a heat flux of 3.8 x 10° Btu hr!
ft—
the cooled surface. In a comparable test in which
the heat flux was 2 x 10° Btu hr™?! ft "2 (hot-spot
heat flux for 100 Mw operation of the AARR) and
the exposure time was 2000 hr, no cracks were
found. The cause of the obseived cracks was oot
determined, but it is possible that they were
related to high thermal stresses existing in the
specimen.

2, numerous shallow cracks were preseat on
6. Chemistry of High-Temperature Aqueous Solutions

ELECTRICAL CONDUCTANCES OF AQUEDUS
ELECTROLYTE SOLUTIONS FROM
0 TO 800°C AND TO 4000 BARS

A, S. Quist W. Jennings, Jr.
W. L. Marshall

There is relatively little information available
on the properties of aqueous electrolyte solutions
at superciitical temperatures and pressures. The
simplest and most direct method f{or obtaining
information about the existence and behavior of
ions in these solutions is to measure their electri-
cal conductances, By using this method, we have
studied aqueous solutions of K,50,, KHSO,, and
H,50, to 300°C and to 4000 bars; from these
measurements we have calculated [lirst and second
After
extensive conductance measure-

ionization constants of sulfuric acid.!-?
these studies,
ments were made on NaCl solutions (0,001 to
0.1 m) in order to observe the behavior of a strong
uni-nnivalent electrolyte under the same condi-
tions.?  The experimental data were evaluated
with the usual theoretical equations that describe
equivalent conductance as a function of concentra-
tion. By using a digital computer to fit the data
to the equations by nonlinear least-squares methods,
limiting equivalent conductances were calculated
at integral temperatures (0 to 800°C) and densities
(0.4 to 1.0 g/em?®). These values are shown
graphically in Fig. 6.1, where limiting equivalent

 

1A. S. Quist et al., Reactor Chem. Div. Ann. Progr.
Rept. Jan. 31, 1962, ORNL-3262, pp.
Chem. Div. Ann. Progr. Rept Jan. 31,
3417, pp. 77—82; Reactor Chem. Div. Ann. Progr. Rept.
Jan. 31, 1964, ORNL-3591, pp. 8488, Reactor Chem.
Div. Ann. Progr. Rept. Jan. 31, 19635, ORNL-3789,
pp- 13943,

2A. s Quist et al., J. Phys. Chem. 67, 2453 (1963);
69, 2726 (1965); 70, 3714 (1966).

3. S.. Quist ef al., Reactor Chem. Div. Ann. Progr.
Rept. Dec. 31, 1965, ORNL-3913, pp- 63--64.

73-75; Reacilor
1963, ORNI.-

65

URRL-DWG €5-TRY/R

 

 

 

 

 

 

 

 

 

 

1500 & . e I s Q00 ]
L”\/’Tx e » S00°
das TR
[ . s §00°
sy ® 007
. 1000 — ——r W o v B00° -
2 e T i
£ : T,
- i — TR i
= ; T, ljuo*’
e
m ,,,,,,,,,,,,,,,,,,,,,,, :T,,E:{,,
|
#orem i
|
ol L ) ao
04 Q3 06 o7 08 G2 10
DENSITY tg/bm)
Fig. 6.1, Limiting Equivalent Conductance of MNaCl

as a Function of Density at Temperatures to B00°C,

conductances are plotted as a function of density.
This graph indicates that the limiting eguivalent
conductance of NaCl is a linear function of scolu-
tion density (at constant temperature)., Figure
6.1 also reveals that at tempervatures of 400°C
and above, the limiting equivalent conductance of
NaCl appears to be independent of temperature
(at constant density).  Although the values at
800 are slightly helow the 400 to Y00°C values,
this difference is mostly a consequence of in-
creasing ion-pair formation at higher temperatures.
As NaCl becomes a weaker electrolyte, the extrap-
olated limiting equivalent conductances become
somewhat uncertain and tend to fall below their
true values, These solutions showed this behavior
al densities below 0.4 g/cm® and at temperatures
above 700°C., Disscciation constants for ion-pair
formation in Na(Cl sclutions were calculated from
the Shedlovsky equation® at densities from 0.3
to 0.75 g/cm?® and temperatures from 400 to 800°C.

 

4. Shedlovsky, J. Franklin Inst. 225, 739 (1938).
Figure 6.2 shows the dependence of the logarithm
of these dissociation constants on the logarithm
of the density of the solution at the several
temperatures,

An extensive series of measurements were made
using 0.01 demal (0.01 mole per 1000 g of solution)
KCl solutions. As was mentioned previously,® a
KCl solution of this concentration appears to be
the logical choice as a reference for conductance
measurements at elevated temperatures and pres-
sures, since it is used as a standard solution for
cell constant determinations at 25°C. Accordingly,
many measurements have been made in this Lab-
oratory over a period of several years on 0.01
demal KCl solutions. Some of the results are
shown in Fig. 6.3, where the specific conductances
are plotted as a function of temperature at a con-
stant pressure of 4000 bars.
the results of 50 separate runs, using three con-

This graph represents

ductance cells and seven different inner electrode
assemblies,

A comprehensive investigation of the electrical
conductances of alkali metal halide solutions and
solutions of related compounds was initiated and
is presently nearing completion. Measurements
have been completed from 0 to 800°C and to 4000
bars on 0.01 m solutions of the following electro-
lytes: NaCi, NaBr, Nal, KCi1, KBr, KI, Rbl", RbCI,
RbBr, Rbl, CsCl, CsBr, Csl, NH Br, (CH3)4NI3T,
HBr, and NH OH. Some preliminary results are

ORNLDWGE €6-BO96R
DENSITY (g/£m®)
030 040 050 060 070 080 090 100
O - .‘ = =
\ I ‘ ‘ | [ I

 

log A (NaCl=Ng*+Ci™)

 

 

 

log DENSITY

Fig. 6.2, Log K(NuCl = Na '+ Cl”} as a Function
of Log Density at Temperatures from 400 to 800°C. K in

molar units.

 

ORNL-DWG £7-530

 

 

 

 

 

 

 

 

 

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-~ | !
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2 400 | _ — —
= i [
Q !
e e R | — — | —
P ! | ‘ 1
Q ! l .
© 200 g | 71 o b
o i ‘ i
e | : | !
§100, e e
a ! ! !
[ — 1 { i i L,,iJ . -
0 200 400 £00 800
TEMPERATURE (°C)
Fig. 6.3. Specific Conductance of 0.01-demal KCI as

Pressure, 4000 bars.

a Function of Temperature.

ORNL-DWG 67-531

800 | e e 5

 

700

600

500

400

300

200

SPECIFIC CONDUCTANCE (ohm™" e hx 1058

100

 

 

 

|

4

\

|
Q 200 400 600 800
TEMFERATURE (°C)

Fig. 6.4, Comparison of the Specific Conductance of
0.01 m Solutions of RbF, RbCI, RbBr, and Rbl as a
Pressure, 4000 bars.

Function of Temperature,

shown in Fig. 6.4, where specific conductances
of the rubidium halide solutions are plotted against
temperature at a pressure of 4000 bars.

DISSOCIATION CONSTANT OF MAGHNESIUM
SULFATE TO 200°C FROM SOLUBILITY
MEASUREMENTS?

W. L. Marshall

There are relatively few experimentally deter-
mined values for dissociation constants at high

 

5Jointly gsponsored by the Office of S8Saline Water,
U.S. Dept. of the Interior, and the USAEC.
temperatures, as contrasted to the very large num-
ber at 25°C.. In this Division, values above 100°C
have been obtained from conductance® and sol-
ubility measurements.’ Any contribution to the
equilibrium behavior of electrolytes at high tem-
perature, especially the behavior of 2-2 electio-
lytes, is of very much fundamental and applied
interest. The equilibrinm of considerable interest
to the high-temperature behavior of seawater and
ofher saline wsters is the dissociation equilibrium
of magnesium sulfate represented by '

0 L 2
MgSO,° w=f=Mg*" +50 *7, 0

where (0, is the dissociation guotient at an ionic
strength I, From the increase in solubility of
calcium sulfate (or its hydrates) in seawater con-
centrates compared to its behavior in aqueous
sodium chloride solutions, and with the assumption
that this increase is due predominantly to the
formation of the magnesium suifate complex, dis-
soriation gquotients could be calculated at many
ionic strengths. The equilibrium of Eq. (1) and the
following solubility product equilibrium,

K__ . — .
CaSO (s) gl Ca?t 150,77, (2)
where K_  equals the solubility product at [ and

includes any contribution from a neutral species,
{338040, wetre used to obtain the tollowing equa-
tions:

I" = [ (formal) - 4« [MgSO4°] , {3)

K;p = a2 |[total sulfate]

= [Ca 2+]1_SO4 27+ M50 0

 

=K+ a1 Mgs0,%), (4)
(MgsG 1 = (K, — &K )/ICa’"], (5)
[total magnesium — MgS0 “lltotal sulfate — MgSO 9]
Ry = -t et (6)
(Mps0O,°) '

Values for K5p, [Ca 2, total magnesium, and total
sulfate were obtained from the experimental results

 

6A. S, Quist et al., “*Conductance of Electrolytes to

300°C and 4000 Bars,’” contained in another part of ibis
section; J. Phys. Chem. 67, 2453 (1963); 69, 2726
(1965); 70, 3714 (1966).

“W. L. Marshall and . V. Jones, J. Phys. Chem. 70,

4028 (1966); Reactor Chem. Div. Ann. Progr. Rept.
Jan. 31, 1965, ORNI.-3789, p. 145.

given elsewhere in this report.® The values for
K., as a function of ionic strength were calculated
from the solubility behavior in sodium chloride
solutions. /1% By an iterative process, where I’
is initially assumed to equal I (formal), Egs. (3-6)
were solved to obtain values of Qd and [’ at the
many different ionic strengths. These values were
extrapolated at each temperature by means of an
extended Debye-Hiickel equation,

log Q= log K5+ 85\ I'/(L+ VI, (7)

where S ig the Debye-Hickel limiting slope for
a 1-1 slectrolyte, to obtain the dissociation con-
stant KS (at I =0).

Values of the constant to 2007 obtained by this
method are plotted as the logarithm of Kg vs
1/T(°K) in Fig. 6.5. Included in Fig. 6.5 are many
published values for this constant obiained by
several different methods al temperatures up to
40°C, the highest temperature of previously pub-
lished work. The present results appear to be in
good agreement and extend the values for the con-
stant to 200°C.

By least-squares fitting the values of Nancollas, !
Jones and Monk,’? others at p°C,'#:1* and those
to 200°C from this present study to the van’t Hoff
isochore,

d1n Kg
1/7)

 

- ~AH%/R (8)

with AH® expressed as a function of [\C; {a con-
stant) and T(°K), thermodynamic gquantities were
calculated; the results are given in Table 6.1.

 

3W. 1.. Marshall and R. Slusher, **Solubility of Calcium
Sulfate in Sea Salt Solutions to 2007C; Temperature-
Solubility Limits for Saline Waters,” included in this
Annual Report in another section; (Chemical Support for
Saline Water Program.

gw. L. Marshall, R. Slusher, and E. V. Jones, J. Chem.
Eng. Data 9, 187 (1964).

w. L. Marshall and R. Slusher, J. Payvs. Chem, 70,
4015 (1966); Reacior Chem. Div. Ann. Progr. Repl.
Dec. 31, 1865, ORNIL.-3913, p. 113.

]‘1(3~ H. Nancollas, Discussions Faraday Soc. 24, 108
(1G57];.

le. W. Joneg and C. B. Monk, Trans. Faraday Soc.
48, 929 (1952).

13_]. Kenntamaa, Suomen Kemisfilehti 298, 59 (1956).

My G. M. Brown and J. ¥. Prue, Proc, Roy. Soc.
(London) 232A, 32G (1955).
Table 6.1.
Dissociotion of Magnesium Sulfate in Aqueous
Solution, 0 to 200°C; Average &CS

Found = 25 cal mole“fi] deg

68

ORNL-CWG 06-7797R

TEMPERATURE (°C)

200 150 100

 

4.5? """" T T l}

® PRESENT WORK
(MARSHALL —1966) SOLY.

 

50 25 0

........ — T T i

ATKINSON, PETRUCCI (1966) , ULTRASUNICS
EIGEN, TAMM (1952), UL TRASOMICS
MANCOLLAS (*957),COND., pH, SOLY,

KENT FAMAA #1956, COND.

 

 
 
 
 
 
 
 
 
 
   

0
“

BROWN, PRUE (1955), MIDVALUE ,Fi. PT, —|
DUNSMORE | JAMES - REEVALUATED BY
JONES, MONK {1952) , COND.

JONES, MONK (1952) £ M.F.
DUNSMORE , JAMES (1951) COND.
DEUBNER , HEISE (1951}, COND.
MASON, SHUTT {1940}, DIEL. CONST.
DAVIES (1928), COND.

MONZY  DAVIES (1932), COND.

DAVIES (#927), COND.

0Ooocodo e

> 0O ¢ 0 <49 0 0

 

0
—log &, (Mg$S

 

 

 

Fig. 6.5. Dissociation Constant of Magnesium Sulfate from 0 to 200°C,

Thermodynamic Quantities for the

-1

 

3 O !XSO

T("C) --log K, AF AI? (cal mole !
(kcal/moale) (kcal/mole) degul)
0 2.129 2.06 —4.08 —24.7
25 2.399 3.27 —4.01 24,4
50 2.631 3.89 -—-4,272 -25.1
100 3.057 5.22 —5.41 —28.5
150 3.501 6.78 —7.67 ~34.1
200 4,002 8.66 —11.0 —41.5

 

DISSOCIATION CONSTANT OF CALCIUM
SULFATE TO 350°C OBTAINED FROM
SOLUBILITY BEMAYIOR IN MIXED
ELECTROLYTES®

L. B. Yeatts W. L. Marshall

The dissociation quotients, Qd’ for CaSO4° were
determined from extensive solubility measurements
of calcium sulfate (or the dihydrate at 25°) in an
aqueous system of mixed electiolytes, varying from
pure NaNO_ to pure Na SO, when possible. Mea-
surements were made at several constant ionic
strengths from 0.25 to 6 m at 25, 150, 250, and
350°C.

When solid calcium sulfate dissolves in aqueous
solutions, it may be assumed that an equilibrium
is reached with undissociated molecules:

CaS0 ,(5)==>CaS0,%aq) , (9)
which at saturation can be expressed by

Q, =[Cas0,°] (molal units) . (10)
The neutral species can be considered also to
undergo partial dissociation:

CaS0,°(aq)=>Ca’"(aq) + 80,7 (aq) , (11)

in which case the equilibrium gquotient expression
is
Q,=Ica™[s0,271/Cas0,"] (molal units) .  (12)
The solubility product quotient in this study was
then defined as

0., =[ca® 50,271 0,0, . (13)
With these assumptions, the molal solubility of

calcium sulfate, s, at various ionic strengths and
temperatures can be expressed by

s [Cas0 °l+ [ca®]. (14)
Since
[ca®*] -0, /150,%7] (15)
and
[$0,27] = total sulfate — {CaSO,°]
=[s0,],, ~ [Cas0,°},  (16)
then
s =[Cas0,%1+ Q_ /(50,1 ,, ~ [CaSO %) . (A7)

A plot of values of s vs 1/([3()4](0 — [CaSOf]),
with the assumption of constancy (or near con-
stancy) of activity coefficients at constant ionic
strength, should yield a straight line of slope 94,
and intercept [CaSO4°]. The results obtained at
25°C at constant ionic strengths of 0.25, 0.5, 1,
2, and 6 m, at 150°C at 0.25, 0.5, 1, and 6 m, at
250°C at 0.25, 0.5, 1, and 6 m, and at 350°C at 1
and 6 m were treated by a method of least sguares
to yield values for Q , Qsp, and [CaSO4°] for each
set of measurementis. As examples, data at 25°C,
and a representative series at [ = 1 from 150 to
350°C, so {reated are shown in Figs. 6.6 and 6.7,
the solubilities are observed to be a linear function
of 1/([304](0 — [CaSO‘tO}) within the precision of
the measurements.

69

ORNL-DWG 67532
20 e

| A= VALUE iN PURE NaNO
(NO Na,$S0, PRESENT)
[SO4 ]=TOTAL SULFATE |
~[cas0
= 25°C

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- =05 _fl,,fl

et | e
» ./fl"' i’
./"'"'—’“ L=0.25 |
| gt i
|
|

20 25 30 35 40 45 50

/150,577 (m)

Fig. 6.6, Solubility at 25°C of Calcium Sulfate Di-
hydrate in Sodium Sulfate~Sodium Hitrote Solutions at
Several lonic Strengths,

ORNL-DWG 67--53%3%

 

 

 

 

 

 

 

 

 

SCLUBILITY OF CaS3, x 103 (m)

 

 

 

 

 

 

 

 

 

 

 

1/ (5027 tm

Fig. 6.7.

peratures in Sodium Sulfate~Sodium Nitrate Solutions at

Solubility of Calcium Sulfate at High Tem-

a Constant lonic Strength of 1 (Representative also of
Results at Qther lonic Strengths Stated in Text).
In order to determine whether the above linear
relationships were an artifact of Harned’s rule,
values of the logarithm of the analytical solubility
product, m{total calcium)mitotal sulfate), were
plotted against the molality of added sulfate at
constant ionic strength. Harned’s rule would be
expected to yield a linear relationship if the plot
were made at constant molality (which for 1-1
electrolyte would be
strength). The plots were not linear nor were they

linear by estimating the nature of the plots at

mixturas constant ionic

constant molality.

The values of Qd, Qsp, and Qu obtained at the
several coastant ionic strengths and temperatures
were extiapolated to zero ionic strengths to yield
dissociation constants for CaSO40,
solubility product constants, and Kg.

values for

When the
previously obtained '® values of Kg were corrected
for neutral CaSO40, there
agreement (within ~10%) between the two sets of
With the solubility behavior in the mixed
electrolyte system, the mean activity coefficients

was relatively good

values.

of calciuin sulfate were obtainable over essentially
the entire four-component system by the relation-
ship

0

), + & = I T Ty
(Caso 4) \/-‘jn .

calcium sulfate

(18)

where the m’s in this expression refer to the total
molality of calcium and sulfate, respectively, and
K:p is defined as the themmodynainic solubility
product constant defined by Eq. {(13) at I -: 0.
Values for the constants of CaSO, are tabulated
in Table 6.2, from which thermodynamic quantitics
were calculated. No values for dissociation con-
stants for CaSO40 have been publistied at tem-
peratures higher than 40°C. % OQur own ng value
at 25°C of 2.00 compares with a literature value
of 2.31.'7 Characteristic of the behavior of sev-
eral other 1-1, 1-2, and 2-2 electrolytes studied
at elevated temperatures in our Laboratory, !® the
dissociation constant of CaSO4° decreased mark-

W. L.. Marshall, R. Slusher,
J. Chem. Eng. Data 9, 187 (1964).

lfij. Bierrum, G. Schwarzenbach, and L. G. Sillén
(Compilers), **Stability Constaints,?’ Part II, Inorganic
Ligands, The Chemical Society, Burlington House,
London, 1958.

17R. 1. Bell and J. H. B. George, Trans. Faraday Soc,
49, 619 (1953).

and E. V. Jones,

Table 5.2. The Megative Lagarithms of tha (" Trua®")
Solubility Product Constant (KS ), Dissociation
Constant (Kg), and KS (molality of C05040 at

I = 0) of Caicium Sulfate, 25 to 350°C
(the Dihydrate at 25°C)

 

-0 ~0 0

(OC) szp de P u
25 4.70 2.0 2.68
150 6.0 2.7 3.55
250 8.05 4.46 4.68
350 11.0 8.8 7.0

edly with rising temperature, reflecting the pre-
dominating effect of the decreasing dielectric
constant of water over that of increasing kinetic

energy.

SOLUBILITY OF Fe 0, AT ELEVATED
TEMPERATYRE!?

. H. Sweeton R. W. Ray
C. ¥. Baes, ]Jr.

This study of the solubility of Fe O, in HCI
solutions is being carried out because Fe O, is
steel -water

a coriosion product of the systems

used in pressurized-water reactors. Measurements
previously reported?? have been extended in tem-
perature and in HC! concentration.

The method used was essentially the same as
before. An HCI solution in a reservoir of Pyrex
glass was equilibrated at room temperature with

H, at a pressure of 1.0 atm. Then it was pumped

'8A. S. Quist et al., J. Pays. Chem. 70, 3714 (1966):
69, 2726 (1965); 67, 2453 (1963); J. Chem. Fng. Data 9
187 (1964); J. Phys. Chem. 70, 4028 (1966).

!

19Essentially the same material was included in a
paper entitled “‘The Solubility of Fe304 in Aqueous
Solufions at Elevated Temperatures,’ presented at the
18th Southeastern Regional Meeting of the American
Chemical Society held in T.ouisville, Ky., Oct. 2720,
1966.

2°¢. H. Sweeton, R. W. Ray, and C. F. Baes, Jr.,
Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1965,
ORNL-3913, pp. 64-65.
through a pressurized heated column of the spe-
cially prepared Fe O,. The equilibrated solution
was cooled and then passed through a cation ex-
changer to strip out the dissolved iron. The iron
was later eluted and analyzed spectrochemically
by means of o-phenanthroline.

After obtaining most of the data reported here,
we modified the system for eventual use of alkaline
solutions. A copper reservoir was substituted for
the glass one, and provision was made for injecting
an HCI solution into the stream of hot equilibrated
solution to prevent precipitation of dissolved iron
on cooling.

The data are shown in Fig. 6.8, where the log-
arithm of the iron concentration is plotted against
the C17 concentration obtained from the electrolytic
conductivity of the makeup solution. The lines on
the graph have been calculated on the basis of
iron being in solution as the two ions, Fe?" and
FeOH", formed in these equilibria:

Ly N + 1
7JFe 0 (s) + 2H" + L H (2)

 

 

   
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

+
=Fe®™ + % H 000, (19
Y Fe,0,(s) + H' + YLH (8)
3 3774 3702
= FeOH" + LH,O(D),  (20)
12 ORNL-DWG  67--534
t.
. s
- A
&
10 """"""" .a//" """""""
:
Tog | 200%C/ A c__ Lo\
2 It . 300 °C ‘
® A
[T 0 / /
5 Q6 Lo A e —
5
2
- '
T —
= A
Q
ool /
B oz Seden Ay L L - ]
]
J |/
®
O g
¢/
- A
-0.2 ® i . i [
Q 10 20 30 40 50 50 70

CONCENTRATION Cl {pm)

Fig. 6.8. Measured Solubility of Fe304 in HCI Solu-
tions Saturated gt 25°C with 1 atm H2.

with the corresponding solubility products:

2t
_ _h[gfm_._.].__. ~1 g =173
F 2 ? '
° [HT1PPL?
2
and
- +
[FeOH™] st/

FeOH [H+]Pé[/23
Using trial values of these products at each tem-
perature, the known Cl~ concentration for each
point, and the known ionization constant of HQO, 21
we calculated the H' concentration that gave a
charge neutrality for the ionic species at the ex-
periment temperature. The logarithm of the corre-
sponding concentration of iron was then compared
with the experimental value. After treating all the
data this way, a least-squares procedure was used
to adjust the solubility products to values that
minimized the differences between calculated and
observed iron concentrations. These constants,
which include the effective: I, pressute as cal-
culated from the solubility data of Gilpatrick and

Stone, 22 are given below:

 

€
Temperature (7C) KFe KFeOH
200 1.79 ($0.10) % 10° 0.2
260 0.0037 (£0.0045) X 10°  0.53 (£0.05)
300 0.0102 (£0.0015) x 16°  0.072 (#0,026)

These calculated solubility products indicate that
the fraction of iron in the Fe?" form was 100% at
200°% 15% or less at 260° and 55 to 85% at 3007
The apparent change in the signs of the tempera-
ture coefficients of the solubility products over
this temperature range is of interest, as this is
in the same temperature range where the dissocia-
tion constant of water goes through a maximum.

In future work we will extend the H' concentra-
tion range by using alkaline solutions, for which

 

12 12

2lhe values used were 5.01 % 10712, 6.76 x 107
and 6.45 X 1072 ™) at 200, 260, and 300°C respec-
tively; these values sare based on the data of A. A.
Noyes, Yopgoro Kato, and R. B. Soesman in J. Am. Chem.
Soc. 32, 159 (1910).

22; . O. Gilpatrick and H. . Stone, Reactor Chem.
Div. Ann. Progr. Rept. Jan. 31, 1961, ORNL-3127,
pp. 60--61 and Reactor Chem. Div. Ann. Progr. Rept.
Jan. 31, 1962, ORNL-3262, pp. 64--65.
the present system is adapted. Such data should
increase the precision of the calculated solubility
products and their temperature coefficients, and
perhaps indicate the presence of other complexes
of iron. We will also go to lower temperatures if
we are able to achieve equilibrium at practical
flow rates.

HYDROLYSIS OF BERYLLIUM IO IN 1.0 ¥
CHLORIDE AT 25°C

R. . Mesmer C. F. Baes, ]Jr.

The aqueous chemistry of Be(ll) solutions is not
yet satisfactorily defined. been a
great number of studies on the hydiolysis of beryl-

There have

lium beginning with the work of Prytz?? in 1929.
Probably the most authorative studies are those
coming from Sillén’s sclicol in Stockholm since
1956. %4727 These workers favor a scheme of three
species to explain potentiometric data below 1 M
beryllium concentration in acidic perchlorate
media, that is, Bez(OH)3+, BeS(OH)33+, and
Be(OH’)Q. many other
schemes have been proposed which were not sup-
ported by the work of Sillén and his co-workers.

The neutral species Be(OH), is the least accept-
able of those in the above scheme. The existence
of such a species implies a lower limit for the
solubility of beryllium hydioxide.
be 10~* M based upon hydrolysis equilibria®?
and the solubility product.?® However, the work
of Gilbert and Garreit?® has indicated solubilities
less than this and even as low as 107 M. A
cursory examination of the data in ref. 24 indicates
that comparable or better fits to the data are ob-
tained with the scheme Be 3(OI{)24", Bes(OI'{)33+,
and Be3(OH)42Jr as well as other schemes con-
taining Bes(OH)33+. The logarithm for the sta-
bility constant for the species Be:&(OH);‘Jr is
—8.94 +0.01, compared with the value of --8.66 %
0.01 reported by Sillén?? in 3 M perchlorate.

As pait of a program to reinvestigate the hydrol-

Over the years a great

This limit would

ysis behavior of beryllium ion, potentiometric mea-

72

surements have been completed at 25°C in the pH
region 2 to 7 and at beryllium concentrations be-
tween 0.001 and 0.05 m using quinhydrone and
calomel electrodes. The factors limiting the pH or
hydroxide-to-metal concentration ratios which can

be attained are the equilibration rate and the
solubility. Generally, equilibrium ratios up to
about 1.1 were achieved without precipitation

within 1 hr.

Analysis of our data at 25° indicates that no
pair of species with up to five metal jons or hy-
droxide ions per species is sufficient to explain
the data within experimental error. Least-squares
calculations for 33 pairs led to this conclusion.
The schemes involving three or more species have
not yet been examined in detail. However, pre-
liminary results indicate that a polynuclear species
is more suitable than Be(OH),.

As Kakihana and Sillén?** have pointed out, the
presence of the species BG(DH)2 leads to an inter-
section of all A vs pH curves at A of 1.0 and
pH ca. 5.5. Cleatly such an intersection is absent
in the data shown in ig. 6.9, This observation
is consistent with data reported in 1965 by Bertin
et al., %% although these investigators also chose
to interpret their data in terms of Be(OH)2 as
suggested earlier by Sillén.

In order to better define the most representative
hydrolysis schemeg, similar studies are being con-
ducted at higher temperatures.

 

23M. Prytz, Z. Anorg. Allgem. Chem. 180, 355 (1929).

2411, Kakihana and L. G. Sillén, Acta Chem. Scand.
10, 985 (1956).

253, Carell and A. Olin, Acta Chen. Scand. 15, 1875
(1961).

265 Carell and A. Olin, Acta Chem. Scand. 16, 2357
(1962).

273 Hietanen and L. G. Sillén, Acta Chem. Scand. 18,
1015 (1964).

28%. A. Gilbert and A. B. Garrett, J. Am. Chem. Soc.
78, 5501 (1956).

29F. Bertin, G. Thomas, and J. Merlin, Compt. Rend.
260, 1670 (1265).
73

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ORNL~ DWG 87— 535
TR e 7] ]
X X A o ' L |
M » 3 P
; NG AN L ‘ ,,,,,,,,,,,,,,,,,,,,,,, L]
E 1.9 & \A\ \\ . | ;
- \ \,: Il\\\\ i
1 A \ ™™ | ;
™ i
& oo N\ ™ \\x TOTAL Be {II) CONGENTRATION ;
fi,J ’ \ (AT START AND END OF TITRATION)
=
o
© o) {0.0464 - 0.0323) m
= 0.8 e (00221 - 0.01769)m
% A (000732 —0.00544)m
5 . (0.00270--0.00193)m
07 - B T -
m i o
i g
Lot \ P
% 06 b A
@ B
g n
" 05 ........... - : ,,,,,,, ]
2 :
> ' :
o : :
i . :
Soa bl X A R DL :
e i i
W | ‘
o .3 el ORI
i e
@ o
2 o
2 H
TODR2 e e W b A Rt N e ‘—
] :
w0
<1 i
o :
4 o \ ................. i
T N |
ol N : i
h"'o.:q\ 3 ‘ !
o I I A L, i
sx1077 4 ® 5 0 2 5 1072
—log A4
Fig. 6.9. Average Number of Hydroxide lons per Beryllium Complexed as a Function of ~log h as Determined

by Potentiometric Titration with NaOH.
7. Interaction of Water with Particulate Solids

SURFACE CHEMISTRY OF THORIA

C. H. Secoy
Heots of Immersion and Adsorption

E. L. Fuller, Jr. H. . Holmes
S. A. Taylor!

Heats of immersion of samples of thoria in water
at 25.0°C have been measured for various pre-
treatments. The base material for this series was
obtained by a 650°C calcination of thorium oxalate

(ORNL lot No. DT 102W) to produce thoria with
a specific sutface area of 35.5 m?/g. Portions
of this material were calcined for 4 hr at higher
temperatures to successively decrease the specific
surface area. The specific surface areas () and
crystallite dimensions {from x-ray line-broadening
data) are given as a function of calcining tempera-
ture in Fig. 7.1. This graph also shows the vari-
ations of the heats of immersion with regpect to

 

lprofessor of Chemistry, Centenary College, Shreve-
port, I.a.

ORNL-DWG 67- 536

 

1200 (==

1000

800

600 —

TOTAL HEAT OF IMMERSION (ergs/cm? )

400

 

 

 

 

 

 

 

 

 

 

QUT-GASSING TEMPERATURE (°C)

 

 

200 i OO S SO i
98.5 235 655 >> 2500
CRYSTALLITE SIZE (A)
N I o 0
800 1000 1200 1400 1600

CALCINING TEMPERATURE (°C)

Fig. 7.1. Heats of Immersion of Thortia in Water.
these parameters for outgassing temperatures rang-
ing from 25 to 500°C. The samples were maio-
tained at a pressure less than 107% torr for 24 hr
at each temperature, prior to sealing under vacuum.
Fach point represents the mean of at least two
experimental determinations. The calorimetric and
associated techniques have been described previ-
ously.

As observed earlier,? the materials from the
lower-temperature calcination liberated a portion
of the heat slowly after an initial instantaneous
burst of heat. The slow heat effects for the
650 and 800°C calcined thoria are best character-
ized as two concurrent first-order processes: a
slow heat effect with a half-life of ca. 4 min and
a very slow heat with a half-life of ca. 40 min.
The 1000°C calcined thoria exhibited only one
slow process (half-life of ca. 12 min) in addition
to the instantaneous process. The higher cal-
cination temperatures (1200 to 1600°C) produce
materials which release all the immersional heat
instantaneously, with no detectable slow processes
present,

The heat of immersion at any given outgassing
temperature shows an increase with increasing
crystallite size (decreasing specific surface area)
for the lowest-fired materials, followed by a de-
crease to an essentially constant value for the
larger crystallites. The variation of the heats of
immersion with respect to crystallite size is not
well understood and has been the object of inves-
tigations with other oxides, as reviewed recently
by Zettlemoyer.? The invariance of the immer-
sional heat per unit area for our larger crystallites
suggests that these values may be truly repre-
sentative of an ““ideal’’ thoria surface, with some
sort of perturbations present for the smaller crys-
tallites. Such a conclusion must receive justifi-
cation from other techniques that will allow con-
stuction of a creditable model. The inversion
of the data for the 300°C-outgassed, 1000°C-
calcined material is probably due to experimental
error, with the heat of immersion being 850 * 20
ergs/cm? for 300 to 500°C outgassing of this
material.

In an attempt to see if the aforementioned be-
havior is the result of the oxalate preparation of

 

2. F. Holmes and C. H. Secoy, J. Phys. Chem. 69,
151 (1965).

3A. C. Zettlemoyer, R. D. Igenyar, and Peter Scheidt,
J. Colloid interface Sci. 22, 172 (1966}.

75

thoria, a similar series was checked for a thoria
prepared by the steam denitration of thorium nitrate.
The low-fired, high-surface-area materials also
showed slow heat effects, with none present for
the higher-fired materials.

Adsorption of Water and Nitrogen on Porous
and Nonporous Thoria

H. F. Holmes {. L. Fuller, Jr.

Previous calorimetric?'* and gravimetric® inves-
tigations of the interaction of water with the sur-
face of thoria have revealed the very complex
nature of this process. Material used in the pre-
vious work consisted entirely of porous thoria
derived from the oxalate. ['wo samples of compa-
rable surface area were used in the present work.
Sample C is a typical oxalate material, while
sampie S was prepared in an attempt to minimize
the internal surface, that is, '
rosity.fi’7

decrease the po-

The porous and nonporous character of samples
C and 8, respectively, are clearly revealed from
the typical adsorption isotherms shown in Figs.
7.2 and 7.3. Inflections in the desorption branch
for sample C at a relative pressure of about 0.35
indicate pore sizes ranging down to about 10 A in
radins, The nonporous character of sample § is
further confirmed by the fact that the x-ray crys-
tallite size is compatible with the nitrogen sur-
face area. However, both samples exhibited the
low-pressure hysteresis and irreversible retention
of water discussed previously.® The quantities
of irreversibly adsorbed water listed in Table 7.1
are much greater than would be required to form
a single layer of surface hydroxyl groups. On the
basis of the present resulis, it is concluded that
the irreversible retention of water in such large
quantities is not a unique property of porous mate-
rial. The phenomenon has been described’® as an
immobile association adsorption to give the surface
analog of a hydrated bulk hydroxide.

 

411. F. Holmes, E. L. Fuller, Jr., and C. H. Becoy,
. Phys. Chem. 70, 436 (1966).

*E. L. Fuller, Jr., H. F. Holmes, and C. H. Secoy,
. Phys. Chem. 70, 1633 (1966).

6Sample S was kindly supplied by F. H. Sweeton.

7F. H. Sweeton, Reacfor Chem. Div. Ann. Progr. Rept.
Jan. 31, 1961, ORNL-3127, p. 71.
ORNL-DWG 67-537

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5 e i e
© ADSORPTION 5 ’ . :l
* DESORPTION ‘ .
. | : " c/ c/
z i i o /0/
i : A A
3 i Fo -° o
Lj ” | ./ O\/O
a | oA < /
= | /./ T .
3 | A/
. ~ ..
o | e* S/
= ia’. A 4
= ot —° /
0 K | 0 *
= A o° nd
v 32 ! o 4
m .. i d_/o i /./ I o
é l. \“‘. 'Lo/. o/
<L 1 : _Q/:‘-‘.". : O/
& ¢ -~
N | o § Iy O
O JENDI o / 13
= ’/ /0’0 /. o= 0
. .
o Ze o -t O/‘T/')
i Q’O | ./fi 0/0/
Q 0 | .fi’ /D/
o
I / ,41
L S
S st -
E’ ¥4 Oood‘ 3
o J g 1 T AFTER OUTGASSING l .
g %0 SAMPLE WEIGHT: 500.052 mg
o g ! ‘ [l AFTER PREVIOUS Hy0 ISOTHERMS
¢ SAMPLE WEIGHT: 501.134 mg
o ! [
i i |
$-=-END II ‘ |
0 l N ‘ J ,,,,,, :
O 0.1 0.2 0.3 0.4 0.5 0.6 o7
PPy
Fig. 7.2. Adsorption of Water on Somple C at 25.00°.
ORNL-DWG 67-538
G oo :
— ! I
T ! i
0 : '
g o ADSORPTION S
" » DESORPTION &
/-
o * 0
= .)’ /
a MNP
3 - .M o
o A o]
£ o o
: S
o
; /./’. /O/ ./0;/0
e O rd
w - ~ 50
> -« ° o
% " ~ o
. e o o
a 2 art o e T
o P o /-/0
Q ® o O ;
= o /0/ ./*b/
- _.,o : O/o ’,./o/
5 | e T
o f o e
~ e c° N[')‘\ e i
QS EVLSTT f
£ W;_T”__fl I AFTER OUTGASSING | .
s B | ! SAMPLE WE!GHT: 199.929 mg
g i |
o %fi’ I AFTER PREVIOUS H,0 ISOTHERMS
- ; I ; SAMPLE |WE|GHT 200.142 mg
& ' ! i
IS ND ! | ‘
_—ENC U | ‘ : .
ol | - ]
0 0.1 0.2 0.3 0.4 0.5 0.6 C.7
/R

Fig. 7.3. Adsorption of Water on Sample S at 25.00°.

76

Table 7.1.

Summary of Nitrogen

Adsorption Measurements

 

 

Micrograms
Sample of HQO N2
Sample Weight? per Square Surfa(ie BET C
) Meter ) (ArE? Constant
of Surface m "/ ¢)
C 499.810 4 5.52 1050
C 500.276 174 4,78 152
C 501.264 533 4,40 40
C 499,803 2 5.48 1270
8 199.815 56 5.96 490
S 200.012 222 5.46 85
s 200.108 302 5.50 32
S 200.170 354 5.25 72
S 200.338 496 5.29 53
S 196.813 55 5.96 390
aSampl weight in vacuum at start of experiment.

Chemlsorhpd H,O in excess of sample weight in
vacuum at 500°C." Computed u%mg N, surface area
measured after cutgassing at 500°C.

“Assumes that an adsorbed nitrogen molecule occupies
16,2 A-.

Application of the BET theory to the water iso-
therms yields interesting information concerning
the apparent surface areas. The data for sample C
indicate that the effective surface area decreases
with increasing amounts of irreversibly adsorbed
water., Ultimately the effective surface area de-
creased to about 75% of the area determined by
nitrogen after outgassing at 500°C. This is not
since chemisorbed water should de-
However,
the effective

une xpected,
crease the total pore volume,
case of the nonporous sample S,
surface arca decreased to about 40% of the original
nitrogen value. In opposition to this, the amount
of water adsorbed at the higher relative pressures
indicated an effective surface area roughly equal
Surfac:
areas were calculated from the water data on the
assumption of a liquid-like monolayer, that
each molecule occupies 10.6 A?  Tentatively,
the anomaly with sample S can be explained by
assuming that each water molecule in the first
physically adsorbed layer is hydrogen bonded to
two underlying chemisorbed water molecules. Cal-
culations based on this assumption give surface

in the

to the original nitrogen surface area.

is,
areas which are much more compatible with the
nitrogen surface areas.

Results from the nitrogen adsorption experiments
are summarized in Table 7.1, which gives the
BET parameters obtained from the isotherms. It
is evident that increasing amounts of irreversibly
adsorbed water decrease both the measured nitro-
gen surface area and the BET C constant., A
decreasing value of C implies a smaller average
heat of adsorption in the first monolayer, a less
shatp ‘‘knee’’ of the isotherm, and that higher
pressures are required to fill the first statistical
monolayer. Decreasing nitrogen sutface areas
for sample C, like those obtained from the water,
can perthaps be explained on the basis of the chem-
isotbed water decreasing the pore volume. This
concept cannot, however, explain the decreasing
nitropen surface areas observed with sample S.
Explanations for the effect of chemisorbed water
on the nitrogen surface areas of sample S are, at
best, tentative. The effect of the magnitude of
the BET C constant and of the substrate lattice
on the area occupied by a nitrogen molecule is
curtently being investigated.

Infrared Spectra of Adsorbed Species on Thoria

C. S. Shoup, Jr.

Infrared spectral studies of adsorbed species on
thorium oxide have continued with higher outgas-
sing temperatures than were previously attainable.?
Initial room-temperature outgassing of a pressed
disk of ThO, which had been calcined in air at
650 (see Fig. 7.1) produced a prominent absorp-
tion band at 3740 cm ™!, attributed to unperturbed
surface hydroxyl groups, and a band at 3660 cm™ .
In addition, a dramatic reduction in the intensity
and increase in the frequency at maximum absorb-
ance of a broad band due to the stretching mode
of hydrogen-bonded OH groups occurred. In con-
trast to most other oxides, ?'1? however, increasing
the outgassing temperature above 100° reduced
{he intensity of the band at 3740 cm™'. After
outgassing at 5007 this band had virtually dis-

 

BC_ H. Secoy and C. 8. Shoup, Jr., Reactor Chem. Div,
Ann. Progr. Rept. Dec. 31, 1965, QORNL-3913, pp. 7071,

°M. R. Basila, J. Phyvs. Chem. 66, 2223 (1962).

Wy B. Lewis and G. D. Parfitt, Trans. Faraday Soc.
62, 204 (1966).

77

appeared, leaving a doublet at 3645 to 3660 cm ™1

and a weaker band at 3520 cm~' superimposed
on a weak, broad contour, indicating some residual
polymeric hydrogen bonding.

Although this thoria had previously been calcined
in air at only 650° the sample was nevertheless
outgassed at increasingly higher temperatures,
thus producing some sintering action.!! OQutgas-
sing at 650° left only weak bands in the OH stretch-
ing region at 3656 and 3503 cm™!. No surface
hydroxyl groups were spectroscopically evident
after outpassing at 800° but other bands in the
1200 to 1700 em™ ! region were present. These
latter bands disappeared after outgassing at 950°.
However, bands at 1040 and about 730 cm ™! were
very little affected by the outgassing treatment,
and their intensities were hardly reduced after
in vacuo sintering at 950°. Examination of other
samples of 'I‘hO2 indicated that these bands were
properties of the total mass, either as thotium
oxide or bulk impurities.

The discoloration of ThO,, which has been ob-
served often after outgassing at 500° % has been
shown to be due to the presence of a small amount
of carbon residue from organic contamination de-
posited during the initial outgassing steps (indi-
cated by the presence of carbon-hydrogen stretch-
ing bands in the 2850 to 3000 cm™?! region). This
catbon residue can be removed by oxidation at
400° with an accompanying weight loss, or by
volatilization at 900 to 1000°.° [o the absence
of prior organic contamination, heating to 500° in
vacuo produced no discoloration. The fact that
the infrared spectra and the enerpetics of water
adsorption? were independent of color and oxygen
treatment (in contrast to rutile*?) indicates  that
oxygen defects, as proposed elsewhere,'? do not
play an important role in the discoloration of thoria.

In the presence of pure water vapor at a relative
humidity of 60%, the H-O~H deformation mode of
physically adsortbed H,0 at 1630 em™! was the
only absorption band to appear (except in the OH
stretching region) that was not present after out-
gassing at 950°  After rcom-temperature evacu-
ation, a brief exposure of the thoria to the labora-
toty atmosphere at 40% relative humidity produced

 

after 24 hr each at 630, 800, aand 950° in vacuo, the
specific surface area was reduced to 12.3 mz/g and the
average crystallite size was increased to 319 A,

12M. E. Wadsworth et al., **The Surface Chemistry of
Thoria,’’ Progress Report, University of Utah, Salt Lake
City, Utah (Jan. 31. 1939).
several absorption bands. Even after evacuating
the cell for 24 hr, strong and fairly sharp bands
remained at 861, 1020, 1303, 1416, and 1583 cm ™!,
in addition to those present before the ThO 6 was
exposed to the atmosphere. It is apparent from
this that atmospheric catbon dioxide is rapidly and
strongly adsorbed on thoria, at least in the pres-
ence of water vapor.

When water vapor was adsorbed on ThO, which
had been outgassed at 950°, sharp bands at 3740,
3695, and 3668 cm~ ! appeared with constant rel-
ative intensities up to approximately one mono-
layer coverage (10 to 12 doses of H,0), as shown
in Fig. 7.4. With increasing coverage, the 3668-

ORNL-DWG 66-12569

TRANSMISSION ——

 

 

 

b

4000 3800 3400

3600 3200 3000
FREQUENCY {crn™')
Fig. 7.4. Infrared Spectra of the OH Stretching Region

After Adsorptien of the
Volumz Doses of H,0 Vaper on ThQ,.
placed slightly for clarity.

lndicoted Mumber of Egual-

Ordinates dis-

78

cm ™! band continued to increase in intensity, but
the intensities of the 3740- and 3695-cm™' bands
decreased. In addition, a broad band at 1630
cm ' began to appear, proving the adsorption of
molecular water. Nevertheless, the growth of the
broad band in the OH stretching regicn showed that
polymeric hydrogen bonding was present at sig-
nificantly less than one monolayer coverage.

The nonequilibrium nature of water adsorption
was further demonstrated by successive adsorption-
desorption cycles between 4.58 torrs (for 1 to 150
hr) and <10~ ° torr (for 24 hr).
in vacuo spectium showed a decrease in the inten-
sity of the 3740-cm~ ' band and an increase in
the intensity of the broad band around 3440 cm ™ .
In addition, bands at 1564.5 (strong), 1430, 1375
(sharp), and 1362 cm~ ' (sharp) appeared. These
bands increased in intensity with each successive

FEach successive

adsorption-desorption cycle, although their relative
intensities remained unchanged. These bands
appear to be due to the adsorbed water rather than
to any contamination, but a firm interpretation of
their nature awaits further experimental data.

BEHAVIOR OF GASES WITH SOL-GEL
URANIUM.THORIUM DXIDE FUELS

D. N. Hess
H. F. McDuffie

B. A. Soldano
C. F. Weaver

Sol-gel microspheres of ThO, or UO, have been
found to evolve gases when heated, as did the
thoria--3% UO, sol-gel materials previously re-
ported. 1*1*  Efforts have been made to remove
these gases, as well as the carbon, which are
generated by interaction and pyrolysis of the water,
nitrates, organic solvents, and surfactants included
in the sol-ge! materials during their preparation.
Such removal is considered desirahle becanse ex-
cess gas pressure or reactions between the gas
and metal might occur in sealed fuel elements,
causing rupture during reactor operation.

The air-dried ThO, microspheres yielded CO,,
CO, H:z’ NO, Nz’ and organics upon heating in

vacuum. The first three were dominant. The

 

ISD. N. Hess, W. T. Rainey, and B. A, Scldano, Reac-
for Chem. Div. Ann. Progr. Rept., Jan. 31, 1965, ORNL-
3789, p. 177.

145, N. Hess and H. A. Soldano, Reactor Chem. Div.
Ann. Progr. Rept. Dec. 31, 1965, ORNL-3913, p. 72.
largest amount of gas evolution occurred at the
following centigrade temperature intervals: 240
to 260, 400 to 460, and 700 to 760, Above 760°C,
only a negligible amount of gas remained.

The wet UO, microspheres yielded CO,, CO,
Hz’ N,, 02, NO, and organics when heated with
steam and evacuated. The temperature intervals
of maximum gas evolution appear to be 150 to
250°C and 400 to 650°C. The principal gas
evolved (primarily in the 300 to 350° range) was
CH,. This is in direct contrast with the production
of higher-molecular-weight organic products previ-
ously noted in the case of the ThO, microspheres.

A possible explanation'® for the difference is
that in dry ThO, catalytic cracking of the included
organics occurs, while in the wet UO, matrix
thermal cracking is dominant. The lailter case is
expected to produce lower-molecular-weight prod-
ucts.

A conditioning scheme, successful with respect
to producing a low carbon content, a low O:U
ratio, and high density, is shown in detail in Table
7.2. The flow of water vapor was necessary for
the removal of carbonaceous material and, thus,
for production of a low final concentration of car-
bon in the UO, microspheres. The water vapor
also seems to prevent fragmentation of the spheres.

The mixture of CO, and H, 0 was superior to
either compound separately for increasing the rate
of sintering. Tt was of prime importance that the
carbon removal be complete before the mixture of
Co, and H?_O was added, since the densification
step trapped any remaining carbonaceous material.
The final use of pure H, to counter the oxidizing
effect of the mixture of CO, and H O produced the
low O :17 ratios.

It is expected that the same processing scheme
would be applicable to microspheres consisting of
any member of the ThO,-UO, solid-solution sys-
tem, although no experimental information is avail-

 

1SP. H. Emmett, personal communication.

79

able for such materials. For pure ThO,, the H,
gas in the processing scheme is both unnecessary

and hamless.

Table 7.2. Successful Gas Flow Conditioning Scheme
for Sol-Gel UO, Microspheres

Batch p-9-12-1153; weight: 9.976 ¢

 

Treatment Schedule

 

 

 

 

16 hre 170°C Ar—H O
2 250 Ar--4% szHZO
2 350 Ar—-4% H2--—I{20
2 450 Ar-4% H2—~H20
Cool Ar—4% Hz-»HzO
156 br Store Helium
25 — 550 Ar—4% P12-~~112O
2 550 Ar 47 HQ-«H2O
2 650 Ar—A4% HZ—H2O
2 750 Ar—4% HQ*HZO
Cool Ar—4%; I-I2»~E{20
Store Helium
25 — 8350 Ar—4% HQ»-HzO
2 A0
1 7, hr 8§50 Ar—4% Hzmflz()
2 I
2 850 C02 1120
1
A B850 H,
3 -
A 1000 H,
Physical and Analytical Data
Weight Percent  Density
Loss o/u C at 210 psi Appearance
15%  2.001 0.008 10.82 Shiny black;

no fines;

nonuniform size

 
Part Il

Gas-Cooled Reactors

 
8. Diffusion Processes

TRANSPORT PROPERTIES OF GASES

Gaseous Diffusion Studies in Noble-Gas Systems

A. P. Malinauskas

Gaseous diffusion experiments have been con-
ducted with the gas pairs He-Kr, Ar-Kr, and Kr-Xe
over the temperature range 0 to 120°C to provide
additional data for use in an investigation of the in-
teraction characteristics of noble-gas molecules, !
This work represents the second phase of a sys-
tematic study of the binary diffusion process in
systems involving all possible combinations of
the noble gases.? The third part of this research,
involving the systems He-Ne, Ne-Ar, Ne-Kr, and
Ne-Xe, is in progress.

All the diffusion data available concerning the
systems He-Kr, Ar-Kr, and Kr-Xe are graphically
summarized in Figs. 8.1 to 8.3, where the dif-
fusion coefficient D.,, at 1 atm pressure, is
plotted as a function of the absolute temperature
T. The solid lines in the figures have been con-
structed using the usual Chapman-Enskog ex-
pression for the diffusion coefficient,® in which
the Lennard-Jones (12-6) potential energy pa-
rameters presented in Table 8.1 have been em-
ployed. These parameters yield the best repre-
gsentation of the experimental data; they likewise
give favorable agreement with diffusion coefficient
values deduced from measurements of the com-
position dependence of the viscosities of the cor-
regponding pas mixtures. On the other hand, a
similar comparison with values obtained from the
measured thermal conductivities of the mixtures
iz not as good, and the deviations appear to

 

1A. P. Malinauskas, J. Chem. Phys. 45, 4074 (1966).
2a. P. Malinauskas, J. Chem. Phys. 42, 156 (1965).

3]. O. Hirschfelder, . F. Curtiss, and R. B. Bird,
Molecular Theory of Gases and Liquids, chap. 8, Wiley,
New York, 1954,

83

worsen as the mass difference of the gas pair
increases.! Unfortunately, it is not possible to
pursue this aspect further,
thermal

since experimental
conductivity data are neither plentiful

ORNL-DWG £6-4900

 

 

0.9

 

 

0.8

 

 

Dip lem?/sact

 

 

0.6 |eenee

 

 

 

 

 

 

 

 

 

 

 

 

0.5
250

300 350

 

400

Fig. 8.1,

ficient of the System He-Kr at 1 otm Pressure.

Experimental Yalues of the Diffusion Coef-
Solid
line: Lennard-Jones {12-6) potential. BH. Watts, Trans.
Faraday Soc. 60, 1745 (1964). & B. M. Srivastava and
R. Paul, Physica 28, 646 (1952). # L. Durbin and R.
Kobayashi, J. Chem. Phys. 37, 1643 (1962). A K. P.
Srivastava and A. K. Barua, Ind. J. Phys. 33, 229 (1959).
O this work.
 

 

D,» {cm?2/sec)

 

 

 

 

 

200 300 400 500
7 {°K])

Fig. B.2,

ficient of the System Ar-Kr ot 1 atm Pressure.

Experimental Values of the Diffusion Ceef.-
Solid
line: l.ennard-Jones (12-6) potential. & M. Watts, Trons.
Faraday Soc. 60, 1745 (1964). a R. Paul, Ind. J. Phys.
36, 464 (1962}, & L., Durbin and R, Kobayashi, J. Chem.
Phys. 37, 1643 (1962). O K. Schafer and K, Schuhmann.
Z. Elektrochem. 61, 246 (1957). /\ B. N. Srivastava and
K. P. Srivastava, J. Chem. Phys. 30, 984 (1959). O

this work.

 

Oyn (cm?/sec)

Q.08

0.06 |-

 

 

 

Fig. 8.3.

ficient of the System Kr-Xe at 1 atm Pressure,

Experimental Values of the Diffusion Coetf-
Solid
line: Lennard-Jones (12-6) potential. & H. Watts, Trans.
Faraday Soc. 60, 1745 (1964). O this work.

Tabie 8.1.

Parameters Obtained from the Diffusion Studies

Lennard-Jones (12-6) Potential Energy

 

 

Gas Pair T (A) €/k (°K)
He-Kr 3.071 58.8
Ar-Kr 3.600 121
Kr-Xe 3.923 170

 

nor sufficiently accurate for other than a super-
ficial analysis.

Thermal Transpiration

8. A. Cameron* A. P. Malinauskas

We had demonsirated earlier that thermal tran-
spiration data may be employved to obtain infor-
interchange of energy as-
sociated with translational and internal molecular
5 Experimentally, this entails

mation regarding the

motion by collision.
the careful measurement of small pressure drops,
under steady-state
imposing a temperature gradient across a gas
confined in a capillary.

conditions, which result by

In the previcus work, fourteea 0.1-mum-ID cap-
illaries had been employed, and the method used
was satisfactory from all aspects except one: &
hr were required to obtain a single datum point.
In an attempt to remove this shortcoming, we
sought to replace the capillaries with a fritted
glass disk.

Experiments wete conducted with the porous
disk arrangement; although the time interval for
a given experiment was reduced to 5 min, it was
no longer possible either to measure or to comn-
trol the temperature gradient accurately, since
the use of the porous plate permitted the thermal
conductivity of the gas to markedly atfect the
temperature conditions. Coansequently, the re-
producibility of the data obtained was quite poor,
and the method was therefore abandoned.

In the meantime, a more sensitive pressure-
sensing device than that employed in the earlier
work has been procured. Since the instrument
will permit us to employ capillaries of a larger
size, thereby increasing the rate of attainment
of steady-state conditions, we are cuirently con-
structing a modified version of the original cap-
illary design.

 

*Summer participant, Hanover College.
SA. P. Malinauskas, J. Chem. Phys. 44, 1196 (1966).
Gaseous Diffusion in Porous Media

A. P. Malinauskas R. B. Evans III
E. A. Mason® '

A generalized freatment of gas transpost in porous
media has been realized; as a result, it is now
possible to account for a variety of phenomena
involving gas transport from a single viewpoint
having a sound theoretical foundation. 7

The treatment has been developed on the basis
of the ‘“‘dusty-gas’’ model, a model in which a
porous septum is described as consisting of uni-
formly distributed giant molecules (dust) held
stationary Although this description
is not as palatable as the more common ‘‘bundle
of capillaries’’ concept, it does possess two dis-
tinct advantages from a mathematical standpoint.
First, the geometrical part of the problem neatly
separates from that part which deals with the
dynamics of the gas transport. Second, the im-
portance of gas-surface collisions relative to
gas-gas readily assessed; one
mesely varies the ‘““mole fraction” of the dust.
In other words, it is unnecessary to postulate
one mechanism for free-molecule conditions and
ancther for hydrodynamic transport and then at-
tempt to reconcile the two in the transition region;
the validity of the dusty-gas model applies through-

in space.

interactions is

out. Moreover, the results are likewise applicable
to describe gas transport through capillaries simply
by a suitable substitution for geometric parameters.

The following brief summary indicates the num-
ber of phenomena which have been described as
special forms of the general case (some of these
have been presented previously).®

Binary Gaseous Diffusion at Uniform Temper-
ature and Pressure. — This pheaomenon is treated
as three-component diffusion in terms of the dusty-
gas model; the most notable result is a theoretical
formulation for the observation that the ratio of
the fluxes of the two diffusing gases is approxi-
mately inversely proportional to the square roots
of their molecular weights, not only under con-
ditions of free-molecule transport but in the region

 

 

®onsultant, University Institute for

Molecular Physics.
7. A. Mason, A. P. Malinauskas, and R. B. Evans 111,
accepted for publication in J. Chem. Phys.

SR, B. Evans III, (. M. Watson, and E. A. Mason,
J. Chem. Phys. 35, 2076 (1961); 36, 1894 (1962); E. A.
Mason, R. B. Evans I, and G. M. Watson, J. Chem.
Phys. 38, 1808 (1963); E. A. Mason and A. P. Malinaus-
kas, j. Chem. Phys. 41, 3815 (1064).

of Maryland,

85

of hydrodynamic transport as well. Parenthetically,
this observation appears to have been made first
by Graham in 1833, but has either become for-
gotton or confused with his work on effusion
(transport into a vacuum), which was done about
13 years later.

Isothermal Transport in the Presence of a Pres-
sure Gradient. — When only a single gas is con-
sideted, the mathematical expression which results
from an application of the model represents an
analogous form of Poiseuille’s flow equation with
viscous slip. However, the viscous part arises
from Stokes’ law, whereas the slip term is de-
veloped from a consideration of pure gas diffusion
in a static dust environment. Moreover, the
Knudsen minimum, which has yet to be rigorously
described utilizing the capillary concept, is sat-
isfactorily taken into account by considering the
next-higher to the diffusion co-
efficient.

Whether a pure gas or a binary gas mixture is
considered, it turns out that it is possible to
separate (mentally) the viscous flow contribution
from that due lo diffusive transpori; but in the
latter case the equations remain coupled through
the composition dependence of the characteristic

approximation

transport coefficients.  Consequently, few sit-
uations can be conveniently handled without re-
course to numerical methods. Cne of these is
known as the Kramer-Kistemaker effect, wherein
a pressure gradient develops in a closed system
because of diffusion. In this instance the equation
derived from the dusty-gas concept appears to
be the first description of the phenomenon that
is applicable at all pressures.

Pure Gas Transport Under the Combined Influ-
ence of Gradients of Pressure and Temperature., —
This phenomenon is koown as thermal transpi-
ration or the thermomolecular pressure difference.
Priot to the application of the dusty-gas model,
the effect was primarily investigated only be-
cause of its influence on zccurate pressure meas-
urement. As stated in the section ‘‘Thermal
Transpiration,”” the phenomenon now appears to
provide a particularly simple approach to an in-
vestigation of inelastic collisions.

Binary Gas Transport Under Nonuniform Tem-
perature and Pressure Conditions. - All the pre-
vious items represent special cases of this rather
complex situation, and only two aspects of the
problem have been considered in detail. The
first of these is the pressure difference, at steady-
state conditions, which is produced in a binary
gas mixture under the influence of a temperature
gradient, without regard for variations in com-
position; the resultant expression describes thermal
transpiration in a binary gas mixture.

The second case also concerns a steady-state
solution; the problem considered is the relative
separation produced in a binary gas mixture as
the result of a temperature gradient, without re-
gard for pressure variations. The expression
derived for this case describes the variation of
the thermal diffusion factor with pressure, as one
proceeds from the free-molecule region to the
region of hydrodynamic transport.

It is interesting to note that in neither of these
last two cases does there appear to be any ex-
perimental data with which the theoretical re-
lationships may be veritied.

Gas Transport Studies Relojed to Yented
Fue!l Elements for Fast Gas-Cooled Reactors

R. B. Evans III D. E. Bruins®

Tentative reference designs for a 2500-Mw (thet-
mal)} fast-flux helium-cooled reactor call for utili-
zation of fuel pins comprising (U,Pu)O2 bushings
stacked in free-standing stainless steel cladding.
Requirements for high power densities demand
very high fuel temperatures, close pin spacing,
and minimum cladding wall thickness. One might
envision 0.38:in.-OD claddings (0.010-in. walls)
containing 6-ft sections of 80% effective density
Cladding temperature will be 815°C at
steady-state conditions; helium coolant pressure
will be about 1000 psi.

If, for any reason, the coolant pressure exceeds
the internal pin pressure by 200 psi at 815°C,
the claddings might collapse. A means for pres-
sure equalization is clearly needed. In the present
investigation the possibility of using direct vent-
ing devices'!® to ensure pressure equalization
is being examined under the realization that direct

fuel.

vents pose special problems regarding fission
product release and coolant contamination.

 

Summer participant, Camegie Institute of Technology,
Pittsburgh, Pa.

1OAlternate methods for pressure equalization of fast-
reactor fuel elements are reviewed by F. R. McQuilkin
et al., GCRPF Semiann. Progr. Rept. Mar. 31, 1965, p. 169,
ORNIL.-3951. Additional descriptions and views of col-
lapsed fuel pins are presented by D. R. Cuneo et al.,
Ibid., pp. 179--90.

86

We have elected to initiate our work by giving
first consideration to the release problems and
means for their mitigation. This approach was
adopted to take advantage of the only established
criterion available for studies of this type, namely,
the maximum allowable release-to-birth ratio, R/B,
of fission products as legislated in the PBRE1!!
fuel specifications, modified to account for zero
retention by fuel at fast flux reactor tempera-
tures. !? The reference criterion turns out to be
a constant R/B of 1.8 x 10~ 2 for all fission prod-
ucts.

Venting efficiencies have been left as somewhat
of a dependent variable because, once a suitable
venting device has been contrived, dimensioned,
and evaluated, venting capabilities as well as
inherent limitations can be determined. Under
this approach we need not be concerned (for the
time being) with rather nesbulous considerations
that might govern venting criteria in the future,
for example, ““maximum credible’’ coolant pressure
and /or fuel-temperature excursions.

In summary, we seek selective venting devices,
materials, or configurations that can discriminate
gradients in total pressure and partial pressures;
thus forced-coolant flow admittances will be high
when Ap # 0, and diffusive-flow admittances will
be low when Ap = 0. Both laboratory and desk
studies have been initiated to obtain the ob-
jectives cited above.

RECOIL PHENOMENA IN GRAPHITES

R. B. Evans III J. 1.. Rutherford
R. B. Perez 13

Pyrolytic-carbon-coated (Th,U)C, and (Th,U)C,
microspheres have demonstrated adequate irradi-

 

1A, P. Fraas et al., Preliminary Design of a 10 Mw(t)
Pebble Bed Reactor Experiment, ORNL-CF-60-10-

63(Rev.) (May 1961).
12 he PBRE criterion for R/B was 1 x 1075 1/2

1/2) 7
where tl,/z’ sec, is the fission product half-lives. But
this value takes full advantage of the low R/B of the
fuel which is 5.3 x 10”5(t1/2)1/2. For fair comparisons
the PBRE value must therefore be promoted to 1.8 x 1072
(then t1/2 terms cancel) to account for a fuel R/B of
The R/B
values cited were reported by P. E. Reagan et al., Re-
actor Chem. Div. Ann. Progr. Rept. Jan. 31, 1963, ORNL-
3417, p. 213,

13Consu1tant, the University of Florida, Department
of Nuclear Engineering Sciences, Gainesville.

unity as anticipated for fast gas reactors.
ation stability 14 under design conditions of existing
high-temperature gas-cooled reactors. However,
extension of current techoology to include per-
formance estimates for advanced concepts and
optimization of coating designs!® has required
additional and supporting studies of phenomena
related to coating failure.

One mode of failure concerns spearheads, gaps,
and fractures that are initiated in recoil damage
and fragment-densified regions at coating layer
interfaces. Recent irradiation tests!® have in-
dicated, for example, that premature failure oc-
when inner-buffer layers are too thin to
sufficient

curs
provide recoil-damage protection for
cuter containment layers. Accordingly, our present
objective to determine what fraction of the
total coating thickness constitutes effective stop-
ping regions for recoiled fission fragments. Studies
of this kind should provide supporting information
for one aspect of the optimization analysis men-
tioned above.

Results obtained in this
directly from the experimental determination of
the penetration distances of fission fragments
that have recoiled into target specimens from thin-
layer source regions. These regions are formed
by placing 235U on one sutface of dense pyro-
carbon coupons using an electromagnetic-separator
technique. !” Target specimens are placed against

.
15

investigation stem

this surface, and source-target pairs are subjected
to a thermal-neutron flux to induce fission and
subsequent fragment recoil. Under this config-
uration, fragments enter the target (and source) as
nearly monoenergetic but randomly directed beams.
The targets are then ground,!® and grinding in-
crements are assayed by gamma counting so that
integral forms of the distribution curves can be
constructed.  Activity values for several frag-
ments are extracted from the overall count data
with the aid of computer program ALPHA.1® The

 

14_]'. L. Scott, J. G. Morgan, and V. A. DeCarlo, ZTrans.
Am. Nucl. Soc. § 42627 (1965).

15]. W. Prados and J. L. Scott, Trans. Am. Nucl. Soc.
8, 38788 (1965); alsc issued as ORNL-TM-1405 (March
1965).

164 R. Olsen et al., GCRP Semiann. Progr. Rept.
Mar. 31, 1966, ORNL-3951, pp. 41~-53.

17J. Tritt, &G. D, Alton, and C. M. Blood, Appl. Phys.
Letters 3(9), 15052 (1963).

18R, B. Evans II[, J. L. Rutherford, and R. B. Perez,
GCRP Semiann. Progr. Rept. Sept. 30, 19265, ORNI.-
3885, pp. 14146,

9% Schonfeld, Nucl. Instr. Methods 42, 21318 (1966).

87

integral curves yield information concerning re-
coil distances along z normal to the source plane.
They also give information concerning the frag-
ment distributions f(z) as they occur in coatings
and the distributions f(¢) as they occur either in
“r space’ or along z in collimated beam experi-
ments.

The basic range concept given in micro-
scopic terms by simple relationships between n,
the density of scattering centers in the target,
T, the specific energy loss to target electrons
and atoms; and do, the differential cross section
for such loss. The macroscopic range-energy
[R(E) — E] relationship is derived from

18

0 g
RE) dE 1 dE o
U J @ER) 0, ) sB) |
g’ £’
where
S(Ey= [ TdT (2)

is the stopping cross section per scattering cen-
ter, and

dE/dR = ~n,- S(E) (3)

is the average energy loss per unit path length.
One assumes that this average loss is a contin-
aous function and finds that (dE/dR)e for elec-
tronic contributions is similar to energy losses
in constant deceleration processes in elemerntary
mechanics. The theoretical problem reduces fo
derivation of relationships between atomic prop-
erties and 5(F), T, and do

In most treatments the range-density product is
taken to be constant, as in Eq. (1). Thus, in
principle, range values for all target deunsities
are known if this product is determined for any
one given target density. Implied restrictions are:
materials in question must be homogeneous to
satisfy the condition dn,/dR = 0, and n, must be
the same for all beams. Along these lines we
note that current local trends favor the use of
very-low-density carbons (d ™~ 0.8 to 1.2 g/em?)
for inner coatings of fuel particles. These are
in many instances quite porous. The implied
restrictions on n, will obviously be violated, and
this feature of the macroscopic theory provides
the major justification for our work.
A solution to Egs. (1 to 3), as given by Lindhard
et al.,*% is simply
Fle)=7.(e) — Alk, €)= (217 D/(k) — Alk,€). (4a)
Here the reduced quantities €, 7(<), and ¢ cor-
respond to the original variables £, R(E), and T.
Contributions of nuclear stopping are given as a
correction term A(k,€) to be applied to the elec-
tronic range r_(¢). The k is soit of an integration
constant involving the masses and changes of
Lindhard presents
values of A(k,¢€) in graphical form as well as
theoretical fluctuation values Alp(€)]? which re-
late to the straggling factors a@. These factors
are negligibly small compared with those intro-
duced by instrumentation.

fragment and target atoms.

 

At fission recoil ener-
gies, 21 A(k, €) values are such that
(e}Y~ Ce?/3and R(E)~ C'E?/3,  (4h)
p

where C and C are arbitrary constants. This is
merely Rutherford’s equation, modified forms of

88

which form the basis for a recent generalized
range-energy reported by Frank. ??
Some predicted and experimental range values are
presented in the top section of Table 8.2.
Successful correlation of penetration data de-

correlation

pends heavily upon a proper selection of an r-space
distribution function as dictated by the structure
of the target material under investigation. For
uniform structures we have assumed validity of
the normal-Laplace distribution given in Fig. 8.4.
Since r-space distributions are not measured di-
rectly in our experiments, the f(r) function must
be converted to f(z) to derive applicable equations.
The connection between these functions is ob-
tained by pretending that all source points can

 

201, Lindhard, M. Scharff, and H. E. Schiott, Kgl.
Danske Videnskab. Selskab, Mat.-Phys. Medd. 33(14),
1 (1963).

21y, M. Alexander and M. F. Gazdik, Phys. Rev. 120,
874—86 (1960).

22p. W. Frank, Bettis Technical Review, WAPD-BT-30
(April 1964), pp. 47-53.

Table 8.2. Comparison of Experimental and Predicted Receil Ronges

Applicable to ldeal and Porous Carbon Matrices

 

 

 

Farget R*d, Range -Density Product (mg,/cm2) Number of
Target Material Density - : '
(o /om® 141Xe(Cea) 14°Xe(Laa) 103, a 95Y<2ra) Experiments
{g/cm"”)
Ideal Target Materials
Predicted” for carbons 2.21 2.24 2.91 2.93
Dense pyrolytic carbon 2.192 2.54 2.50 2.94 3.01 6
(massive deposit)
Predicted® for carbons 2.63 2.70 3.54 3.59
Parous Target Materials
Isotropic pyrolytic 1.49 2.47 2.42 2.89 2.96 3
carbon (disk coating)
Impregnated graphite 1.865 2.52 2.52 3.09 3.16 3
(CGB)
Very porous graphite 1.63 2.94 3.33 3.41 3

(C-18)

 

“Species counted.

®Rased on Frank’s correlation and tahulated energies (ref. 22).

®Basad on Lindhard’s correlation (ref. 20) using energies as plotted by Alexander and Gasdik (ref. 21).
be collected at a single point corresponding to the
origin of a spherical coordinate system, where @ is
the polar angle measured from z, r=+/ p? + 22 ig
the radius vector, dw is an element of solid angle,
and p = 277r sin . Thus

f(z) = f
@

An additional integration of f(z) over z is required
to obtain the result sought, namely,

2£(r) cos 0 dw f”of(p,z) p dp ©)
477 o (p2+ z2)

FR = [THz)dz=1~z/R f(z a/R) . (6)

Zz
This approximate relationship gives F.R., the frac-
tion of activity remaining after grinding to a pen-
etration z, as a function of z in homogeneous

89

(uniform) targets. The curve is essentially a
straight line with a small tail accounted for by
the f_ term that reflects effects of straggling.
These are greatly attenuated by the integration
and can, in theory, be safely neglected. Most of
the straggling shown in our results is the result
of experimental imperfections. Experimental data
that follow the normal-distribution correlations
appear on the left side of Fig. 8.5.

An equation giving a close approximation to
f(z) in uniform particle coatings develops when
the point source pertaining to Eq. (5) is extended
as an infinite line source positioned along the
negative z axis. This gives rise to an additional
integration and equations that are one degree
higher with respect to z/R than Eqs. (5) and (6).
It turns out that the f(z) for the line source is
identical to the right side of the integral function

ORNL-DWG €6--12373R

 

 

 

R#r)

     
 

 

 

DISTRIBUTION FOR IDEAL TARGETS

e =R ep-(B)?]

T anT a

 

F=rm =R

 

o

 

o
74

 

 

 

 

<
=t}

o
n

  
 
 

MAXWELLIAN DISTRIBUTION FOR

 

NORMALIZED DISTRIBUTION FUNCTICNS,

oz

 

POROUS TARGETS

2

m)2 exp [ (—”!;n) ]

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 8.4,

tributions correspond to concentration vs penetration curves for perfectly collimated beam experiments.

 

Plots of the Normalized Disfribution Functions Used to Correlate Recoil Dara.

2.0 2.4 2.8

r/R

These *‘r-space’” dis-
P

In the pres-

ent investigation beams are not collimated, but the observed stopping behavior does relate to the curves shown above.
90

ORNL-DWG 66-12374R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

F.R e R
\ 1 ] ‘ \
i i | "
E | .9\ : ‘ I
og - e L] IR i o S e e
3 * \ |
« ; | \ |
~ 08 i&g DEMSE-ANISOTRCORIC PYROLYTIC CARBON \‘ ISOTROFIC PYROLYTIC CARBON i
B % (DEAL TARGET MATERIAL, d=2 49g/cm’) 'Bg,\ (POROUS TARGET MATERIAL, d=4.49g/cin?) ‘ -
@ ¥ | \\ \ | J :
Z g | |
Z 0.7 - k 1 T k\ - ' e ; ! i
= e \ 5 |
& ;m. \ \ i ‘
T 56 & ~ e By — gfl\, e o 9Byizr
v "8 ‘ ¢ "V xelBa) ¥ o ¥O0xe(ga)
= AV \
E .
2 % >\ |
Z 0S5 % e "\ 5 — —
= L \
04 B : J Yoo ! —_— 1
& * '1 \ <
b e RN |
B 03 [ :fi - r, e \ . ‘f S S
= ‘;i’.q ’ oD
c o2 X ‘ °
0P k- of o N o ] L de
i‘;() .5 1 &«. ‘
o " W2 |
w c:fl ° ‘. . ‘ { !
oA . ?é;% e e L ._.\' O L i
\ 3" o \ Q{:
\8%0 » g © ol .
Q b \f%.si_r@, :Sh...! otenmwl e L 1 01 [ ... \.\ " ° T= BW.-..‘_Q”“:BS_-;—V"—_—_W
o 0.2 0.4 06 08 10 12 14 16 18 O 0z 0.4 06 08 1.0 1.2 14 1.6 1.8 20
z/R z/R
Fig. 8.5. Penetiation Data for ‘‘Light and Heavy'' Fragments in Dense and Porous Pyrocarbons. Penstrations

are referred to R values peculior to given experiments and species.

The curve that passes through the porous pyro-

carbon data was plotted on the basis of F.R., vs z/R values as given by Eq. (7).

for a point source [Eq. (6)] multiplied by a factor
slightly less than 2.
normalize the new distribution cutve.

This factor is required to
Therefore
the experimental plot of Fig. 8.4 is also a plot
of the distribution as it would occur in a coating,
diffusion and densification neglected.

For the case of porous carbons and graphites,
we find that Eg. (6) does not give a good fit of
experimental data plots. Obviously, the normal
r-space distribution does not hold, because well-
collimated beams would ‘‘see’’ density variations
in porous targets. The probability for stopping
f(r) dr about a mean R would not be symmetrical,
as it is for a normal distribution. In terms of
bulk volume most of the matrices of our porous
targets are composed of solids (open porosities
range from 10 to 20 vol %), and the most probable
range r, should correspond to a ‘‘low’” value
which might be predicted using a ‘“high’’ porosity-
corrected density. Thus, on the r-space distri-
bution, r_ should reside to the left of the average
{r) or R, while portions of the curve to the right
of R should “‘tail out’” slowly to account for
fragments

(pores).

that encountered nonstopping regions

Several skewed distribution functions were tested
for applicability to this problem; the most suc-
cessful candidate appears to be that shown in
Fig. 8.4 for porous targets. This function is
simply Maxwell’'s speed distribution for gases with
v/v_ replaced by r/r_.

When the Maxwellian distribution is subjected
to the same manipnlations employed to develop
Eq. (6), one finds that

F.

(7)

A

i

®

-

a
1
/’I—‘\
i ‘:i
N
I
\_/
|

exactly. As before, the F.R. function also rep-
sesents the f(z) distribution in coatings. Typical
porous target data are shown on the right side
of Fig. 8.5. The solid line represents a plot of
Eq. (7), and the dotted line represents a plot of
Eq. (6), which clearly does not apply.

Average range values for porous targets appear
in Table 8.2. Values for two of the porous ma-
terials show good agreement with ideal values.
Range wvalues for the very porous graphite are
somewhat high; however, a large fraction of the
pores in this material were several times greater
than R. Our original intentions were to use these
data to establish an upper limit on pore sizes,
and in this respect the experiments failed because
Eq. (7) was followed (except at large z/R) and
the ranges were only slightly greater than the
ideal values.

In summaty, we have found that effects of po-
rosity and variations in n, can be taken into ac-
count through a proper selection of the primitive
r-space distribution function. When this is ac-

91

complished, the mean range for porous targets
should be approximately the same as those ob-
served or predicted for homogeneous (ideal) ma-
terials. Specific details concerning porosity char-
acteristics are of little importance, if the pores
are not too large or well connected, because the
range correlation depends only on the bulk den-
sity and/or total porosity. Finally, we note that
straggling factors, as given by the theory, are of
little consequence with respect to the present
investigation.
9. Behavior of Graphite with Reactive Gases

L. G. Overholser

OXIDATION OF GRAPHITE SLEEVES BY STEAM

C. M. Blood G. M. Hebert

I.eakage of steam into the coolant circuit of
an HTGR having coated fuel particles and graphite
structural elements in the core could result in
damage to the fuel coatings and graphite com-
ponents due to extensive oxidation if appreciable
partial pressures of steam were present during
the period in which the core components were at
high temperatures. At this time, such parameters
as partial pressure of steam and temperature of
the various core components under accident condi-
tions cannot be precisely defined. Consequently
the steam-graphite reaction must be examined
experimentally using fairly wide ranges of tem-
peratures and partial pressures of steam in an
attempt to cover the abnormal conditions which
may prevail. Some fission products catalyze the
oxidation of graphite, and if such products have
moved from the fuel particle to the adjacent
graphite abnormal reaction rates may be expected.
The

trolled at very high reaction rates resulting from

reaction also may be mass transport con-

excessively high temperatures or strong catalysis
at lower temperatures.

The oxidation of AT] graphite spheres by steam
was studied previously.! More recently, AT]J
graphite sleeves have been oxidized at 1000°C
using a helium-steam mixture having a partial
pressure of ~250 torrs and a total pressure of
1 atm. Mullite reaction tuhes and quartz deposition
tubes used in these studies have been replaced
recently with alumina tubes, and studies now are
being performed in the temperature range of 1100
to 1500°C. 1

Graphite sleeves (]f,’/w-in.—OD, s

 

IJ. L. Rutherford, J. P. Blakeley, and L. G. Over-
holser, Oxidation of Unfueled and Fueled Graphite
Spheres by Steam., ORNL-3947 (May 1966).

92

in.-ID) were machined from AT] graphite stock
and various lengths used in the oxidation studies.
These dimensions approximated those of graphite
specimens available from irradiation studies.
Graphite sleeves (2 in. long) were impregnated
with barium using 1338&(112 and Ba(()H)2 solution,
drying the impregnated specimen at ~125°C, and
finally heat-treating in diy helium at 800 or 1000°C
for various periods of time. Sectioning and count-
ing of impregnated specimens are incomplete, but
preliminary measurements suggest that the treated
specimens contained ~0.1 wt % of barium and that
the barinm was not uniformly distributed through
the graphite. One specimen of graphite previously
irradiated in loop 1, experiment 14 (ref. 2) was
oxidized and examined for transport of fission
products.
Reaction rates measured for wvarious graphite
specimens are given in Table 9.1. Average rates
determined from weight changes are expressed on
a weight basts, but essentially the same telative
rates would be obtained if geometric surface areas
were employed instead of weights because of the
geometiy of the specimens. A superficial com-
parison of the rates obtained for the untreated
specimens of various lengths suggests that the
rates increased with decreasing sleeve lengths.
The reaction
burnoff, and any critical comparison of reaction
The
length effect appears to be small in those cases
where burnoffs are comparable. Reaction rates

rates increased with increasing

rates must take this effect into account.

given for the specimens impregnated with barium
show that the steam-graphite reaction was def-
initely catalyzed. Data are too fragmentary to
indicate whether or not length of the specimen
had any effect on reaction rates. Data given for
the one previously imradiated graphite specimen

2A. W. Longest et al., GCR Program Semiann. Progr.
Rept. Mar. 31, 1966, ORNL-3951, pp. 56--64.
Table 9.1. Reactivity of ATJ Graphite Slesves with Steam at 1000°C

 

 

(Pyy_o ™ 250 tores)
Specimen Length  Flow Rate? Reaction  Initial g 0o Reaction Surtace Arca’
Designation (in.)  (em?/min, sTPy  Time  Weight . o Rate” _me
(hr) () (mg g~ he™!)  Original  Final
11-B 0.5 900 3.3 3.931  27.4 96 0.16 8.2
18-8 0.5 900 3.4 3.952  33.4 116 0.15 9.9
UNL-1-EXP 14E? 0.5 900 3.4 4.140  50.6 200 0.09 e
ATJ-2-CYL-9B' 0.5 900 1.1 3.798  32.9 365 e o
13-A 1 900 3.2 7.888  19.6 68 0.15 7.8
11-A 1 900 3.3 7.791  21.6 73 0.16 7.2
UNL-1-EXP 14-A9 1 900 3.3 8.427  24.8 85 0.09 13.8
ATJ-3-CYL-3A 1 900 1.1 8.200 3.1 30 0.14 2.6
ATJ-3-CYL-2A 1 900 2.1 8.530 8.2 41 0.13 5.2
ATJ-2-CYL-GAS 1 900 1.2 7.412  40.9 420 0.17 8.5
10 2 400 3.4 14.9012  14.1 45 e e
12 2 400 2.7 15.608 9.2 36 e e
17 2 900 3.7 15.211  13.0 37 0.15 7.3
IR-GI-28 1.6 900 3.3 12.424 13.4 - 43 0.18 1.4
ATJ-2-CYL-8" 2 500 2.1 15.330  39.8 240 0.16 16

 

a . .
Flow rate given for helium.
bAverage reaction rate based on average burnoff.

“BET surface area obtained with nitrogen.

dSleevr: from A'lJ graphite stock used in loop 1, experiment 14.

°Not determined.
fImpregnated with barium.

gA’I‘j graphite sleeve from loop 1, experiment 14 following irradiation.

from loop 1, experiment 14 (ref. 2) indicate that
neither the prior irradiation nor the presence of
small amounts of fission products had any sig-
nificant effect on the reaction rate.

Surface area data included in Table 9.1 iadicate,
as would be expected, that the reaction rates of
the unimpregnated sleeves increased, in general,
with increasing sutface atea although no quanti-
tative relationship is evident. Limited data avail-
able for the impregnated sleeves suggest that a
different relationship between reaction rate and
surface area development prevailed in the presence
The low value found for the surface
area of the oone irradiated graphite specimen after

of barium.

oxidation may or may not be =significant; additional
data from other irradiated specimens are needed to
resolve this question.

The studies are continuing with emphasis being
placed on the effect of higher temperatures (11060
to 1500°C) on the steam-graphite system.

TRANSPORT OF FISSION PRODUCTS
C. M. Blood

Quartz deposition tubes were used in the runs
in which the barium-impregnated and the previously

irradiated graphite sleeves were oxidized by
steam, as well as during the prior heat treatment
of the impregnated sleeves in dry helinm. These
tubes were utilized to determine any movement
of barium downstream from the impregnated spec-
imens in wet or dry helium and (in the case of
previously irradiated graphite specimens) to cap-
ture any fission products transported by wet
helium. A number of the deposition tubes have
been sectioned by cutting into 1.5-cm lengths,
and the gamma activity due to '°°Ba has been
measured in each section,

Studies using graphite sleeves impregnated with
133Ba and subsequently heat-treated at 800 or
1000°C employing a flow of 330 cm®/min (STP)
of dry helium showed that very small quantities

94

of 1°°Ba deposited on the deposition tubes. Even
smaller quantities of transported 133Ba were found
on the deposition tubes from runs in which heat-
treated impregnated graphite were
oxidized by steam-helium mixtures [900 cm®/min
(STP) of helium| at 1000°C. The data suggest that
barium is less readily transported by wet belium
than by dry helium. The fact that all the impreg-
nated graphite specimens had received a heat
treatment in dry helium prior to oxidation compli-
cates the comparison, however, because there is
evidence that barium is fixed in graphite to some
extent by the prior heat treatment. The flow rate
of the dry helium during the heat treatment at
1000°C appears to have had an important effect

specimens

ORNL-DWG 67- 539

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

13354 ACTIVITY (counts min~! em™! x:073)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\

 

 

 

 

 

 

 

 

o o O o metr Qe ] T = 2ale O e et el Jpmerwraree 3
28 32 36 40 44 48 52 56 60
DISTANCE (cm)

Fig. 9.1. Distribution of 133p, Activity as a Function of Distance (Temperature) Along the Deposition Tube.
on the movemen{ of barium. The one run made at
975°C using a flow rate of 930 cm®/min (STP)
of dry helium showed a much larger transport of
13384 than those runs made at the lower flow rate
of 330 cm?¥/min (STP). Counting data obtained
from the sectioned deposition tube used in the
run at the higher flow rate of dry helium are given
in Fig. 9.1. The marked temperature dependence
of the 1?*Ba deposition suggests that condensation
of some species occurred over a relatively short
length of the deposition tube. At this stage, how-
ever, the transported species has not been identi-
fied, and the available data do not permit an
accurate determination of the condensation tem-
perature or heat of sublimation. A gamma scan
of the tube did not provide a better profile than
that obtained by sectioning. A smaller temperature
gradient and improved sectioning or scanning
techniques are required to provide the more precise
data needed for further study of the transported
material.

Sectioning and counting of the deposition tube
used in the run in which irradiated graphite from

95

loop 1, experiment 14 was oxidized by steam
(Table 9.1) revealed that appreciable activity had
been transported downstream by the wet helium,
The deposition profile is given in Fig. 9.2, in
which the total gamma activity is shown as a
function of distance (temperature) along the dep-
osition tube. The two maxima indicate that at
least two radioactive species had been transported
by the wet helium. Further examination of the
sections by means of a gamma spectrometer pave
the data included in Fig. 9.3, It was possible to
separate the activities due to 'PAg, 1%%Cs, and
137¢s.  Silver-110 was produced from structural
materials which were used in the irradiation ex-
periment and which subsequently migrated to the
graphite. Low levels of activity from other nu-
clides are probably masked by silver and cesium.
Experimental difficulties similar to those indicated
for 1*°Ba transport preclude any rigorous analyses
of data obtained for '!%Ag, '*Cs, and !'*'Cs
transport at this time.

ORNL-DWG 67-540

 

 

&

 

 

an

D

 

 

 

=]

N

 

 

—

 

 

 

 

 

 

y ACTIVITY {counts min™' cm™" x1073)

O

 

 

 

 

 

 

 

 

 

 

 

30
DISTANCE (ecm)

35 40 45 20 35

Fig. 9.2. Distribution of Gamma Radicactivity os a Function of Distance (Temperature) Along the Deposition Tube.
9%

ORNL-DWG 67-5914

 

 

 
 

......... e S

 

 

 

 

 

 

4 b oo _

® TOTAL ACTIVITY
0 137(:8
A ‘HOAG

—_ i

& O 34CS

|

¢ (

T

£ 3 i

T

£

E

o

o

S .

=

=

—

O

e

Q

o 20— —

<

o

>

w2

L

S

Let

o

o

 

 

 

 

 

 

 

 

 

 

 

25 30 35 40

 

 

 

 

 

45 50 55 60

DISTANCE (cm)

Fig. 9.3. Spectra of Specific Gamma Radioactivities as a Function of Distance Along the Deposition Tube.

OXIDATION OF COATED FUEL PARTICLES
BY WATER VYAPCR

J. E. Baker

The pyrolytic-carbon coating on the fuel particle
is expected to retain gaseous fission products
until the coating fails; nonvolatile fission products
may diffuse slowly into the coating during pro-

longed use at high temperatures. If {eakage of
water vapor into the reactor occurs, subsequent
oxidation of the coating on the fuel particle may
cause failure of the coating and release of gaseous
fission products. The nonvolatile fission products
may be transported by the coolant and deposited
on various surfaces within the reactor following
failure of the coating.
97

Fatlier studies?®'? of the oxidation of pyrolytic-
catbon-coated fuel patticles by water vapor were
performed at temperatures of 1000°C orx less and
utilized partial pressures of water vapor ranging
from 4.5 to 560 tous. The effects of higher tem-
peratures and lower concentrations of water vapor
have been examined in more recent studies.®
Rates of reaction of pyrolytic-carbon coatings
present on unirradiated fuel particles were deter-
mined at temperatures of 1100 to 1400°C. Helium—
water-vapor mixtures containing 500 to 1000 ppm
(parts per million by volume) of water vapor and

having a total pressure of 1 atm were used in a
single-pass system. Reaction rates were obtained
from weight changes given by a continuously re-

 

3C. M. Blood and L. G.
Semiann. Progr. Rept. Sept.
12530,

4¢. M. Blood and L. G. Overholser, Compatibility of
Pyrolytic-Carbon Coated Fuel Particles with Water
Vapor, ORNL-4014 (November 1966).

5y. E. Baker and L. G. Overholser, GCR Program
Semiann, Progr. Rept. Sept. 30, 1966, ORNL-4036 (in
press).

Overholser, GCR Program
30, 1965, ORNL.-3885, pp.

Table 9.2. Rates of Reaction of Pyrolytic-Carbon-Coated Fuel Particles with Water Yapor

 

 

Water Vapor Exposure Percent of Reaction Coating
Coat?d Fu:;l Tem%erature Concentration F310w Rate Time Coating Rateb Failure©
Particles e (ppm) {em”/min, STF) (hr) Oxidized (mg g_l hr "'1) (%)
Isotropic V7 1100 1000 200 72 31 5.0 e
1200 1000 200 24 39 16 44
1300 1000 200 20 53 27 52
1400 1000 200 15 52 33 87
Isotropic VI? 1100 1000 200 23 1.7 0.5 0.2
1200 1000 200 23 3.6 1.5 0.2
1300 1000 200 23 12 5.6 0.1
1300 1000 400 24 17 7.4 0.1
1300 500 400 24 10 4.2 0.1
1400 1000 200 24 43 19 32
Isotropic VII? 1300 1000 400 24 19 8.7 1.6
Granular 1V 1200 1000 200 24 12 4.7 51
1300 1600 400 24 46 16 72
OR 688" 1200 1000 200 24 21 8.1 6.7
OR-689 1200 500 200 24 8.3 4.4 e
1200 1000 200 24 19 8.4 e
Y7 1347 1200 1000 200 24 38 20 e
¥7-1357 1200 1000 200 24 48 20 17
vZ-136" 1200 1000 200 24 31 18 16

 

“100-mg sample used in all cases.

PReaction rate based on weipght of pyrolytic-carbon coating; rates given for less than 5% burnoff.

®Calculated from quantity of uranium (thorium) in acid leach solution and total quantity of uranium (thorium) origi-

nally present in 100 mg of coated fuel particles.

dSupplied by General Atomic Division, General Dynamics Corp.

®*Not available.

f:"':u;.mliecl by Metals and Ceramics Division, Oak Ridge National Laboratory.
cording semimicro balance and from analyses of
effluent gases performed by a sensitive gas chro-
matograph, The extent of coating failures was
determined by microscopic examination and acid
leach of the oxidized coated fuel particles.

Experiments performed in mullite reaction tubes
during the early part of these studies gave results
which were so erratic that it was impossible to
determine the effect of temperature and water vapor
concentration on the reaction rates. In virtually
all cases, the observed reaction rates decreased
with time, and in some instances the final rates
were an order of magnitude lower than the initial
rates, Consistent data obtained after replacing
the mullite tube with an alumina tube indicate that
mullite was responsible for the erratic results.
Gearey and Littlewood® also observed a decrease
in reaction rate attributed it to
catalysis of the steam-graphite reaction by silica
from the mullite tube.

Reaction rates obtained for various batches of

with time and

coated fuel particles using an alumina reaction
tube are given in Table 9.2. The most interesting
feature of these data is the large variation in

reaction rates found for the different batches.

GD. Gearey and K. Littlewood, Nature 206, 395 (1965).

98

Isotropic VI and VII particles, for example, were
less The
coating density (1.55 compared to 2.00

reactive than Isotropic V particles.
lower
g/cm?® for Isotropic VI) of Isotropic V particles
may be responsible for the higher reaction rates.
Some of the YZ and OR batches of particles,
however, have properties (including density) very
similar to those of Isotropic VI particles but were
considerably more reactive.
The long exposure times combined with rel-
atively high reaction rates resulted in oxidation
of a large portion of the coatings present on a
number of batches of particles. In view of this,
it is not surprising that severe damage and high
percentages of failures occurred. Microscopic ex-
amination suggests that a pitting attack followed
by cracking of the coatings occurred at all tem-
peratures.
Isotropic VI particles because of lower reaction
rates and less burnoff of the coatings. These and
other data indicate that ~10 wt % of the coatings
may be removed before failure occurs with Iso-
tropic VI particles. Subsequent studies, in which
failures will be detected by bursts of activity

Conditions were more favorable for

from irradiated materials, will utilize higher partial
pressures of water vapor to reduce failure time,
the

particularly when oxidizing resistant

types of coatings.

moreg
10.

Irradiation Behavior of High-Temperature

Fuel Materials

0. Sisman

IRRADIATION EFFECTS ON PYROLYTIC-
CARBON-COATED FUEL PARTICLES

P. E. Reagan  J. G. Morgan  J. W. Gooch
M. T. Morgan M. F. Osborne

We are studying the irradiation effects on coated
fuel particles by measuring the fission-gas release
rates during irradiation at high temperatures, and
by postirradiation examination to determine what
damage was done. These studies are being pér-
formed in cooperation with the groups who are
developing the coatings for fuel particles  at
ORNL, General Atomic, and the Carbon Products
Division of Union Carbide Corporation. The in-
tegrity of production-run coated particles for the
AVR reactor was studied during a long-term test.
The effectiveness of silicon carbide barrier to
solid fission product release is being studied,
and the effectiveness of a gas gap in retarding
sclid fission product release is also being in-
vestigated.

Pyrolytic-carbon-coated
to 1) carbide particles, prepared commercially for
fuel elements for the pebble-bed gas-cooled
German AVR reactor, were irradiated to 10 at., %
heavy-metal burnup at 1300°C.'} These were a
blend of several batches of duplex-coated pat-
ticles that were representative of the production-
run coated particles.? The fractional fission-gas
release for 28Kr was 5 x 107 ° at the beginning of
the test and increased with burnup to 4 < 1075,
No bursts of fission gas were released during the
test, and no broken coatings were found on post-
irradiation examination.® Metallographic exami-
nation revealed some damage to the inner coating,
but nothing that indicated potential failure of the
coated particles.

thorium-uranium (4,59

9G

J. G. Morgan

Pyrolytic carbon coatings on fuel particles will
contain essentially all of the fission gases, but
at elevated temperatures some of the solid fission
products {(notably St, Ba, and Cs) will diffuse
through the coating. To reduce the mipration of
fission solids, a silicon carbide barrier layer
may be deposited between layers of pyrolytic
carbon., One experiment (capsule AS7) contain-
ing thorium~uwranium carbide—coated particles of
this type (batch GA-327) was irradiated at 1300°C
for 1800 hr. The fission-gas release was low
(the R/B for ®%Kr was in the 10~% range), and
postirradiation examination at 30x showed no
Postirradiation metallography
and fission solid analysis are not complete, but
preliminary. results on postirradiation heating ex-

damaged coatings.

periments have indicated very low release of the
solid fission products at temperatures up to
2000°C.

One irradiation capsule assembly (A9-6), after
being irradiated for 1100 hr at 1500°C, was ex-
amined for the location of the principal solid
fission products. In this experiment the graphite
fuel particle holder was isolated from a graphite
shell by an annular gas gap as shown in Fig
13.1. Seven components of the assembly, ine
cluding the fuel particles, were either acid leached
or dissolved for recovery of 39Sr, 9°Zr, 1%7Cs,

 

lM. N. Burkett, W. P. Eatherly, and W. O. Harms,
““Fueled-Graphite Elements for the German Pebble-Bed
Reactor (AVR),’”’ paper presented at the 1966 AIME
Nuclear Metallurgy Symposium on High Temperature
Nuclear Fuels, Delaran, Wis., Qct. 3--5, 1966 (to be
published).

2R. A. Reuther, Nucl. Sci. Eng. 20(2), 219 (1964).

3p. E. Reagan, *‘Fission-Gas Release and [rradiation
Damage to AVR Pyrolytic Carbon Cozted Thorium-
Uranium Carbide Particles’’ (in press).
and '%*Ce. Some data were obtained for **%Ba,
but the decay time was too long for conclusive
results. No appreciable fractions (<10 2) of the
zirconium and cerium were found outside the fuel.
About 5% of the cesium had escaped from the fuel
and graphite holder and had condensed on the
relatively cold metal surfaces of the capsule.
Strontium appeared to be considerably more mobile;
more than 50% of the ®°Sr had escaped the fuel
and was found in significant fractions (>1%) in
the graphite holder and shell and on the metal
capsule. With the exception of some 2% of the
137Cs  found the lead tube,
fractions of any of the fission products were found
outside the irradiation capsule. This is the first
of a series of experiments to study the release of
solid fission products during irradiation. Future

in no significant

ORNL-DWG 66-4C62

,,flg—rflmm HOUSING
Y - FILTER THIMBLE
: gfl 28 ALUMINUM

—-COOLING CQIL, 12 TURNS
STAINLESS STEEL TUBE
3.2 mm QD

FILTER
rLANDERS AIRPORE

 
 
   
   
  
  
  
   
 
    
    
   
 

  

FILTER HOUSING
EXTENSION ——w

 

 

 

 

STAINLESS STEEL TUBE
THERMOCOUPLE HOUSING

CAPSULE CAP AND He INLET

INCONEL-—2E7 :
= INCONEL CAPSULE

THERMOCOUPLE SHEATH —PyC SHEETS

RHENIUM FOIL

7—HOLDER CAP
OXIDE WOOL*HT :
g |

COATED PARTICLES

 

i _—-DEPOSITION PLATE
THERMOCOUPLES (2)

T = SHELL
THERMOCOUPLES (4)

CENTER It
THERMOCOUPLES (3)-— 1 THEF

 

 

 

DEPOSITION PLATES
2 HALVES

I“"HOL,DER PIN

 

HOLDER

 

 

 

 

DEPOSITION END PLATES, (4 LAYERS)

}(\Pyc SHEETS
TOP AND BOTTCOM

.

 

PyC SHEETS,
TOP AND BOTTOM

Fig. 10.1, Gas-Gap Fission Solid Capsule.

100

studies in this series will include an improved
filter design for othet
products, especially !*°Ba.

and analysis fission

IN-PILE TESTS OF A MODEL TO PREDICT
THE PERTFORMANCE OF COATED
FUEL PARTICLES

P. E. Reagan
E. L. Long, Jr.

J. G. Morgan
J. W. Gooch

A mathematical model that will predict the
conditions under which a pyrolytic-carbon-coated
particle will fail has been formulated by Prados
and Scott.* This model takes into consideration
the combined effects of fuel swelling and fission-
gas pressure from fuel burnup, and fast neutron
damage to the coating., To test the model, we
irradiated a capsule containing about 4000 coated
particles with structural characteristics for which
the model was readily applicable. These were
sol-gel U0, particles, coated with a high-density
laminar pyrolytic carbon inner coating and a
high-density isotropic carbon outer coating. The
average fuel particle diameter was 202 y, and the
average coating thickness was 123 u. These par-
ticles (designated OR-YZ66) were irradiated at
1400°C in the B9 facility. After 1209 hr (15 at. %
uranium burnup), the particles began to release
bursts of fission gas at irregular intervals. The
bursts continued, and at the end of the test (23%
burnup) we had observed 197 bursts. Postirradia-
tion examination of the coated particles showed
that about 5 to 10% of the coatings had indeed
broken and in many cases were completely sep-
arated from the uranium oxide particles, as shown
in Fig. 10.2. Mectallographic examinations are in
progress, and postirradiation burnup determina-
tions are being made to check the values ob-
tained by argon activation during irradiation.

The mathematical model had predicted coating
failure at 14.2 at. % burup.® Since the first
failure occurred at 15% burnup and a rather small
fraction of the coated particles had failed even
at 23% burnup, it appears that the calculation gave

 

*J. W. Prados and J. L. Scott, Nucl. Appl. X8),

403-14 (1966).

°D. M. Hewette ¢t al., Preparation and Preirradiation
Data for Pyrolytic Carbon Coated Sol-Gel Uranium Oxide
Particles for ETR X-Basket 2 [Irradiation Experiment,
ORNL-TM report (in preparation).
a low value. If the postirradiation burnup value
proves this to be tme, the constants in the
equations can be changed to increase the accuracy
of the calculation.

To study the effect of the ratio of thickness of
porous carbon to dense pyrolytic carbon in the
particle coating, uranium oxide particles were
coated with three different ratios of dense to
porous coating but with the same total coating

 

10.2.
Particles from Batch OR-YZE8E, lrradiated to 23 ot, %
Uranium Burnup ot 1400°C in Copsule B9.31. 30

Fig. Pyrolytic-Carbon-Coated Uranium Oxide

Table 10.1. Pyrolytic-Carbon-Coated uod, Particles
lrradiated to Study Optimum Coating-Thickness Ratio

 

 

OR-491 OR-493 OR-494
Average particle
dimensions {{0
Total diameter 713 703 704
Fuel particle 429 410 4172
Porous carbon coating 28 a1 93
thickness
Pyrolytic carbon coating 114 81 63
thickness
Total ¢ceoating thickness 142 142 146
Porous ceating to 0.25 0.70 1.47
pyrolytic coating ratio
Number of particles 5 10 570

irradiated

 

101

thickness.® The coated-particle dimensions are
given in Table 10.1. These coated particles were
irradiated in capsule B9-30 at 1400°C for 159 hr
At the beginning of
the test, the fractional release of ®3Kr was in the
1077 range, but it increased suddenly to the
10~ * range near the end of the test when pasticles
from batch OR-494 (which had the largest ratio of
porous coating to pyrolytic coating) began: to
rupture. None of the 15 low- and intermediate-
ratio particles (batches OR-491 and -493) failed;
in both cases about half of the thickness of the

(1.9 at. % uranium burnup).

porpus carbon coating had been consumed during
irradiation; but there was no evidence of uranium
carbide formation. Most of the high-ratio parcticles
(batch OR-494) either failed or were damaged
almost to the point of failure. Failure was by
unilateral movement of the Jo, through heth
layers of the coating. The fuel in these particles
appears to have operated at a higher temperature
than particles with the thinner porous carbon
coating. Crystals of UC_ were found at the fuel
sutface of the totally failed particles.

When the experiment was designed, it was as-
sumed {hat the five coated particles with the
thinnest porcus carbon layer would fail first,
because there is little room for expansion of the
fuel and accumulation of fission gas. The coated
particles with thick layers of porous carbon were
not supposed to fail. The burnup was too low
to cause failure in any of these particles. We
have observed, however, that a very thick buffer
layer of porous carbon will insulate the fuel
particle and may cause failure because of ex-
cessively high temperature in the fuel.

POSTIRRADIATION TESTING OF COATED
FUEL PARTICLES

M. T. Morgan C. D. Baumann
R. L. Towns

To aid in the development of better coatings for
coated fuel particles, a series of postirradiation
anneals have been made on four types of pyrolytic
carbon coatings applied to UC, and U0, fuel
particles. Fuel migration studies by metallography

 

6P. E. Reagan, E. L. Beatty, and E. L. Long, Jr.,
“Pearformance of Pyrolytic Carbon Coated Uranium Oxide
Particles During Irradiation at High Temperatures’ (in
press).
and microradiography have been made, and fission
products released during anneals are being ana-
lyzed. Anisotropy, density, and crystallite size
were the wvariables in the four types of outer
coatings, the details of which are given in Figs.
10.3 and 10.4. These coatings were applied over
identical inner coatings. The coated particles
were all irradiated in the same capsule at the
same temperature (1400°C) and to the same burnup
(14.5% of the heavy metal). Samples of ten
coated particles from each batch were annealed at
1700°C, and duplicate samples were annealed at
2000°C for a total of 191/2 hr for each sample in
steps of 61/2 hr. 78

Evoluation of Fuel Migration by Meta!lography
and Microradiegraphy

Microradiographs and metallographic photomi-
crographs were made on each of the eight batches
of coated particles as coated, on the irradiated
coated particles before heating, and on the ir-
radiated coated particles after the 1700 and
2000°C anneals. The photomicrographs are shown
in Figs. 10.3 and 10.4. The microradiographs are
not shown, but the outer diameter of the dense
areas in the radiographs corresponds to that of
the shaded or spotted ateas of the coatings in the
photomicrographs, indicating uranium diffusion
into these areas.

Reaction areas in the coated uo, particles
were restricted to the inner coating. Spearhead-
type attack occurred during irradiation in coated
particles with low-temperature outer coatings, and
the UO, seemed to expand to fill the void areas.
In the coated uo, particles with outer coatings
deposited at 2000°C, spearhead attack is not
evident, and penetration of the inner coating is
slight: this is probably a result of the 2000°C
heat treatment received during the deposition of

the outer coatings. The uo, particles lost

 

TM. T. Morgan, D. C. Evans, and R. M. Martin, ‘‘Fis-
sion-CGas Release from High-Bumup UOQ," GCR Program
Semiann. Progr. Rept. Mar. 31, 1963, "ORNI.-3445, pp.
103.--6.

M. T. Morgan et al., ‘‘Fission-Gas Release f{rom
High-Burnup Coated Particles,’” GCR Program Semiann.
Progr. Rept. Sept. 30, 1963, ORNL-3523, pp. 149-52,

102

and void
No

crystalline detail during irradiation,
were redistributed as smaller voids.
carbide formation was evident.

The UC, fuel showed no apparent fine porosity
after irradiation. Heat treatment at 2000°C seemed
to have consolidated the fine porosity into larger
pores or voids. Some graphite flakes were seen
around the periphery of the fuel particles. The
particle-coating reaction areas in the coated
carbide particles were not as local as the spear-
head attack in the coated oxide particles and
were asscciated with a progressive, more ex-
tensive diffusion of fuel into the coating. The
microradiographs of the unirradiated carbide par-
ticles indicated that fuel diffusion into the inner
coatings started during the manufacture of the
particles.

areas

Penetration of the inner coatings was
complete in all coated carbide particles after
irradiation, as indicated by the spotted areas in
the micrographs. Diffusion of fuel deep into the
coating was evident in the
irradiated YZ-13 coated particles. Large voids
and separation of the inner coating froin the outer
coating were apparent in the YZ-15 and YZ-19
coated particles after heat treatment at 1700 and
2000°C.

low-density outer

Coating Stability

No coating failures were observed in the eight
groups during irradiation at 1400°C. Coatings of
the UC, particles failed during the postirradiation
tests at both 1700 and 2000°C, but no failures
occurred with the UQ, particles. Based on ®°Kr
release, approximately 30% of the YZ-4 sample
and 25% of the YZ-13 sample failed at 1700°C.
At 2000°C, 80% of the YZ-4, 70% of the YZ-13,
and 45% of the YZ-19 samples failed. Failures all
occurred during heatup or within a few minutes
after the annealing temperature was reached.

The two samples of coated uc, particles which
did not fail at 1700°C (one of which also survived
the 2000°C anneal) had undergone the heat treat-
ment during fabrication. We think that the migra-
tion of fuel into the coatings of UC, particles
weakened the coating and caused the high per-
centage of failures.
BATCH NUMBER
CORE MATERIAL

DEPOSITION TEMPERATURE, °C
CH, SUPPLY RATE, em¥min-cm?
GUTER COATING THICKNESS, s
OUTER COATING DENSITY, g/ctn®

CUTER COATING CRYSTALLITE 5|ZE,LC,E
CUTER COATING AMISOTROPY FACTOR, o,/

BEFORE  IRRADIATION

AFTER IRRADIATION

IRRADIATED AND
ANNEALED 19% ir
AT 1700°C

IRRADIATED AND
ANNEALED 19V, tr
AT 2000°C

“ox

INNE R
INNER
INNER
INNER

103

DUTER COATING DEPOSITION CONDITIONS

YZ--2
U0,
1300
2.0
24
1.99
30
5

 

INNER COATING [)EPOSITION CONDITIONS

U0, FUEL PARTICLES
DEPOSITION TEMPERATURE, °C:1600
CH, SUPPLY RATE, cm®/ min-cm?:4.0
COATING THICKNESS, ;2 26

YZ-4
UC2
1300
20
83
2.0
30
45

COATING DENSITY, g/cm>:1.56

COATING CRYSTALLITE SIZE, L., 8:50
COATING ANISOTROPY FACTOR, ¢y /9oy !

ALL BATCHES OF UG, PARTICLES

HAD THE SAME INNER COATING

YZ-14
uo.
1600
20
76
1.44
50

uc

2

PHOTC 234909

YZ -3
uc
500
20
74
149

 

FUEL PARTICLES

DEPCSITION TEMPERATURE, "C: 1600
CH, SUPPLY RATE, em/min-cm?:1.0

INNER COATING
INMNER COATING
INNER  COATING
INNER COATING

THICKNESS, 140 34
BEMSITY, g/cm>:1.63
CRYSTALUITE SIZE, L, K80
ANISOTROPY FACTOR, oy /oo,

ALL BATCHES OF UC, PARTICLES
HAD THE SAME INNER COATING

Fig. 10.3. Photomicregraphs of Pyrolytic-Carbon-Couated U02 and UC, Particles Befere and After lrradiation and
After Postirradiation Heat Treatment at 1700 and 2000°C Respectively. Groups YZ-2, YZ-4, YZ.14, ond YZ-13.
QUTER COATING

BATCH NUMBER YzZ-41
CORE MATERIAL uo,
DEPOSITION TEMPERATURE, °C 2000
CH, SUPPLY RATE, cm¥min-cm? 02
OUTER COATING THICKNESS, u 78
OUTER COATING DENSITY, g/cm’® 200
OUTER COATING CRYSTALLITE SIZE,LC,B. 120-130

OUTER COATING ANISOTROPY FACTOR, o, /00y 15

 

BEFORE IRRADIATION

AFTER IRRADIATION

IRRADIATED AND
ANMEALED 19% hr
AT +700°C

IRRADIATED AND
ANNEALED 19% hr
AT 2000°C

UO2 FUEL

104

DEPOSITION CONDITIONS

YZ-15
uc,
2000
0.2

78
2.05
120-130
15

INNERR  COATING

PARTICLES

DEPOSITION TEMPERATURE, *C:4600

CH4

SUPPLY RATE, cm¥min-em©:4.0

INNER COATING THICKNESS, pu@ 26

INNER COATING DENSITY, g/cm?:1.56

INNER COATING CRYSTALLITE SIZE, LC,KSZ5O
INNER COATING ANISOTROPY FACTOR, oy, /Uy !

AlLL BATCHES OF us, PARTICLES

HAD THE SAME

INNER COATING

PHOTO 284908 -

YZ-18 YZ-19

uo, uc,

2000 2000

10 1.0

82 76

194 194

135 15

1-1.2 -1.2

DEPOSITION CONDITIONS

UC2 FUEL PARTICLES

DEPOSITION TEMPERATURE, *C: 1600
CH, SUPPLY RATE, cm®/min-cm? 1.0

INNER
INNER
INNER
INNER

COATING THICKNESS, & : 34
COATING DENSITY, g/cm3:1.63

COATING CRYSTALLITE SiZE,L.R 50
COATING ANISOTROPY FACTOR, UOZ/onii

ALL BATCHES OF UC, PARTICLES
HAD THE SAME INNER COATING

Fig. 10.4. Photomicrographs of Pyrolytic-Carbon-Coated UD2 and UC, Particles Before and After Irradiation and
After Postirradiation Heat Treatment at 1700 and 2000°C Respectively., Groups YZ-11, YZ-15, YZ-18, and YZ-19,
IRRADIATION EFFECTS ON COMPATIBILITY
OF FUEL OXIDES AND BERYLLIUM
OXIDE WITH GRAPHITE

. R. Cuneo
C. A. Brandon”®

H. E. Roberison
E. L. Long, Jr.'°

The compatibility of (U,Th)D2 with graphite and
the compatibility of beryllium oxide with graphite
were studied in separate experiments in the ORR.

(U, Th)O zquphite Experiment

The objectives of this experiment were to study,
at a fuel center temperature of 1650°C and a sur-
face temperature of 1370°C, irradiation etfects on
chemical compatibility of (U,Th)O, with graphite
and fission-gas release, and to determine possible
fuel swelling and its effects on the graphite.

The assembly was swept with helium containing
250 ppm of CQ to suppress the reaction of the
fuel oxides with graphite. The fuel loading and
operating conditions are given in Table 10.2.

During the last six weeks of irradiation, the
133%e release rate increased from 1.5 to 18%;
the temperature drop from fuel center to surface

105

increased from 315 te 390°C. The gamma scan of
the capsule after irradiation and before disassembly
showed about 12 individual peaks of activity for
the space occupied by eight hollow pellets in the
upper fuel region. This indicated that the pellets
were broken and pieces had separated; this was
verified upon disassembly.  The seven solid
pellets in the upper region showed general de-
terioration of their outer surfaces. The 12 pellets
in the lower region appeared intact; however,
attempts to determine dimensions caused one of
them to powder.

Metallography revealed that because of densi-
fication there was much less porosity in the
irradiated sample (compared to an unirradiated
control), and shrinkage cracks were apparent. In
spite of this shrinkage, the diameters of two pel-
lets which we were able to measure did not
change. Weight and dimensional changes of the
graphite disks and spacers were negligible.

We conclude that no carbide formation occurred
from the following:

 

YORNL Reactor Division.
10ORNL Metals and Ceramics Division.

Table 10.2. Fuel Leading and Operating Conditions for (U,Th)oz-Gmphife Compatibility Test

 

 

 

 

 

5 Bumup
Composition Density Power _ _ Integrated Flux™ ——rmmommeeees
Fuel Pellet (st %) (% Density 1D OD  Height . tions/em?) ot % U (7 total
Position e e 3, {in.) (inJ) (in,} —————————————— he A heavy
Uoz ThO, of theoretical) (w/cm™) Thermal >4 Mev fissioned) metal
x 10729 x 10'8
Upper Region 1.3 3.3 8.9 0.8
1-8 (hollow) 9.2 90.8 87 600 0.140 0.258 0.125
9.-15 {golid) 9.2 90.8 87 &00 0.156 0.100
Lower Regi-:mb 1.3 3.5 7.3 2.4
16, 18, 20, 33 67 a5 2400 0.070 0.200 0.069
22, 24, 26
(hollow)
17, 19, 21, 33 67 87 2200 0.200 0.061
23, 25, 27
(solid)
=

16507 C for 54 days.

bThe pellets in the lower fuel region were individually inserted in graphite disks and sandwiched between

fueled graphite spacers.

Experiment operated for 83 days at equivalent full power of 30 Mw in the ORE; it was at design temperature of

LI
106

 

 

00X

 

0.035 INCHFS
[ro

 

 

 

 

 

 

100X

0.035 INCHES
Jro

 

 

 

Fig. 10.5. Comparisen of (a) Unirradiated BeO with (b) an lrradiated Specimen from the B=20.Graphite Compatibil-
ity Test. Etched; 100x.
1. No reaction was obtained with powdered fuel
pellets in aqueous media (2 M HCI at 80°C
for 8 hr).

2. The x-ray diffraction pattern of a graphite
spacer which had been in contact with a fuel
peliet showed no carbides.

3. Metallographic studies showed no evidence of

any reaction products on the surfaces of five
pellets or on matching graphite pieces.

Be0O-Graphite Experiment

The BeO-graphite test was designed to study
irradiation effects at 1500°C on the chemical
compatibility of graphite with BeO, to investigate
the distribution of °Li formed in the Be(, and to
determine weight,
changes in the graphite and BeQ. A BeO ring and
a graphite ring were in close contact on flat sur-
faces and were nested over a solid BeD core
within a graphite sleeve.

The experiment was not fueled. The desired
temperature was achieved for about four months
by pamma heating. During the last five months
of irradiation, the temperature gradually fell to
about 1280°C, probably because of formation of
soot {CO » CO, + C), resulting in loss in effec-
tiveness of reflective insulation in the capsule.
The total irradiation dose was 1.0 x 102! uneu-
trons/cem? (E > 0.18 Mev).

There was no physical damage to the BeO or
graphite during irradiation. Metallographic exami~
nation of selected surfaces showed no reaction
products, and x-ray diffraction of the BeO ring
showed no Be,C. There was a deposit of Mo,C
on the graphite sleeve adjacent to several large
holes  which appeared in molybdenum heat
shield.

Examination of the irradiated graphite showed
no microstructural changes. Examination of the
irradiated BeO showed that the grain size doubled,
and each grain was outlined by gas bubbles, as
seen in Fig. 10.5. Analyses of experimental com-
ponents for °Li concentrations showed that the
major portion of the °Li stayed in the BeO core
where it was formed. A surface-to-volume rela.
tionship for °Li retention was found by comparing
the Be() ring with the BeO core. This is in
agreement with work by Stieglitz and Zumwalt. !

Weight and dimensional changes for the BeO and
graphite showed about 1% shrinkage for the
graphite and 0.5% expansion for the BeO.

dimensional, and structural

&l

107

FAST GAS-COOLED REACTOR DEVELOPMENT

D. R. Cuneo
. F. Robertson

E. L. Long, Jr.*°
j- A. Conlin”

During the past year we have begun postirra-
diation evaluations of experimental assemblies
related to fuel elements for fast gas-cooled re-
actors. This is a cooperative program with
General Atomic Division (GA) of General Dy-
namics Corporation. Details of test conditions
and procedures were determined by the ORNL
Reactor Division in cooperation with GA, and the
fuel specimens were supplied by GA. The present
series of tests is designed to investigate the
effects of irradiation, thermal cycling, external
pressure, and fuel-cladding interactions on . the
integrity and behavior of metal-clad UO, fuel
elements having design features that approximate
those for the General Afomic fast pas-cooled re-
actor design. Design and operation of these ex-
perimental assemblies and details of postirradia-
tion findings have been reported elsewhere.!?271°

Two types of elements are being studied in the
first group of tests: the fuel-supported type, in
which the cladding is designed to collapse onto
the fuel pellets at the outset of pressure and
temperature application, and the free-standing fuel
or ‘“flexcan’® type. The latter has a deformed
section of cladding above the fueled region that
is capable of flexing with pressure changes, and
the internal void is filled with sodium.

We have completed examinations of four as-
semblies (lacking metallography for the fourth)
which contained six of the fuel-supported elements
and two of the sodium-filled flexcan types. All
elements were fueled with UOQ pellets,

Operating data and metallograpghic findings are
given in Table 10.3, while the appearance of two
of the test elements is shown in Fig., 10.6.

 

11y.. J. Stieglitz and L. R. Zumwalt, Nucl. Appl. 2(5),
394401 (1966).
126 gr. McQuilkin et al., **Fuel Irradiation Tests in

the ORR Poolside Facility,”” GCR Program Semiann.
Progr. Rept. Sept. 30, 1965, ORNL-3885, pp. 244-5G.

1312 R. McQuilkin et al., ‘‘Fuel Irradiation Tests in
the ORR Poolside Facility,’’” GCR Program Semiann.
Progr. Rept. Mar. 31, 1966, ORNL.-3951, pp. 169-.79.

14]'_). R. Cuneo et al., “*Examination of Irradiated Cap-
sules,’”! GCR Program Semiann. Progr. Rept. Mar. 31,
1966, ORNIL.-3951, pp. 179-80.

. R Cuneo ef gl., ““Examination of Irradiated Cap-
sules,’” GCR Program Semiann. Progr. Rept. Sept. 30,
1966, ORNL.-4036,
Table 10.3. Operating Conditions and Metailogrephic Gbservations for Fast Gas-Cooled Reactor Fue! Elements

 

 

Burnup Maximum
Experiment Fuel Thermal Flux Cladding Cladding
No. Element Dose . (7 235 (% heavy Material Temperature Metaliographic Observations
No. (neutrons/cm ) ¢ metal) ;O
{70
P4B1 GAl 4% 101° 1.6 0.35 Hastelloy X 760+ Capsule badly flattened (see Fig. 10.6). Columnar grain
growth observed in the fuel. No evidence of fuel melting or
lenticular voids. General subsurface voids in inner surface
region of cladding to depth of 1 mil. Formation of thin oxide
layer (™2 u) and thin metallic layer found on ID surface.

GA2 4 % i0t? 1.6 0.35  Hastelloy X 760+ No evidence of fuel melting. Partial coliapse of UO, resultedin
slight wrinkling of cladding. Pellet interfaces had sintered
together. Large radial cracks found at pellet midlengths. No
evidence of attack on cladding.

GA3 (unfueled,

unconnected
flexcan)

P4B2 GA4 5.2 % 10'? 1.4 J.46 Hasielloy X 650 Fuel element not sampled for metallographic examination.

GAS 5.2 x 109 1.4 0.46 Hastelloy X 650 No evidence of fuel melting. Columnar grain growth noted.
Central hole in fuel pellet moved 14 mils off center. Reshap-
ing of central hole was observed as seen in Fig., 10.6. No
microstructural changes observed in cladding.

P4B3 GAO 5.2x 107 2.4 0.58 304 sS 659 No evidence of microsiructural changes in cladding. Neither
{fueled high- nor low-density fuel peliet showed evidence of columnar
flexcan) grain growth or central voids. No deleterious eiffects from

sodium in capsule.

GA7 5.2 x 10%° 2.4 0.59 Hastelloy X 650 No evidence of microstructural changes in cladding. No evi-
dence of ceniral voids or columnar grain growth in fuel peliets.

P4B4  GAS 1.3 % 10%° 4.0 0.86 304 S8 650 Examination continuing.

GA9 1.3 x 10%° 3.0 0.90 Hastelloy X 650 Examination continuing.

 

801
 

 

Fig., 10.6, (a) Transverse Section of Element GAl, Showing Columnar Grain Growth in Central Region of U02 as

Well as Severe Flattening of Cladding; () Longitudinal Section of Element GAS, Showing that U02 Moved Qut Against

Cludding in Groovad Portion of Pellet, Accounting for Increase in Size of Central Hole in Pellet,

To summarize the postirradiation findings:

In all cases the cladding showed no indica-
tions of failure, despite the fact that element
GAl was badly deformed and GA2 was deformed
to a lesser extent. (Some more recent elements
have been prepressurized to avoid large pres-
sure differences across the cladding.)

There was no evidence of incompatibility be-
tween the cladding and the fuel in the ele-
meats tested, except for subsurface voids (fto
a depth of about 10% of the wall thickness)
noted in the inner surface regions of GAl
(Hastelloy X).

There was no evidence cf fuel melting in any
element.

The flexcan elements which contained sodium

in the free volume did not reveal any dele-
terious effects from the sodium.

FISSION-GAS RELEASE DURING FISSIONING
OF U0,

B. M. Carroll ). Sisman
R. B. Perez'® G. M. Watson
T. W. Fulton

In all reactor systems the fission gas which
escapes from the fuel must be coontrolled by
containment within the fuel element or by re-
moval from the fuel coolant. In high-temperature
reactors this problem is especially important.!’
Some empirical formulas have been developed for
estimating the fission~gas release under certain
conditions,'® but most values are obtained by

 

16 . . : -
University of Florida, Consultant to Reactor Chem-
istry Division.

172, M. Carroll, Nucl. Safety 7(1), 3443 (1965).
YW, B. Lewis, Nucl. Appl. A2), 171-81 (1966).
experiments simulating specific reactor operating
conditions. We are seeking a general solution
to this problem by studying the behavior of
fission-gas release under well-controlled condi-
tions in an attempt to understand the basic mech-

anisms of fission-gas release.!? 21

Defect-Trap Medel

As a result of these studies, we believe that
the fission gas is released by a combination
trapping and diffusion process, where the trapping
is the dominant factor determining the time for the
gas to escape. We have formed a defect-trap
theory which postulates that defects in the uo,
crystal structure will trap migrating xenon and
krypton atoms.?? Some defects, such as grain
boundaries and closed pores, are naturally present
in the UO_, and others can be formed by irradia-
tion. The irradiation-formed defects start as
point defects which may be destroyed by annealing
but may cluster to form larger defects that will
require longer times to anneal. Experiments have
that the model has the correct general
23 and that grain boundaries will
24 This model predicts that
which will produce more

shown
characteristics
trap fission gas.
higher fission rates,
fission gas, will also produce more traps; there-
fore, the fission-gas release rate will not change

much with a change in fission rate.
Fission-gas release measurements have been

made during fission-rate and temperature oscil-
These data are being fitted into a com-
puter code to determine the parameters of the
defect-trap model. One complication was caused
by the mixing of the fission gas with the flowing
helium sweep gas. We developed an apparatus to
measure the amount of mixing, using argon to
simulate the fission gas and a thermal conduc-
tivity cell to measure argon concentration in the
The mixing transfer function has now

lations,

sweep gas.

 

lgR. M. Carroll and P. E. Reagan, Nucl. Sci. Eng.
21, 141-46 (1965).

20R. M. Carroll and O. Sisman, Nucl. Appl. 2(2),
142--50 (19606).

21R. B. Perez, Nucl. Appl. 2(2), 151—57 (1966).

22R. M. Carroll and O. Sisman, Nucl. Sci. Eng. 21,
147--58 (1965)

BR. M. Carroll, R. B. Perez, and O. Sisman, J. Am.
Ceram. Soc. 48(2), 5559 (1965}.

24R. M. Carroll and O. Sisman, J. Nucl. Mater. 17(4),
305-12 (1965).

110

been determined and will be used to correct the
fission-gas release transfer function for the in-
pile experiment.

Fission fragments which recoil free of the speci-
men surface will knock out UO, molecules along
with any fission gas in the immediate vicinity.
The knockout release depends on the fission rate,
the total surface area of the specimen, and pos-
sibly the condition of the surface., In order to
study the eifects of surface condition on fission-
gas release, the gas release rates from UO, single-
crystal specimens with highly polished surfaces
were compared with those from unpolished speci-

mens. 5

The fission-gas release from unpolished
specimens decreased with time in the early stages
of irradiation, while the fission-gas release from
the polished single crystals showed no change
We feel this confirms our theory that

the surface is smoothed by irradiation, causing a

with time.

reduction in area,

Measurement of krypton release from a polished
single-crystal specimen is compared with that
from a polycrystalline specimen in Fig. 10.7. At

 

25Kk, M. Carroll and O. Sisman, ORNL-TM-1400 (Dec.
31, 1965).

ORNL-DWG 65-13320R2

 

 

 

 

 

 

 

 

 

 

 

 

 

N

 

 

 

 

 

 

 

RELEASE RATE {atoms /sec)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

106 - o1 A
MM/ - FLUX s [
5 - o=2.3x10" ]
o i x=4.0x 10"
o—o— o0=3.9x 10" |
L L L +=43 410" | .
|
exi0?b v b L :
500 600 700 800 900 1000 00 1200
TEMPERATURE (°C)
Fig. 10.7. Releose Rats of 88+ from Fine.Grain and

Single-Crystal uo,.
temperatures up to about 600°C, knockout release
predominates, and the gas release shows little
temperature dependence. The tougher surfaced
polycrystalline specimen shows greater knockout
release (both specimens are the same size). At
temperatures above 700°C the gas release is from
point-defect traps for both specimens. At 950
to 1050°C the point defects are mobile encugh
to form clusters, slow down the gas release, and
thus cause the step in the curve for the single-
crystal specimen. The point defects are trapped
by grain boundaries in the polycrystalline speci-
men before they can cluster, and therefore the
step is absent. At higher temperatures, the
clusters in the single-crystal specimen may an-
neal and migrate, and once again the release rate
has an exponential temperature dependence, but
the dependence differs from that where release
is primarily from point defects. The temperature
dependence of gas release from the polycrystalline
specimen, where the gas release is always from
point defects, is constant to 1200°C. Al of
these observations were predictable by the defect-
trap model.

THERMAL CONDUCTIVITY OF UD,
DURING IRRADIATION

C. D. Baumann J- G. Morgan
R. M. Carroll M. F. Osborme
R. B. Perez'®

Studies of the effect of irradiation on the thermal
conductivity of UO, have, for the most part, been
limited to postirradiation tests. Annealing tests
show that irradiation at a low (<100°C) tempera-
ture causes a decrease in themmal conductivity
which can be recovered in stages near 150 and
400°C. 2% Effects of density, grain size, and
burnup have also been studied after irradiation. a7
Data are lacking, however, on the actual thermal
conductivity during irradiation as it is affected
by fission rate and temperature. We have begun
our in-pile testing with a single crystal of U0, to
eliminate the parameter of grain-boundary effects.?®

 

zfij. L. Daniel et al,, Thermal Conductivity of UO?,
HW-694945 {(September 1962),

27M. Asagones and M. Guerrero, The Effect of Density
and Grain Size on the Thermal Conductivity of [0,
During frradiation, ARCL-2564 (April 1966).

28R, M. Carroll and J. G. Morgan, Fuels and Materials
Development Program Quart, Progr. Rept. fune 30, 1866,
QENIL-TM-1579, pp 85--95.

111

We are evaluating data obtained while the speci-
men was at three different neutron flux levels and
three temperatures at each neutron flux. The evalu-
ation is not complete, but we can see some rather
definite trends: (1) the thermal conductivity is
lowered by increasing the fission rate at a given
temperature, and (2} the themal conductivity in-
creases with temperature up to about 900°C and
then decreases with further increase in temperature.

These changes of thermal conductivity with tem-
perature, while fissioning, are consistent with
the concept that disorder lowers the thermal con-
ductivity. When the temperature of the specimen
is changed (for example from 600 to 700°C) while
at a constant fission rate, two different processes
are in competition which affect the thermal con-
ductivity: (1) the conductivity of the undamaged
matriz tends to decrease as the temperature (s
increased, and (2) the increased temperature would
increase the conductivity by annealing the accu-
mulated radiation damage. When the temperature
is increased from 600 to 700°C, the themmal ¢on-
ductivity increases because annealing the radi-
ation damage is the dominant process. At tem-
peratures higher than about 900°C, the equilibrium
amount of damage is small enough so that the
change in conductivity caused by the increase in
temperature will result in a net lowering of the
conductivity.

The specimen is moved sinusoidally in the reac-
tor neutron flux. The fission rate and, thus, the
heat production within the specimen respond in-
stantaneously to the change in neutron flux, but
the temperature does not. I the oscillations are
so rapid that the temperature does not have a

 

 

  

 

   
 
 

 

 

 

 

 

 

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x F
5 &
35 2
o
g
Q. 5
= E
~ £
O
% ~ Mg o1 ]
o505 el ILH_T eI
‘:: b b LS -
F)j f{J _ TEMPERATURE AXI5: 1405°C !
g2 FLUX AXIS: 4J,:s|8;<4r.)‘~“' neulrns /5 cm® "\
TP Q2 bt o L N
o) ; ' : NG
o4 L { ....... A el
s o2 5 1077 2 5 w2 sao
CSCILLATION FREQUEMNCY (redians/sec)
Fig. 10.8. Uranium Dioxide Single Crystel Tempera-

ture Response to Oscillating Fission Rate,
chainice to reach its maximum value, then the
amplitude of the temperature oscillations will de-
crease as the oscillation frequency increases (see
Fig. 10.8). The time delay or phase shift between
the neutron flux oscillations and the resultant
temperature oscillations will also change with
frequency. The phase shift and amplitude relation
of the temperature oscillations in comparison with
the neutron flux oscillations over a range of
frequencies will yield data to determine the ther-

mal conductivity of the specimen.

112

The analysis of the temperature response of the
sample to small oscillating changes in neutron
flux (heat generation) has been completed, and
the necessary mathematical relations to obtain
the thermal conductivity have been derived.??
Measurements of the thermal conductivity of single-
crystal UO, at three neutron flux levels and three
temperatures at each neutron flux are in progress.

%R, M. Carroll, J. G. Morgan, and O. Sisman, Fuels
and Materials Development Program Quart. Progr. Rept.,
ORNL-TM-1500 (June and September 1966).
1.

Behavior of High-Temperature

Materials Under Irradiation

EFFECTS OF FAST-NEUTRON iRRADIATION
ON OXIDES

G. W. Keilholtz R. E. Moore

A comparison of the effects of fast-neutron
irradiation on sintetred compacts of BeO, MgO,
and a-Al O, has been reported previously.!~3
During the past year we have initiated an io-
of the irradiation behavior of
commercially available translucent aluminum
oxide of high density* which is of interest as a
possible iasulator in themmionic emitters. The
objectives were (1) to establish its limits of
stability toward fast aeutrons and (2) to determine
whether grain boundary separation would cccur at
very high doses, as it does in BeO at doses well
velow 1 x 102! neutrons/ecm?. The experimental
techniques used in the irradiations, which are
carried out the Engineering Test Reactor
(1daho), are described elsewhere.®

The results of postirradiation examinations and

vestigation el

in

measurements of translucent a-AIZOS specimens
irradiated at 300 to 600°C over the dose range

 

s, w. Keilholtz, R. E. Moore, and M. F. Osbome,
Reactor Chem, Div. Ann. Progr. Rept. Dec, 31, 1965,
ORNL-3913, p. 1905,

’G. W, Keilholtz, J. E. Lee, Jr., and R. E. Msoore,
“Properties of Mapnesium, Aluminum, and Beryllium
Oxide Compacts Irradiated to Fasi-Neutron Doses

Greater than 1021 Neutrons em™ 2 at 150, 800, and
1100°C,** Proceedings of Joint Division Meeting of the
Materials Science and Technology Division eof the
American Nuclear Society and the Refractories Division
of the American Ceramic Society held in conjunction
with the 68th Annual Meeting of the American Ceramic
Society, May 8—11, 1966, Washington, D.C., pp. 133~48.

3(}. W. Keilholtz, J. E. f.ee, Jr., and R. E. Monore,
Nucl, Sci. Fng, 26, 320~38 (1966).

4Luca10x, trade name of a proprietary product of
General Electric Co., Cleveland, Chio.

113

0.6 to 5.2 x 102! neutrons/em? (>1 Mev) are
given below.

Gross Damage

At low irradiation doses, translucent AIZO, is
much more resistant to fracturing than sintered
A1203. Below 3 x 10%! neutrons/cm? (> 1 Mev),
there was virtually no fracturing of translucent
““203! but severe fracturing along intergranular
paths occurred at higher doses.

Metallographic Exuminations

Grain boundary separation begins to appear at
a dose of 2.3 x 107! neutrons/cm? (>1 Mev),
This separation is very extensive al a dose of
4.7 x 10%! {>1 Mev). The same degree of grain
boundary separation occurs in BeO at doses
smaller by a factor of 5 to 10. Grain boundeary
separation in translucent 1—"&1201 is probably the

o

predecessor of gross fracturing.

Dimensional Measurements
The translucent Alzo3 expanded in volume by
ahout 1.3% over the dose range 0.6 to 1.4 x 107}
neutrons/cm? (>1 Mev);, this is approximately
the same expansion observed in the cold-pressed—
sintered Al O irradiated previcusly. Above 2 x
1021 peutrons/cm? (>1 Mev), the volume expansion

 

5(30 W. Keilholtz, R. E. Moore, M. F. QOsbome, B. W.
Wieland, and A. F. Zulliger, “*Techniques forlrradiating
High Temperature Materials in a Steep Flux Gradient,®?
Irradiation Capsule Experiments, FProc. of USAEC
Conf. on Developments in Frradiation Capsule Technal-
ogy, Pleasanton, Calif., May 3--5, 1966, TIR-7697 (2d
ed.).
of translucent AIZO
with neutron dose to about 5% at a dose of 5 x
102! neutrons/cm? (>1 Mev).

increases nearly linearly

X-Ray Diffraction Examinations

The volume increase calculated from the lattice
parameter expansion was less than 1,5% over the
dose range 1.3 to 4.7 x 102! neutrons/cm?. The
anisotropic expansion ratio (AC/CO)/(Aa/aO) was
about 2 to 3, as compared with that of BeQ, which
about 20 for less than 1 x 102!
neutrons,/cm? (>1 Mev).

is doses

Grain boundary separation in BeQ compacts
irradiated at low temperatures is generally
believed to result from anisotropic lattice

patameter expansion. In the case of AIZOS, how-
ever, the lattice parameter expansion in itself
probably does not produce the grain boundary
separation, which is observed in increasing degree
as the neutron dose is increased from 2.3 to 4.7 x
102! neutrons/cm? (>1 Mev). The lattice parameter
expansion in A1203 is small and does not increase
with increasing neutron dose over this range. It is
possible, however, that defect agglomerates too
larpe to affect the lattice parameters cause an
additional crystal expansion large
enough at high doses to produce separation of

anisotropic

grains. The grain boundary separation appears tco
limit the usefulness of translucent Al O, to fast-
neutron exposures of less than 2 to 3 x 102!
neutronis/cm? (>1 Mev) at temperatures below
600°C. The neutron dose limit at higher tempera-
tures will be established when specimens now
being irradiated in high-temperature assemblies
are examined,

BEHAVYIOR OF REFRACTORY METAL
CARBIDES UNDER IRRADIATION

G. W. Keilholtz R. E. Moore
M. F. Osboine

Refractory metal carbides have potential uses
in nuclear reactors designed to operate at very
high temperatures. We investigating the
changes in the physical and mechanical properties
of these materials during exposure to fast-neuntron
doses as high as 5.4x 102! neutrons/cm? (>1 Mev)
at temperatures ranging upward to 1400°C.° The
monocarbides of Ti, Zr, Nb, Ta, and W in the

are

114

1 1

% % cylinders made by (1) hot
pressing, (2) slip casting and sintering, and (3)
explosion pressing and sintering were selected for
these studies. Low-temperature (300 to 700°C)
irradiations are now complete, a high-temperature

(~1000°C) irradiation is in progress, and as-

form of e in.

semblies for higher~temperature irradiations are
being designed. The experimental techniques used
in the irradiations are described elsewhere.’
Hot-pressed specimens and slip-cast—sintered
specimens of each of the five monocarbides
behaved very similarly in low-temperature
radiations (300 to 700°C) over the neutron dose
range 0.7 to 5.4 x 10?! neutrons/cm? (>1 Mev).

The results are summarized below.

ir-

Gross Damage

Tungsten carbide and ftitanium cartbide were
generally undamaged over the entire dose range of
the Figure 11.1 is a bar graph
showing approximate neutron dose ranges

irradiations.
the
where fracturing occurred in each of the five
carbides.

 

W. Keilholtz, M. F. Osbomes, and R. E. Moore,
Ann. Progr. Rept. Dec, 31, 195635,

ba
Reactor Chem. Div.
ORNI.-3913, p. 104,

ORNL-DWG 6610268
FRACTURING

5 SEVERE

   
  

w

N

FAST-NEUTRON DOSE
(neutrons/scme, =1 Mev)

 

|

O

ZrC TaC NbC

TYPz OF CARBIDE

TiC

Fig. 11.1. Gross Domage to Specimens of Hot-Pressed
and Slip-Cast-Sintered Refractory Metal Monocarbides in
the Form of ]/2 X ]/2 in. Solid Cylinders Irradiated at Law
Temperatures {300 to 700°C) as a Function of the Fast-

Neutron Dose.
Metallographic Examinations

No grain boundary separation could be seen in
photomicrographs of any of the carbide specimens.

Dimensional Measurements

As shown in Fig. 11.2, in which the volume ex-
pansion data are summarized, the carbides ex-
panded on exposure to a dose of 1 x 102!
neutrons/cm? (>1 Mev) and then shrank at different
rates as -the neutron dose increased to 5.4 x 10721
neutrons/cm? (>1 Mev).

X-Ray Diffraction Examinations

The volume increases calculated from the lattice
parameter changes account for only 30 to 50% of
the gross volume expansions.

Fracturing of explosion-pressed specimens with
~ 1% nickel additive generally occurred at lower
neutron doses than with the corresponding het-
pressed and slip-cast carbide specimens with
no additive, This was particularly true for
explosion-pressed titanium and tungsten carbides,
which were damaged on exposure to doses above
~72 % 102! neutrons/cm? (>1 Mev). The volume
expansion of explosion-pressed
approximately the same as that of the corresponding
carbides fabricated by the other methods,

The fellowing conclusions were drawn from the
results of the low-temperature (300 to 700°(C)
irradiations.

samples was

1. The gross volume expansion is the sum of
the lattice parameter expansion and agglomer-
ates of defects too large to affect measured
values of the lattice parameters,

2. A collapse of the large agglomerates as the

neutron dose is increased from 1 to 5 x 107!

- neutrons/cm? (>1 Mev) is the principal cause

of the shrinkage in gross volume over this
dose range.

3. The leveling effect on gross damage to the
five carbides when explosion-pressed samples
were irradiated indicates that the nickel
additive is primarily responsible for the in-
creased fracturing observed.

4. Hot-pressed and slip-cast—sintered tungsten
carbide and titanium carbide prepared without
additives are more resistant to fast-neutron
irradiation at low temperatures (<700°C) than
any of the other types of carbide specimens

    

 

 

 

 

 

 

 

 

 

investigated.
ORNIL-DWG 6E-6547R
BT ——— T T ey
Z”;EC:‘-NIUM NN O e
~2, | | ONCCARRDE
— "‘\“MTM’MNCM RBIDE
2 \ e
— > :'V/O l —————
]&}'_)] 2 3 T _:—---:3‘&»8 ““““ t """"""""" 7 -
< % Moy
oA
w T ) \*\54,?&
(2':) - ~ 0
= /l/o(’ \\‘\.\_
2 e ™~
DA e L N s
B Y
> T —
s e WO |
! . '
0 b 5
0 1 2 3 4 5 (x10%7)

FAST-NEUTRON DOSE (neutrons /emZ, >1 MeV)

Fig. 11.2. Volume Incrense of Menocarbides of Ti, Zr,
Ta, Nb, and W irradiated at Low Temperatures (300 to
700°C) as u Function of Fast-Meutron Dose.
Part IV
Other ORNL Programs

 
12. Chemical Support for the Saline Water Program

SOLUBILITY OF CALCIUM SULFATE

IN SEA SALT SOLUTIONS TO 200°C;

TEMPERATURE-SOLUBILITY LIMITS
FOR SALINE WATERS'

W. L. Marshall Ruth Slusher

In a previous study? the extensive solubility
measurements of calcium sulfate and its hydrates
in sodium chloride solutions? were used to estimate
solubility limits for these gpecies in saline waters
in general, since such estimates are of great value
to those concerned with distillation of saline
waters. The effect of divalent ions (magnesium
in particular), which can partially complex the
sulfate ion and thereby increase the solubility of
calcium sulfate over that ‘‘normally’® expected,
could not be taken into account in these estimates.
We have, therefore, measured the solubility of
calcinm sulfate (or of the dihydrate at temperatures
below 100°C) in sea salt solutions (aqueous sclu-
tions containing the salt composition of seawater
bul with varying degrees of concentration or dilu-
tion) at temperatures from 30 to 200°C and at ionic
strengths to 6 m, and have compared the results
with those obtained in sodium chloride solutions.
Figure 12.1 shows all scolubility values obtained
experimentally; in this figure the logarithm of the
solubility product, K . is plotted against a func-
tion of the formal ionic strength, /. The term _K;p
is defined as the molal solubility of calcium sulfate
times the molality of total sulfate. The analogous,
comparative behavior in sodinm chloride solutions

 

ljointly sponsored by the Office of Saline Water, U.5.
Department of the Interior, and the U.S. Atomic Energy
Commission.

*W. L. Marshall, Reactor Chem. Div. Aan. Progr, Rept.
Jan., 31, 1965, ORNL.-3789, p. 294,

9. L. Marshall, R. Slusher, and E. V. Jones, j. Chem.
Eng. Data, 9, 187 (1964).

119

is shown by the undashed lines, the experimental
data for which are given elsewhere.?'* Qur evalua-
tions of some published solubilities in sea salt

solutions?

are also included in this figure; the
comparative results appear to be in good agreement.
In the temperature interval 30 to 90°C the only
significant difference in solubility in the two
systems is observed at very high ionic strengths.
The solubilities for temperatures from 100 to 200°C
are greater in sea salt solutions, but the values
do not show a continuous (monotonic) increase
in the difference with temperature.
This behavior can be explained by the formation
of an MgSO " neutral species that allows the
molal solubility of calcium sulfate to increase
until the limiting value of the solubility product,
K, obtained from the comparative solubility in
soé’ium chloride solutions, is reached. The change
with increasing temperature in solubility between
that in sea salt (containing approximately 2 moles

increasing

of magnesium per mole of sulfate) and in sodium
chloride soluticns is then explained by an inter-
related function of an increasing association con-
stant of magnesium sulfate and the decreasing
solubility product constant of calcium sulfate.
With the dissociation quotients (at ionic strengths,
[) and constants (at [ == 0) for magnesium sulfate,
calculated from the solubility behavior shown in
Fig. 12.1 and presented elsewhere in this report,®

 

“W. L. Marshall and R. Slusher, J. Phys. Chem. 70,
4015 (1966).

5}'& Hara, Y. Tanaka, and K. Nakamura, Sendai Tohuku
fmp. Univ. 11, 199 (1934); R. Hara, X. Nakamura, and
K. Higashi, ibid., 10, 433 (1932). W. F. Langelier, D.
H. Caldwell, and W. B, Lawrence, Ind. Eng. Chem. 42,
126 (1950); An Investigation of the Solubilify of Calcium
Sultate in Seawater Concenirates at Temperatures from
Ambient to 65°C, Office of Saline Water Research and
Development Report No. 191, UU.3. Dept. of Interior
(May 1966); E. Posnjak, Am. J. Sci. 238, 559 (1940).

‘w. L. Marshall, **The Dissociation Constant of
Magnesium Sulfate to 200°C,?* Chap- 6, this report.
120

a previously developed computer method? was
revised to obtain refined temperature-solubility
limits for saline waters in general within which
precipitation of calcium sulfate or its hydrates
can be avoided. The values needed for the cal-
culations are the molalities of calcium, magnesium,
and sulfate, and the (molal) ionic strength of the

 

 

    
   

 

 

A
e

 

-2 :
107° |- | ;
I t | ST eC)
! D'D?\ =z
. ; gl N 30
SATURATING = \0\
soLip - yd 5
A0
2 .
1073 7
CoS04- 2H,0
: : O
/. - - T
7 G
// /c:;fia ‘
-4
o e CaS0,-2H,0
~ ) / .
s a ; e,
v . Cogr - ://o *, 93
- ) /fl‘/ 7/0 '
Ll Z \
/ A
s o REGION OF
107% 7 . o HYORATE INSTABILITY
-7 CaS0,-2H,0, e -
s ' :
Ksp 7
] » 100
8 Ooé/
1079 e e
7 o4
-
7
/
e
(o5 / COMPARATIVE
BEHAVIOR IN
NoCl~H20 SOL’NS.
| | IN SEA WATER CONCENTRATES:
T | & PRESENT {MARSHALL-SLUSHER),
60-200°C
s o HARA, 30-200°C (1934)
10 ' o POSNJAK, 30°C (1940}
- & LANGELIER, 100°C (1950)
. & AMF, 20-60°C (1966)
CIN HEO:
@ REFS.11,12(IN Hp0 ONLY)
wore - i .
0 0.1 0.2 0.3 0.4 0.5 0.6

N7/ (1 Asey/ 7 )

saline water, A representation of the revised
solubility limits for seawater is shown in Fig.
12.2; in this figure CF equals the concentration
factor on a molal, molar (at 25°C), or weight frac-
tion basis. If calcium and magnesium are removed
initially from a saline water, other sets of concen-

tration !imits can be obtained with ease,

1073

o

1075

   
 

Asp

   
  

1.51

/ 60 1.56
= / 95 1.59 =

100-200 1.60

 

 

10 IR ! S ] i
" DEBYE - HUCKEL 1
1 LIMITING SLOPES ... .
¢ - . i .
i | | |
/ : f :
/ ‘ i
10'? e e
0 0.1 0.2 0.3 Q.4 0.5 0.8

Fig. 12.1. Solubility Products, K;p vs \/7/(] + Asp\/.I-), of CaS0,*2H,0 ond Co50, in Seawoter Concentrates
Compared with Their Behavior in NaCl-H,0 Solutions, 30-200°C,
 

 
  
 
 

zonem.}
n

{C.Fi={sal"d. conc’n. /init.

T UNSATURATED REGIONS BEL
CURVES

COMNCENTRATION FACTCR

 

 

e et ot e . e ]

121

  

 

SEA WATER (molsl units)
1, =0709 ‘
Ca= 001034

Mg=00542 |
2763 e

|
! T
" S MoLaL
L MOLAR (25°0)

SSWEIGHT FRACTICN
(BASIS OF CONC'N.
; FACTOR) |

    
   
 
  
 

 

 

 

 

 

 

 

TEMPERATURE (°C)

Fig. 12.2.

Calculated Limits of Solution Stability to Avoid Precipitation of Gypsum (CaS50,°2H,0), Hemi-

hvdrate (C{:SOA' 1/2&‘!2‘0), and Anhydrite (CaS0,) from Seawater Concentrates. Iy = molal fonic strength of seawates;

Ca, Mg := original molalities.of Ca and Mg; X == molal ratio, 504/Cn.

CORROSION OF TIiTANIUM
IN SALINE WATER

E. G. Bohimann
J. F. Winesette

J. C. Griess, Jr.
F. A. Posey’

Crevice Attack

Titanium has excellent resistance tosaline waters,
but at temperatures of 150°C and higher it is subject
to severe corrosion damape in crevices.® Studies
to determine the cause and to find ways to mitigate
such attack are in progress,

Flectrochemical polarization studies at tempera-
tures up to 150°C showed that it was not possible
to preduce corrosion of the same magnitude as
observed in crevices when the pH of 1 M NaCl
solutions was 3 or higher; this suggests that the

 

7Chemistry Division.

j¢. C. Bohlmeann and J. C. Griess, Reacfor Chem.
Div. Ann. Proge, Rept. Jan. 31, 1965, ORNL.-3789, pp.
267305,

small volume of solution in the crevice becomes
appreciably acid (more than 107 ° M) before crevice
corrosion can occur.® To determine the approxi-
mate pH of the solution in a corroding crevice,
specimens were prepared in which the tip of a fine
titanium capillary extended into the crevice. The
capillaty tubing passed through the head of an
autoclave, and with appropriate valving a small
volume of the solution from the crevice region
could be sampled while the specimen was at 150°C.
The pH of single small drops of solution was about
1 in those cases where crevice corrosion was
observed and 4 to 5 when sipnificant corrosion
was absent. The pH of the bulk solution was 6
to 7.

Figure 12.3 shows a series of paitial anodic
polarization curves obtained potentiostatically in
a solution containing 0.1 ¥ HC{ and 0.9 M NaCl.
The curves originate at the corrosion potential,
and the potential was changed in a noble direction
at 3-min intervals. In the active region the current
density achieved a steady-state value in less than
3 min, whereas in the passive region the current
density decreased slowly for long periods. Although
not shown, the passive region extends up to the
pitting potential, which is higher the lower the
“Pitting’’).  The
increased with

temperature (see section on
critical
temperature, corresponding to an activation energy
of about 11 kcal/mole. The actual maximum cur-
rent density at 150°C corresponds to a corrosion
rate of 360 mils/year, which is close to, but less
than, the maximum corrosion rate observed in the
crevice regions (as high as 500 mils/year) at the
same temperature.

The corrosion of titaninm in crevices
confined to chloride systems.
specimens exposed to neutral aerated 1 M NaBr,
Nal, or NaQSO4 at 150°C underwent corrosion in
the crevices to about the same extent as observed
in chloride solutions. Anodic polarization curves
obtained in boiling acidified solutions of these
salts were essentially the same as those obtained
in chloride systems (though the pitting potential
is different). These results show that crevice
corrosion of titanium is not peculiar to chloride
but that it is associated with the de-
velopment of an acid environment within the crevice.

current for passivation

o
=

is not
Titanium crevice

solutions,

-0.150

-0.250

-0.350

~-0.450

-0.550

POTENTIAL {voits versus S.C.ED

-0.650

-0.750

122

The preceding results imply that a titanium alloy
resistant to dilute acid solutions at 100°C may be
immune to crevice corrosion at higher temperatures;
conversely, an alloy showing active corrosion in
dilute acid solutions at lower temperatures probably
will be susceptible to cievice attack at high tem-
peratures. It was of interest, therefore, to deter-
mine the anodic polarization characteristics of a
number of commercially available titanium alloys
and of alloys prepared in small quantities by the
Metals and Ceramics Division. The nominal com-
positions of the commercial alloys tested were:
0.15% Pd; 4% Mn—4% Al; 5% Cr—3% Al;, 4% V—-6%
Al;, 14% V-11% Cr-4% Al; 2.5% Sn--5% Al; 2%
Nb—1% Ta—8% Al; and 8% Zr--1% (Nb + Ta)-8%
Al. Alloys produced locally contained individually
1, 5, 10, 20, and 30% Mo; 0.5% Nb; 0.5% Ta; 0.5%
Cu; 1% Su; 1% Al; 2% Ni; and 1% Ni--1% Mo.
Potentiostatic anodic polarization curves were
obtained for each alloy in a solution containing
0.1 M HCl and 0.9 M NaCl at the atmospheric
boiling temperature. Of all the alloys tested, only
the 0.15% Pd alloy, those containing 5% or more

Mo, and both alloys containing nickel showed

ORNL-DWG 66-6232

 

5 100 1000

CURRENT DENSITY (microamps/cm?2)

Fig. 12.3.

Effect of Temperature on the Anodic Folarization of Titanium in a 0.9 M NaCl-0.1 M HC! Soclution.
no active region. Open-circuit potentials of freshly
abraded or pickled surfaces for the latter group
of alloys were greater than the active potential
(more than ~0.4 v vs S.C.E.), whereas all the
other alloys had polarization curves similar to that
observed with commercial purity titanium. The
critical current density for passivation, however,
varied somewhat from alloy to alioy.

Additional anodic polarization curves were ob-
tained in boiling 1 M HCl with those alloys ex-
hibiting passivity in the 0.1 # HCI-0.9 ¥ NaCl
solution.  In this environment only the alloys
containing 20 and 30% Mo and the 0.15% Pd alloy
did not develop a potential characteristic of actively
cotroding titanium. The 2% Ni alloy was passive
initially, but after exposure to the 1 ¥ HCI for
several days or affer cathodic polarization, the
alloy developed an active potential and corroded
at about the same rate as pure titanium. The 1%
Ni-1% Mo alloy exhibited borderline passivity.
Most of the time the potential of the alloy was
about —0.3 v vs S.C.E., but periodically it de-
creagsed to —0.47, at which potential the alloy
corroded with evolution of hydrogen. After a few
minutes at the latter value, the potential returned
to - 0.30 v vs 5.C.E. The time between periods
of active corrosion was 20 to 30 min.

On the basis that crevice corrosion occurs be-
cause of the production and maintenance of an
acid solution in the crevice, the alloys which
would tend to resist such attack are 2% Ni, 5% Mo,
1% Ni + 1% Mo, 10% Mo, 0.15% Pd, 20% Mo, and
30% Mo, with the last being the most resistant.
Observations on a large number of specimens
exposed in high-temperature (150 to 200°C) loop
experiments ate qualitatively consistent with this
ranking, but a satisfactory quantitative means of
assessing and demonstrating susceptibility has not
been devised,

Pitting

Titanium, in common with many other metallic
materials of construction, owes its immunity to
corresion to the existence of a passive oxide
layer at the metal-solution interface. Except in
strongly acidic solutions undet reducing coondi-
tions, titanium normally corrodes spontaneously
at a very low rate in the passive state, and the
current density of the anodic process (formation of
a Ti()2 passive layer) is independent of the value
of the electrode potential over a wide potential

123

range, However, in solutions containing a suf-
ficient concentration of chloride ions (also bromide
or iodide ions), localized breakdown of the passive
layer occurs, and titanium exhibits a pitting poten-
tial similar to that observed for aluminum, iron,
stainless steel, and other metals and alloys which
are capable of existing in both active and passive
states.

Electrochemical studies of titanium alloys in
saline waters have shown asharp inverse dependence
of the pitting potential on temperature. Thus,
whereas most such alloys require anodic potentials
of the order of 10 v or higher to initiate breakdown
of the protective oxide and consequent pitting
attack at approximately 25°C, we have found that
the potential required is only 0.5 to 2.0 v at elevated
temperatures, Details of the variation of the pitting
potential with temperature are dependent on alloy
and solution composition but have been shown to
be quite reproducible under a given set of conditions.

The etfect of temperature on the pitting potentials
of a number of titanium alloys is shown in Fig.
12.4. These data were obtained by variation of the
temperature of the 1 4 NaCl solution while the
titanium electrodes were polarized to their pitting

ORML-DWG 65-35558
TEMPERATURE (°F)

 

   
 
 
 
 
 
 
   
  

 

 

 

 

 

 

 

 

 

 

 

 

 

TH 25 TS 225 275 %25 375 425
i TWI! TVTTT
! ALLOY
10 . 50/ fil ;) o0, w TTptTTTTM
o B, 25% 5n
015%
2 045% Pd
~ 28% Cr, 0.3% Fe
1l
o 8T 08% Cr . O3% Fe
"; COMMERCIALLY PURE Ti
o TUBING
< COMMERCIALLY PURE Ti
N UG ]
<
: |
‘Z “
7 s g
}.v
[
D‘ ‘
4 b —
3 \
& i
i i
2 3 e e
< _
-d
Leh /
2 e
{
0
O 25 80 7S 100 125 150 7S 200 225 250
TEMPERATURE (°C)
Fig. 12.4. Effect of Temperature on Pitting Poten-

tiuls of Titanium Alloys in 1 M NaCl.
potentials by passage of a constant high anodic
current., Commercially pure Ti alloys and the 0.15%
Pd alloy, with small amounts of added elements,
show high pitting potentials (9 to 11 v) in the
With increasing
temperature the pitting potentials of these materials
decrease more or less uniformly to values of the
order of +1 to 2 v at 175 to 200°C. A relatively
sudden change (—2 v) in the pitting potential of
the 0.15% Pd alloy occurs in the vicinity of 125°C.
The type 150 A alloy (28% Cri, 0.3% Fe) shows a
similar sharp drop in the value of its pitting poten-

vicinity of room temperature.

tial in the vicinity of 70°C; the pitting potential
changes about 2.5 v over a temperature range of
only a few degrees. Pitting potentials of the type
110-AT alloy (5.5% Al, 2.5% Sn) are generally low
and attain values of only about +0.4 v at 190°C.

Qualitatively, the curves of Fig. 12.4 show that
in general the addition of alloying elements de-
creases the value of the pitting potential of tita-
nium. This is consistent with observations of
Fischer? on effects of various surface treatments
and of intentionally added impurities on the initia-
tion of titanium pitting corrosion. One way to
rationalize these differences in pitting potentials
of titanium alloys is to suppose that the presence
of alloying elements affects the ionic as well as
the electronic conductivity of the protective oxide
layer. It is known that the presence of certain
impurities or lattice defects considerably enhances
the electronic conductivity of rutile (TiOz) crys-

tals. 1%~ 13  Apalogous effects on the kinetics of

124

the anodic process have been observed as a result
of incorporation of alloying elements into the Ti0O
lattice,!*4/1%  In addition to direct effects on

anodic reaction mechanisms, the presence of

additional electronic energy levels in a semi-
conductor of the type in question and the resulting
enhanced electronic conductivity imply a simul-
taneous increase in hole concentration in the
valence band (weakened bonds), so that a further
increase in the mobility of ionic charge carriers
can be expected.

An analysis based on a simple resistance analogy
suggests that differences in the mobility of ionic
charge carriers, which affect the kinetics of the
anodic processes occurring in the passive oxide
film, can account for observed variation in values
of pitting potentials as a function of alloy com-
position.

 

*W. R. Fischer, Tech. Mitt. Krupp, Forsch. Ber. 22(3),
65-82 (1964).

IOK. Hauffe, H. Grunewald, and R. Tranckler-Greese,
Z. Elekirochem. 56(10Q), 93744 (1952).

llR. G. Breckenridge and W. R. Hosler, Phys. Rev.
21(4), 793--802 (1953).

12(. Kleber, H. Pecibst, and W, Schréder, Z. Physik.
Chem. (Leipzig) 215, 63—-76 (1960).

131, E. Hollander, Jr., and P. L. Castro, Phys., Rev.
119(6), 188285 (1960).
14g. A. Grant, RRev. Mod. Phys. 31(3), 64674 (1959).

15j. Maserjian, Conduction Through Thin Titanium
Dioxide Films, Jet Propulsion Laboratory Technical
Report No. 32—-976, October 1966,
13. Effects of Radiation on Organic Materials

W. W, Parkinson

EFFECTS OF RADIATION ON POLYMERS

W. W. Parkinson W. K. Kirkland

Radiation~induced processes in polybutadiene
are significant because the polymer has a simple
hydrocarbon structure related to natural and many
of the synthetic rubbers and because the olefin
groups of this polymer undergo rapid changes
upon irradiaticn.  Furthermore, types of poly-
butadiene are obtainable having a high fraction of
the olefin groups in any one of the three possible
igsomeric forms. These forms are cis, trans, and
side vinyl, as shown:

¥ C——~— -~C H e G

. N |

H H H \C"‘" ™

i
Hy CH2
Cis TRANS SIDE VINYL
Changes in these groups may be studied by
infrared spectra, since each has at least one

absorption band at a characteristic {requency:
cis at 740 em™!, trans at 967, and side vinyl at
010. Previous reports described infrared spectral
measurements on polybutadiene specimens of
various types irradiated in °°Co gamma sources

and in a nuclear reactor.’'? The four types

 

IW. W. Parkinson et al., Reacfor Chem. Div. Ann.
Progr., Rept, fan. 31, 1964, ORNL-3591, p. 226.

§. W. Parkinson and W. . Sears, ‘'The Effects of
Radiation on the Olefinic Groups in Polybutadiene,®?
presented at the American Chemical Society Meeting,
Mar. 29, 1966, Pittsburgh, Pa.; in press, Advances in
Chemistiy Series (1667).

Q. Sisman

studied were high cis (82%), high trans (95%),
and two mixed polymers, one 73% trans and the
other 71% side vinyl.

To derive actual concentrations of the olefin
groups from the spectral measurements, it proved
necessary to  use the unirradiated specimens
themselves as calibration standards. Calibration
curves developed from liguid olefins proved
inapplicable. However, they were adequate to
determine the small content of c¢is groups in the
two mixed polymers. Since the total unsaturation
was known, this pemmitted the solution of simultee
neous equations relating the concentration of
trans and side vinyl groups to the optical
densities at 967 and 910 ecm™!. The absorptivities
thus calculated, coupled with the accepted value
for the total unsaturation in high-cis polybutadiense,
enabled us to determine the cis concentration in
the polymer of this type from its spectrum. The
integrated band area at 740 cm™*', calculated from
theoretical band shapes,®*? was required for the
calibration and concentration determination of the
cis group because of the dependence of the simple
absorptivity on groups adjacent to the cis species
in the molecule, '

From this treatment of the infrared spectra, the
concentrations of olefin groups were obtained for
specimens irradiated undet wvarious conditions.
The changes in these groups for high-cis poly-
butadiene are plotted vs dose in Fig. 13.1. The
decrease in cis groups at the lower doses appears
to be independent of their concentration {(zero-
order), whereas the decrease in trans and side
vinyl groups in the other types of polybutadiene
followed first-order kinetics. In addition, the high-

 

125

8D. A. Ramsay, J. Am. Chem. Soc, 74, 72 (1952).
%S, A. Francis, J. Chem. Phys. 19, 942 (1951).
cis polymer was the only type showing an in-
crease in concentration of one of the olefin iso-
meric forms. In this polymer there was a conver-
sion of cis groups to the trans forimn, the more

stable isomer thermodynamically. The radiation

ORNL—OWG 66—3040R

 

 

 

 

20 : ‘ -
= .0 CIS
$ v TRANS
= ) ‘ a6 SIDE VINYL 4 =
2R " U OPEN SQINTS -GAMMA (RRADIATION L=
£ N CLOSED POINTS -REACTOR IRRADIATION £
— . : A
= \\ P & e, A 13 - t;'l;
o 2o // : Ce
— * + - .- =
g | ‘ a
x ! w2
- ; 2 o
S i g Q=
o ; ~ Z =
O a4
z - i x
3 1 2%
U — ]
0 T Q
O 8 T . g -3

a ‘ T —t ——emA . | o O

o 5 10 15 20 25 30 (k)

DOSE (rads)

Fig. 13.1.
diated Cis Polybutadiene.

Concentration of Qlefin Groups in liro-

Tabie 13.1.

126

yields, rate constants, and activation energies
calculated for all the polymers from these changes
in concentration are listed in Table 13.1.

The activation energies were calculated from
radiation yields at room temperature and at 110°C.
They are similar to activation energies in other
radiation-induced reactions and indicate that low-
energy processes such as the addition of radicals
and the diffusion of relatively small species are
the rate-controlling steps.

The high yields and high rate constants indicate
efficient transfer of radiation energy from the
points of absorption to reactive sites, either
through charge or excitation transfer. It is also
possible that the high rate constants observed for
the side viny! groups result from short chain
reactions.

The products resulting from these reactions are
unidentified as yet. Because the cross-link yields
are far too low to account for the loss in olefin
groups, it has been speculated that intramolecular

Radiation Yields and Reaction Rate Constants

 

Type of Polybutadiene

 

 

High High Amorphous Side Vinyl
Cis=1,4 Trans-1,4 Trans (Sodium)
(Emulsion)
Cis Groups
Initial yield, GO’ groups per 100 ev ~15.2 1 0.8 <+ 1.0 <+ 0.5 0
Activation energy, E, kcal /mole 3.7
Reactor yield, GO, groups per 100 ev = —12
Trons Groups
— — -2
Rate constant, k, g/ev (1.4-3.3) x 10 23 2.2 x 10 23 2.4 x 10 3
Initial yield, GO' groups per 100 ev +6.8 —11 to —-22 ~-18 -7
Activation energy, E, kcal/mole 1.5 + 0.3 3.4 4.0
Reacter yield, GO’ groups per 100 ev +4.4
Reactor rate constant, k, g/ev 0.9 x 10723 ~1x 10723
Side Vinyl Groups
Rate constant, k, g/ev 0. 1.4 0.5 % 10723 4.3x 1023 5.4 x 10723
Initial yield, GD’ groups per 100 ev 0 —~0.2 ~12 ~40
Activation energy, I%, kcal /mole 3.8 3.9
Reactor rate constant, k, g/ev 1.6 x 1023 ~1 % 10723
Met Change, All Groups (Gamma Radiation)
~8.4 ~11 to —-22 -31 —47

Initial yield, GO’ groups per 100 ev

 
ring links are formed, but these have not been
identified definitely in irradiated polybutadiene
because of difficulties in both spectral and chem-
ical observations of such cyclic structures.?:®

RADIATION-INDUCED REACTIONS
COF HYDROCARBONS

R. M. Keyser W. K. Kirkland

The irradiation of coal together with a condensed-
phase source of hydrogen atoms and alkyl free
radicals offers the possibility of the synthesis of
useful hydrocarbon chemicals from inexpensive
raw materials in @ chemonuclear reactor. Initial
jirradiations have been confined to a model system
for experimental convenience: n-hexane (simulating
liquefied petroleum gases) the condensed source
of hydrogen and radicals, with naphthalene simu-
lating coal. It is expected that the overall course
of the reaction, the addition of radiation-generated
hydrogen atoms and alkyl free radicals to aromatic
rings,® will be similar in both the model and coal
systems.

     
       
   

 

 

 

 

 

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100 & e Ll Eow [ ¥ B .
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127

Samples irradiated with ®°Co gamma radiation
have been analyzed by gas chromatography on a
5-m column containing Apiezon L as the separating
medium. The results of a typical run are shown in
Fig. 13.2, from which it is apparent that irradiation
of naphthalene-hexane solutions yields a quite
complex array of products. Efforts to date have
been directed toward identification of the com-
pounds responsible for the peaks in Fig. 13.2.

The six dimers of hexane, the peaks labeled
4,5-diethyloctane through n-dodecane in Fig. 13.2,
have been identified by using chromatographic
retention times to obtain estimates of the boiling
points of the compounds in question and comparing
these with literature values. Apiezon L., a nonpo-
lar liquid, separates saturated hydrocarbons es-
sentially on the basis of their boiling points. It

 

SM. A. Golub, J. Phys. Chem. 69, 2639 (1965).

bW. W. Parkinson et al., Reactor Chem. Div., Ann,
Progr., Rept, Jan, 31, 1965, ORNL-3789, p. 320,

ORNL-DWG &7-542

 

 

 

 

 

 

I L S B -l L. J

 

225 235 et——o ISOTHERMAL AT 235 — i

TEMPERATURE (¢C)

Fig. 13.2. Gas Chromatogram on a 5-m Apiezon L Column of a Sample Consisting of 0.10 Mole Fraction of

Naphthalene in n-MHexane Irradiated to 0 Dose of 1.0 1022 o\ - g

-1
128

can be shown, from thermodynamics, that the fol-
lowing relation should hold:’

log t, =K+ (4.5/T) T, (13.1)
where ¢ is retention time, T is gas cliromatograph
column temperature, T, is boiling point, and K
can be regarded as a constant provided the operat-
ing parameters of the chromatograph are closely
duplicated from run to run.

Values of log t for standard samples of n-
heptane, n-decane, and n-dodecane were plotted
against their respective boiling points, and the
relationship predicted by Eq. (13.1) was found to
hold. Boiling points of the compounds correspond-
ing to the peaks indicated in Fig. 13.2 were then

 

7See, for example, J. H. Knox, Gas Chromatography,
chap. 2, Wiley, New York, 1862.

obtained from their respective retention times and
the log t vs T, plot. The boiling points obtained
in this way agreed within 1 to 2° of the literature
values for the hexane dimers.

The identification of one of the radiation products
as n-dodecane was further confirmed by the coin-
cidence of its retention time with that of a standard
sample of n-dodecane. This same technique was
also used to tentatively identify the peak preceding
naphthalene as 1,2-dihydronaphthalene,

Compounds corresponding to peaks 1 to 8 in the
chromatogram of Fig. 13.2 have been individually
trapped in glass capillaries at liquid-nitrogen tem-
perature upon elution from the chromatograph.
These collected fractions were used to obtain
ultraviolet absorption spectra in cyclohexane solu-
tion of compounds 1 through 8. Some typical
results are shown in Fig. 13.3 for the 240- to 330-
my region. Differences in overall intensities are

ORNL-DWG $57-543

 

2.0 — o

 

OPTICAL DENSITY

 

    

i

; l i

\ ‘ | |
!

\

|

 

1

|

2-METHYLNAPHTHALENE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

330

WAVELENGTH (mp)

Fig. 13.3. Ultraviolet Absorption Specira of Compounds Corresponding te Peaks 2, 3, and 8 in the Chromatogram

of Fig. 13.2,

Solvent is cyclohexane and concentrations are different for each compound.
not significant, since the spectra were obtained
at different concentrations.

The uv spectrum of a-methyvlnaphthalene, in-
cluded in Fig. 13.3 for comparison, exhibits absorp-
tion bands in these same regions also, and the
comparison strongly suggests that compounds 1,
2, 3,4,6, and 8 are c-substituted alkyInaphthalenes.
mince the chromatographic retention time of a-
methylnaphthalene is less than that of compound
1, the alkyl substituents must be larger than methyl.
The uv spectra of compounds 5 and 7 indicate
that these compounds are f-substituted alkyl-
naphthalenes.

Additional amounts of these compounds are being
trapped from the chromatograph effluent, and it is
planned to use these to obtain mass, infrared, and
nuclear magnetic resonance spectra, which, in
conjunction with the uv data, should enable us to
make a positive identification of the unknowns,

ADDITION REACTIONS
OF FURAN DERIVATIVES

C. 1D, Bopp W, W, Parkinson

The radiation-induced addition of saturated furan
derivatives to olefinic groups is being studied to
develop high-yield reactions which could utilize
radioisotopes in chemical synthesis. An earlier
survey ® covered chiefly monoolefinic and saturated
furan derivatives and eyclohexene. The results
were promising in that the array of products was
limited; there were only three or four major products,
consisting of dimers, 1:1 adducts, and a non-
volatile residue probably comprising trimers and
higher adducts. Furthermore, the yield of major
products was enhanced by irradiation at higher
temperatures, while the yield of minor products
was reduced. On the other hand, none of the
mixtures containing saturated furan compounds
gave the high vields characteristic of chain reac-
tions.

The system cyclohexene-tetrahydrofuran has
been investigated in detail, since the radiation-
induced reacticns of cyclohexene are better known
than those of the other reagents, Solutions of
cyclohexene in tetrahydrofuran in concentrations
of 1:3 and 1:7 were irradiated at room tempera-

 

5¢. D. Bopp et al., Reactor Chem. Div. Ann. Progr.
Rept, Dec, 31, 1965, ORNL.~3913, p- 123.

129

ture, and solutions of 1:3 councentration were
irradiated at 150 and 300°C. The irradiated mix-
tures were subjected to chromatographic analysis
on a butanediol succinate polar column and on a
silicone rubber nonpolar column. At room tempera-
ture there were four products in the range of dimers
and 1:1 adducts, plus the nonvolatile residue.
At the higher temperatures one of the products
disappeared.

To identify the radiation products observed by
chromatographic analysis, compounds which were
probable products were synthesized by conven-
tional chemical methods. The chromatographic
retention times of the synthesized compounds were
then matched with those of the radiation products
on columns of at least two types.

The methods used for preparation of the dimeric
compounds were Wurtz-type condensations of or-
ganic halides with active zinc-copper and con-
densations with CHaMgI. For adducts the syn-
thetic method was the Grignard reaction, condensing
the cyclohexeny! ot furanyl magnesium halide with
the proper halide. The halides in all these reac-
tions were prepared by accepted methods. For
example, 2-chlorotetrahydrofuran was obtained by
direct chlorination of tetrahydrofuran at —56°C;
allyl-type bromides, such as 3-bromocyclohexene
and 2-bromo-2,5-dihydrofuran, were obtained by
bromination with a suspension of N-bromosuc-
cinimide in carbon tetrachloride. By these methods
the ten possible allyl-type or alpha isomers of the
dimers and 1:1 adducts of cyclobexane, cyclo-
hexene,
prepared.

tetrahydrofuran, and dihydrofuran were

By comparing retention times of these synthetic
products with those of the irradiated mixture, the
compounds listed in Table 13.2 were tentatively
identified as the major radiation products. The
identification remains tentative because of the
presence of the products of side reactions in the
synihesized compounds. The synthesized com-
pounds are now being isolated, Ly distillation
and extraction, for infrared spectra to eliminate
uncertainties. Chromatographic fractions are also
being collected from the radiation products for
similar examination to confirm our identifications.

The tabulated product yields were obtained from
chromatographic peak areas utilizing dodecane as
an internal standard. Evaporation of the irradiated
mixture indicated that the yield of polymeric non-
volatile residue was pgreater than G = 6, probabiy
exceeding the combined yields of dimers and 1:1
Table 13.2. Radiation Products from Cyclohexene-
Tetrahydrofuran Solutions

 

 

FProduct Yield (moleculefi per 10Qev)
25°C 150°C  300°C
1,79 1:37 139
2(Cyclohexene-3-yi)- @m ~ P ~ & ~8
telrahydrofuran

2,2"-Diltetrahydrofuran) [ouoj ~ 15 ~ 3 ~5
3,3-Dicyclohexenyl @@ ~05 n ~2
~ 0 0

2,2"-Di{dihydrofuran}(?) @_l?]

 

“Concentration of cyclohexene in tetrahydrofuran by voiume.

adducts. Additional derivatives of furan and fur-
fural are being irradiated to explore the nature of
the reactions and to look for the high yields of
chain reactions.

DEVELOPMENT OF RADIATION-RESISTANT
INSULATORS

W. W. Parkinson B. J. Sturm
E. J. Kennedy”

A program has been initiated under the sponsor-
ship of the Office of Civil Defense to develop
radiation-resistant insulators for personnel dosim-
eters. Since the dosimeters are small fiber elec-
trometers, the electrical resistance of the construc-
tion materials must be very high, both before and
after irradiation.

A copolymer of styrene and a-methylstyrene with
a moderate impurity content has been found to have
high resistance before irradiation and extremely
low postirradiation conductivity. Objectives of the
program, then, are to correlate the electrical con-
ductivity of plastics with the molecular structure

 

9Instrument.a*tion and Controls Division.

130

of the base polymers and also with the chemical
nature of the impurities, since these are known to
affect the conductivity prior to irradiation and
appear to play a part in reducing the postirradiation
conduction.

The correlations should permit the selection of
materials andthe development of fabrication methods
suitable for both the bulk insulators and the capac-
itor dielectric in the dosimeters. We also hope to
identify the mechanism of the impurity effects; it
could be a suirface process or a process in the bulk
of the polymer involving supplying and trapping
charge carriers.

Analytical work on styrene-base plastics has
indicated that the content of unpolymerized styrene
varies widely in these materials. The copolymer
of low conductivity showed a monomer content of
~0.1%. The solvent-cast film (the form for capac-
itor dielectric and specimen material for resistance
measurement) showed less than 0.01% monomer.
Since the unsaturated nature of styrene monomer
makes it chemically different from the polymer,
the role of the monomer in electrical processes
inn the plastic will be investigated early in the
program.

A vibrating-reed electrometer has been modified
slightly to permit electrical measurements in the
required low-current range. Tests have indicated
that its minimum sensitivity is of the order of
10~ '% amp. This will be suitable for postirradia-
tion measurements and at least some of the pre-
irradiation measurements.

The procurement of sample materials has been
confined to simple polymers, either hydiocarbons
or polymers of carbon, hydrogen, and either oxygen
or nitrogen, and to materials of relatively high
purity, with the emphasis on the glassy, amorphous
plastics,
ample,

Common commercial additives, for ex-
ultraviolet stabilizers and antioxidants,
have also been obtained.

Measurements and irradiations will be made first
on polymers of simple chemical and physical
structure in the hope of obtaining data amenable
to interpretation in terms of the basic composition
of the material.
14. Chemical Support for the Controlled
Thermonuclear Program

R. A. Strehlow

INTERPRETATION OF DCX-2 MASS SPECTRA

Twenty DCX-2 mass spectra obtained during a
single day were subjected to detailed examination.
Four types of impurity pases (viz., not hydrogen)
were distinguished by peak height comparisons.
These four types were:

1. those species generated or evolved near the
spectrometer,

2. the so-called base pressure gases,

3. impurity gases introduced with intentionally
admitted gas,

4, impurity gases generated during beam injection.

As noted earlier,! the spectrometer species de-
cayed rapidly for 30 to 50 min after the spectrom-
eter filament was turned on. Those species,
primarily organic, appatently have no significant
relation to gases impinging on the DCX-2 plasma.
Quantitative interpretation of the mass spectra is
complicated by their presence.

Of far more significance is the behavior of the
base-pressure species other than hydrogen. Air
leaks have been only occasionally responsible for
a significant part of the background gases. Water
vapor, carbon dioxide, and methane were the prin-
cipal identifiable species in the background gas
on the day of the study (August 11, 1966). Water
vapor constituted about three-quarters of the im-
purity background gases. Although the water par-
tial pressure increased during admission of hydro-
pen, it was possible to notice the variation of
this component of the background during the day.
This variation is shown in Fig. 14.1. This domi-

 

lrhermonuclear Div. Semiann. Progr. Rept. Apr. 30,
1966, ORNL.-3989, pp. 128-32.

131

D. M. Richardson

nant change, except during a period of probe
adjustment at noon, indicates that the water off-
gassing rate increased during the moming and
presumably was a themal! effect. Because of
the dominance of water vapor and its slow removal
from unbaked vacuum systems, some of the param-
eters affecting water behavior (in unbaked stain-
less steel systems) have been separately studied.?
The partial pressure of carbon dioxide has been
generally observed to follow that of water vapor.
The average carbon dioxide partial pressure ap-
peared to be about 15% of the water vapor pressure
during the day. Methane (present during the tita-
nium evaporation) was about 5% of the water par-
tial pressure, Other organic species and carbon
monoxide comprised the balance of the base-pres-
sure gases, but were not studied.

Of the gases which are introduced along with
intentionally admitted gas, water vapor and air
have been observed. Air from leaks in the gas
manifold is present in amounts which depend upon
the length of time subsequent to manifold evacu-
ation, as well as the leak rate. Water vapor is
usually the dominant manifold impurity and was
so for this study. The water partial pressure
variation with ion-gage reading for a hydrogen leak
is shown in Fig. 14.2. The linearity of the data
points and the times involved in the exposure in-
dicated that this increase was not an artifact of
the spectrometer but was a net increase of water
partial pressure admitted with or due to admission
of the hydrogen. The water pressure was observed
to vary only slightly during beam injection. It
was therefore concluded not to be a gas of the
fourth type (gases produced during beam injection).

 

2$ee subsection “Water Vapor Chemisorption on Stain-
less Steel,’’ this section.
ORNL-DWG 67-779

 

 

 

    

£SSURE, H,0 (x107 torr)

 

   

|
|
|
|
‘
‘
|
!
|
|

 

 

 

 

132

o [
0. ‘ !
o |
=
= |
O |
(Cg PROBE ADJUSTMENT —
<
© ;
<L i ‘
o |
o o —
x | |
0w |
o 1 |
O e i ;
4 2 3 4 5 6 7 8
TIME (hours after BAM, 8-11-66)
Fig. 14.1. Pu_a in DCX-2 (8-11-66) as a Function of
2

Time (Bockground lmpuiity).

Figure 14.3 shows the variation of identifiable
species with ion-gage recading during beam injec-
The fractional contribution of these species
is shown in parentheses. Water, though not pro-
duced by the injection process, is shown for com-
parison. The variation of mass 28 (primarily from
cthane, ethylene, carbon dioxide, carbon monoxide,
and nitrogen) after adjusting for CO, and ethane
appeared, from consideration of fragmentation pat-
ters, to be primarily carbon monoxide and ethylene
in about equal amounts. The variation of mass
28 so adjusted contributed slightly less to the
total pressure than did methane. A fractional
abundance of 0.003 to 0.004 for each of these
gases relative to hydrogen was indicated. This
behavior appeared to be typical for the mode of
DCX-2 operation at the time this work was done.

tion.

The principal organic species identified as being
produced during beam injection are methane, ethyl-
ene, acetylene, and ethane. The partial pressure
of these gases was found for one day to be about

2.5% of the indicated pressure.

For many plasma experiments in DCX-2, it has
not been as necessary to maintain a high-purity
environment as to have some knowledge of the
purity level. A purity index can be defined as

CRNL-DWG 57-780

 

 

 

 

 

 

~

p,;nd(MOOTorr)

 

 

 

 

 

 

 

 

 

-
/_\,OHQO (x40 torr)

Fig. 14.2. Increuse of py, with H, Bleed (Manifold
70 2
Impurity).

ORNL-DWG &67-784

 
   

 

 

 

 

 

 

 

 

 

I
| .
15%
‘7 4
i o
I . <
| ‘ /‘. T
| - eogE
,‘, t - _— ' / ......... —_——
i ! ‘
| i
" Hy0(0.045) e — {05
- i
—1 |
o . i | Lo
0 2 4 6 8 10 12 14

£ limpurity) (x10% tarr)

Fig. 14.3. Partial Pressures of CHA' COZ’ C2H2 and
C2Hfi (Produced Impurities) in DCX-2 (8-11-66) During
Injection at Different H2 Partial Pressures. py (see

20

Fig 14.2) is shown for comparison.
or0,/(Prgray — Pr,)- This index (P.L) is approxi-
mately given by
2.2(p.

1

ag ~ Pp) 1 =%

Xpind + pB’

 

’

where
2.2 is the ion-gage factor for H,

Pinq is the indicated ion-gage pressure,

p,- is the partial pressure of “base’’ pressure
contaminants other than H,,

¥ is the fraction of the ion-gage reading (p; ,)
due to contaminants produced by the injection,
introduced with bleed gas, or pgenerated by
these species (a gage factor of unity is as-
sumed).

For the data described here, using the values

ind ° 2 % 107° torr and 9 x 107° terr,
ppr=3.5x 107" torr,

B

x = 0.057 (based on the mass spectral analy-
sis),

one obtains, for the two indicated pressures, val-
ues of the purity index of 7.0 and 20.8, respec-
tively, for the two pressure levels. The data
obtained on August 1l yielded the purity index
variations shown in Fig. 14.4. These data were
obtained over a period of several hours in the
merning when the water partial pressure was in-
creasing as discussed above.  The points at
indicated pressures of 1.0 and 1.3 x 107" torr,
which were obtained later, were, accordingly, some-
what lower than the values interpolated from the
eatlier detemminations. The agreement of the cal-
culated values of P.I. with those shown in Fig.
14.4 indicates that unidentified species are not
present in significant amounts,

Residual gas analysis strongly depends on the
useful mass range of the spectrometer, the exist-
ence of identifiable parent mass peakg, and a
knowledge of fragmentation pattems. We have
little knowledge of species with molecular weights
higher than 44 u for the DCX-2 spectrometer, or
higher than about 92 u for the instrument used in
the calibration and chemistry studies. A spectrom-
eter design study has begun with the goal of im-
proving the analytical capability in the ORNL
mirror machines.

133

ORNL-OWG 67~782

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

/_—‘-.m
s
e ]!
—
o
0 a-- e
0 2 4 S B 10
Patx 10° torr)
Fig. 14,4. Purity Index YVariotioen (DCX-2, 8-11.68) vs

Uncorrected lon Gage Reading.

MASS SPECTROMETER CALIBRATION STUDIES

Dispersion of mass spectrometer ion beams af-
fects both the cracking patterns and the relative
partial pressure sensitivities for different gases.
Calibration studies carried out using the type of
mass spectrometer used on DCX-2 (Veeco model
No. RGA-3) showed that the instrument has a sig-
nificant degree of mass dispersion. This dis-
persion is attrbutable to space-charge effects and
to thermal velocities in the direction parallel to
the magnetic field during the ion’s life.
Space-charge effects at higher pressures were
observed as a diminution of ion intensities upon
increasing the hydrogen pressure. As expected,
this effect is most pronounced for gases of higher
molecular weight (e.g., argon and krypton). Hydro-
gen pressures of as much as 2 to 3 x 1077 torr
were needed before a significant decrease of the
28 and 18 mass peaks occurred.

At lower pressures the mass discrimination can
be attributed principally to thermal drift in the
direction of the magnetic field during the flight
time of ions to the collector. Estimation of this
effect for the analyzer region only, using as
parameters the flight path (23 cm), the ion velocity
derived from the ion energy meter, the average
thermal velocity (at 100°(C), and relative spectrom-
eter slit lengths, leads to the estimated transmis-
sions listed in Table 14.1 along with the trans-
mission as determined from experimental data
described below. The observed transmission frac-
tions were uniformly lower than those estimated.
This is presumably due to mass discrimination in
the source. For this 60° spectrometer (supplied
with almost equal source and collector slit lengths),
it should be possible to decrease the mass dis-
crimination of the analyzer region by perhaps 60%,

Table 14.1, Relative Transmissions for the RGA-3

Spectreameter

E.xperimental
Estimated Fraction

 

Fraction
E Species Transmitted, Transmitted,
e Analyzer Region Entire Flight
Only Path
12 CcO 1.0 1.0
18 HQO 0.9 0.8
28 CO 0.7 =0.4 to 0.5
44 C02 0.5 0.2
78 C6HG 0.2
2 Low mass 0.98
range, Hza
18 T.ow mass 0.1 =0.1
range, HZO

but probably with little overall improvement of this
parameter because of the source contribution.

The method used to experimentally determine the
extent of mass discrimination was simply to meas-
ure the peak height ratios for argon (+1, +2, and
+3) and compate these with the literature values.?
(The electron encrgy for the RGA-3 is 145 1 4 v.)
The same technique applied to the spectrometer
on DCX-2 vielded similar results. The data shown
in Table 14.1 were estimated using a smooth curve
drawn through the points at M/e = 13.3, 20, and
40. Relative calibration of the spectrometer to
an ion gage for nitrogen and water vapor yielded
additional points within 15 and 5%, respectively,
of the derived argon transmission curve. The ion-
gage sensitivity factor for water vapor was inci-
dentally determined from measured flow rates,
assuming the pumping speed for water vapor to
of the trap). This value of the gage sensitivity of
water vapor was found to be 0.9 1 (.03 that of
nitrogen.

These data were used in the measurement of con-
taminant levels in the DCX-2 work described above.
These results should not be indiscriminately ap-
plied to other similar analyzers, because the extent
of mass discrimination is strongly affected by
operating parameters such as repeller potential.
In order to avoid overly harsh evaluation two things
should be noted. The first is that residual gas
analyzers are designed primarily for qualitative
assessment of vacuum conditions and that their
application to quantitative work is usually re-
stricted to individual mass peaks or simple ratios.
The second is that the analytical capability is
restricted generally to masses less than 44 u.
The spectrometer model used in our studies per-
forms these tasks well.
tative data, however,
instrument’s capability.

The need for more quanti-
does markedly strain the

WATER VAPOR CHEMISORPTION
ON STAIMLESS STEEL

The dominance of water vapor impurity in DCX-2
and the lack of information about its sorption
kinetics prompted a study of water sorption at
low partial pressures. The results of the study

 

l.ow masses are obtained by shunting the magnet,
which reduces the field by a factor of about 0.55.

%H. 8. W. Massey and E. H. S. Burhop, Electronic and
Ionic Impact Phenomena, p. 38, Oxford, 1956,
135

showed distinctly that chemisorption occurred in
the experimental vacuum system. The experiments
were conducted in a manner to give assurance
that the observed chemisorption occurred primarily
on stainless steel surfaces. Thus, since a large
body of informaticn on chemisorption exists,* the
description of water vapor behavior in vacuum
systems can be greatly simplified. Specifically,
a selection from empirical relations between the
extent of sorption and time, exposure, pressure,
or temperature which are observed for other chemi-
sorption reactions may be applied to studies of
water vapor.

Chemisorption may be digtinguished from physi-
cal adsorption using several criteria, including the
following:

1. Chemisorption requires an appreciable (>0.1
ev) heat of activation.

2. It usually involves sorption heat values greater
than about 0.5 to 1.0 ev, whereas physical
adsorption is usually associated with smaller
values.

3. Chemisorption or chemidesorption iz (as a con-
sequence of criterion 1) a slow process.

4. The quantity of substance chemisorbed often is
related to time by the empirical relation:

fi - pe” (1)
dt
where ¢ is the amount sorbed at time t and where
a and @ are constant during any single experiment
or at the very least have discontinuous derivatives
with respect to time during a single experiment.
Water chemisorption studies have been made for
Tho,, %% ALO_,>7 Ti0,,%? and silica'® sor-
bents, principally calorimetrically. (Infrared, gravi-
metric, and pressure-change techniques have also

 

*The review article by M. J. D. Low with 342 refer-
ences is especially recommended; Chem. Rev. 60, 267 -
312 (1960).

5M. E. Winfield, Australian J. Chem. 6, 221 (1953),

54, F. Holmes, T.. L. Fuller, and C. H. Secoy, J.
Phys. Chem. 70, 436 (1966).

"R. L. Venable, W. H. Wade, and N. Hackerman, J.
Phys. Chem. 69, 317 (1965).

aC. M. Hollabaugh and J. J. Chessick, J. Phys. Chem.
65, 109 (1961).

. D. Haskins and G. Jura, J. Am. Chem. Soc. 66,
919 (1044).

109 w. Wnalen, J. Phys. Chem. 65, 1876 (1961).

been used.) No report has been found in the
literature on water sorption kinetics at low pres-
sures (<107* torr) or on metals.

The technique used in this study was to observe
the pressure fall after water vapor exposures at
different pressures and durations in a dynamically
pumped vacuum system (Fig. 14.5). The apparatus
was designed to have two regions (A and B) which
could be separately heated. To minimize tempera-
ture nonuniformity, region 4, a steel tube with
1.1 x 10* cm? area, was baked by resistance
heaters strapped to the outside of an aluminum
pipe which fitted concentrically around the steel
tube with a uniform l/g—in, gap. Insulation material
and aluminum foil were applied around the alumi-
num pipe. A copper tubing cooling circuit was
strapped to the steel tube in order to achieve a
rapid cooling capability. Region B, down to below
the conductance-limiting baffle, was heated using
strapped-on and serpentine heaters, and a hot air
assembly to heat the trap. The mass spectrometer
(Veeco RGA-3) was heated with tape heaters,
Glass jon gages (used only for early experiments)
were heated by heat lamps. Only metal seals
were used.

The water was admitted through a valve from a
regulated pressure of 1 or 2 torrs, The gas line
was perodically baked and operated with cold
water cooling to minimize organic contamination.
Water vapor with less than 0.03% total organic

ORNL -DWG 67-783
BAKABLE REGIONS

 

 

COMDUCTANCE
LIMITING BAFFLE —

 

Fig. 14.5.
Water Yapor Chemisorption.

Vocuum System Used for the Studies of
contamination could be admitted to the system.
Two temperature conditions were studied. The
elevated temperature condition was 180 + 20°C
(as monitored by several thermocouples) for all
regions of the apparatus. The lower temperature
condition was reached following a rapid cooling
of region A to 28°C. The temperature of region
B was not changed. At the elevated temperature,
typical partial pressures in torrs were 3 x 1073
H,, 4 to8 x 1010 H,0, 1 to 2 x 1077 CO (pro-
duced almost entirely by gage and spectrometer
filaments). No effort was made to reduce the hy-

 
  
    
     
  
  
 
 

M

 

PARTIAL PRESSURE OF WATER (torr)
3
4

w

1078

 

|
|
-

136

drogen pressure by higher-temperature baking since
the gas under study was water.

Water was admitted for periods of time ranging
from about 0.5 sec to as long as 30 min, after
which the valve was closed and the exhaust curve
was obtained. The partial pressure of water was
followed by monitoring mass 18 with the spectrom-
eter. As was expected, agreement with glass ion
gages was good only when the gages ware heated.
For most of the determinations, however, to mini-
mize the possibility that the glass might be affect-
ing the results, they were replaced by nude gages.

TEMPERATURE ’exp (sec)
1 180 05 ‘
) | 28 05 -
I m 180 600
‘ v 28 600

V EXPONENTIAL EXHAUST - !

cofor o |-
A

    

 

 

 

‘
1079 L mf .....
c.2

0.5 1

Fig. 14.6. Typical Logp

H20

 

TIME {sec)

vs Lag ¢ (sec) Plots for Four Exposure Conditions,
For the hot-system, short-exposure cases, the
slope of log Pi,0 VS t {the usual exhaust relation)
initially corresponded to a value of 540 liters/sec &
10% except when the initial pressure exceeded
1 x 1075 gignificantly, for which cases the mass
18 spectrometer peak was not exactly proportional
to pressure.

Four typical log py,o vs log ¢ plots are shown
in Fig. 14.6 for the cases of hot and cold region A4
and for 0.5-sec and 10-min exposures. The expo-
sute pressures, p_ ., were all within 20% except
for case Il, where p___  was about a factor of 5
lower. Exponential exhaust, indicated for water
by curve V, was observed for N, over almost four
orders of magnitude of pressure.

We observed that the log p,0 vs log ¢ plot of
our data closely approximated straight lines of
reproducible slopes, which, of course, varied with
exposure conditions.!! Integration of each curve
from the extrapolated limit of 0.1 sec to 1000 sec
and multiplying the result by our measured }{20
pumping speed (590 liters/sec * 3%) yielded the
quantity desorbed isothermally following the expo-
sure. The value of g (quantity absorbed) is be-
lieved to be within 20% of the measured quantity
desorbed.

The Elovich plots [Eq. (1)] from ocur data for two
temperatures and normalized to the same exposure
pressure (assuming variation of ¢ with the first
power of p _ ) are shown in Fig. 14.7. The ex-

exp e
posure pressures were from 0.7 to 4.0 x 1077 torr,
Determination of g for exposure pressures as low
as 4 x 1077 torr led to the relation of ¢ and Ploxp)

1.0510.1
9= Poxp f(t

exp) :

The only study found in the literature of the pres-
sure dependence of ¢ is that of Winfield,” who
reported an exponent of 1.2 with thoria at, of
course, higher water vapor pressures. The value
of 1.2 introduces an error band of somewhat less
than 15% to our data. The quantities observed
correspond to about 0.1 monolayer at room tem-
petature following a 10-min exposure and about
0.15 monolayer after 30 min. Dust or other im-
purities could be responsiblé for some or even all
of the observed sorption. We believe, however,
that the method is reliable and that sufficient care

 

llSee: John Howard and H. S. Tay!lor, J. Am. Chem.
Soc. 56, 2259 (1934) and other references in Low, op.
cit., pp. 301-7.

137

was exercised to be able to attribute the observed
chemisorption to the steel surface,

Some effect on water chemisorption kinetics by
ion or electron bombardment is to be anticipated
by analogy to other chemisorption phenomena.
Neither the magnitude nor the direction of change
seems to be presently predictable. Water sorption
by unbaked and baked stainless steel during a
pressure excursion to atmosphere of different hu-
midities has been observed and is reported in the
literature but with inadequate reported data to
permit scaling the parameters studied here over
the required five to six orders of magnitude.
Samples with different specific surface areas wmay
be expected to show different sorption capabilities.

The conclusion that chemisorption phenomena
exist leads to possible practical application in

 

 
  
   

 

HoO-STAINLESS 51

 

 

 

 

 

 

 

 

 

 

EXPOSURE TIME (min)

 

 

 

14x40% em® AT 28°C
- 2x10% e AT 180°C ———

  
  
   

 

 

 

 

 

Qo2

 

 

 

 

 

 

 

004

 

Q.04 0.06 Q.08
g Uore~liters of H;0 desorbed)

GA0

Fig. 14.7.

Stainless Steel.

Elovich Plots for Water Chemisorption on
138

Table 14.2. Proton NMR Spectra of DC-705 Diffusion Pump Qil and a Decomposition Product

 

 

 

 

 

Signal Height
Position of Integral (mm) Height x 0.1344 Protons per Signal Assignment
Structure |, 34 protons
0 (CHS)4Si, reference
19 22 2.96 3 —Me
32 46 6.18 6 2 equivalent Me
435 185 24.86 25 5 equivalent ¢
253 mm
0.1344 protons/mm
Unknown Solid
Signal Height Heipht
Position of Integral (tmm) Assumed Group Proton;;é;Group Assigament
0 Me4Si, reference
34 57 Me 19 } 2 equivalent methyl groups per
436 196 (f) 39.6 4 equivalent phenyl groups

 

operating nonbakeable vacuum systems. First,
since chemisorption occurs at even very low pres-
sures and is a strong function of exposure time,
maintenance of very low partial pressures of water
by liquid-nitrogen cooling of appreciable areas in
the system, followed by intermittent warming of
the cooled areas, desorption, and rapid exhaust,
should result in a lower water impurity level than
the same period of pumping would produce. Anocther
consequence of the observations reported here is
that even small water impurity levels introduced
from gas manifolds are expected to result in sig-
nificant impairment of the ‘‘base’’
water vapor.

pressure of

DECOMPOSITION OF DC.705 DIFFUSION
PUMP FLUID

A white solid was found condensed on the cold
caps and wall of one 10-in. diffusion pump on the
DCX-2 vacuum system.
sharply at 47°C.

It was found to melt

The infrared spectrum'? was

not distinguishable from that of the diffusion pump
fiuid, a pentaphenyl trimethyl trisiloxane, struc-
ture I.

P ¢ ¢
| | |

$—Si—0—Si—0—Si—¢

Me Me Me
¢ = CgHlg
Me = Chig
I

The manufacturer indicated that decomposition
to the dimer might occur if the operating tempera-
ture is too high, in a radiation field, or if the fluid
is heated in the presence of alkali impurities in

’G. Goldberg and H. L. Holsopple, Jr., Analytical
Chemistry Division.
the pump. Proton nuclear magnetic resonance (60
Mc) spectra were obtained'® for the solid and for
a sample of the pump fluid with the results sum-
marized in Table 14.2.

The dimer structure II may therefore be assigned
to the solid material. It would appear that for the
pump in question some period of excessive-temper-
ature operation probably occasioned the decompo-
sition. The observed lowered pumping speed of
this pump was presumably cansed by the deposit.

139

¢ ¢

¢~ Bi— O — Si—~ P

II

 

1 . . Co
SJ. R. Lund, Analytical Chemistry Division.
Part V
Nuclear Safety

 
15. Activities of Nuclear Safety Technical Staff

W. E. Browning, ]Jr.

The Nuclear Safety Technical Staff, comprised
of three persons, was formed early this year to
aid in planning, coordinating, and directing the
research and development activities within the
Nuclear Safely Program.

One of the first activities of the technical staff
was to survey work being done in the Nuclear
Safety Program on the development of analytical
models for fission product behavior in various
stages of reactor accidents and to recommend the
initiation of additional theoretical and experi-
mental activities which were needed, preferably in
existing groups outside the technical staff.

In a new activity, theoretical treatment of the
chemical behavior of gas-borme fission products
at high temperatures is combined with a labora-
tory investigation. ‘There are separate studies
of dynamic transport phenomena (continuing) and
of surface phenomena (new) of fission product
deposition at high temperatures. Models are
being developed for the transport and deposition
of fission products in containment vessels and for
the convective circulation of pases. Other theo-
retical work covers the behavior of aerosols. A
new program on behavior of fission products in
gas-liquid systems was started with studies on
the fission product removal by sprays at ORNL
and by suppression pools at General Electric,
San Jose, under subcontract. Another subcontract,
with Battelle Memorial Institute, supports work
on the theoretical chemical yield of methyl iodide.

A continuing program has been maintained for
computing thermochemical equilibria of fission
product fuel mixtures at high temperatures and
for predicting vapor pressures in the gas phase
above a condensed-phase solution. A multi-
component equilibrium program has been obtained
and debugged, and thermochemical data tapes
have been made for U, Sr, 5r0, UO, UOZ, and

M. H. Fontana

143

B. A. Soldano

U0, in addition to 140 compounds from the JANAF
Tables., Computer programs have been written to
generate thermochemical tables in the JANAF
format for compounds of interest, given some
spectroscopic or experimental data. An experi-
mental program to check theoretical calculations
by mass spectrometric measurements has been
initiated, and preliminary shakedown runs have
been made.

One member has assisted in the development of
a theoretical model describing fission product
behavior in containment vessels and in planning
an experimental program of scaling, flow visual-
ization, and pilot-plant experimeats to test the
model, At the invitation of the AEC, the staff
has participated in the management of an AEC
contract with Professor N. C. Ozisik at North
Carolina State University, who is working on the
theory of diffusional transport as it applies to
containment vessels.

One member has been assigned full-time to the
program on spray effectiveness (Sect. 18) during
its formative stages. A study undertaken by the
Chemical Technology Division at the instigation
of the technical staff has led to a report on the
theory of foam decontamination as it might be
applied to a teactor containment vessel.!

Recently, one man has spent half fime on the
special core melt-through problem assisting the
task force directed by W. K. Ergen. The work
involved estimating amounts of fission products
of molten core materials
Estimates were

released from masses
in large water-cooled reactors.
made on release by diffusion from the melt in
various configurations, and rough approximations

 

g, u. Jury, Foam Decentamination of Air Containing
Radicactive lodine and Particulates Following a Nu-
clear Incident, ORNL-TM=158%9 (Oct. 3, 19066).
were made for the case where natural convection
within the melt exists.

The Technical Staff participated in several in-
formation dissemination activities. Assistance is
given at the Nuclear Safeiy Information Center in
abstracting literature on accident analysis, heat
transfer, thermodynamics, and fluid mechanics,
and providing consultation services.
on filter and adsorber efficiency and pool de-
contamination in reactor safeguards was prepared

Information

144

for the United States representative at a meeting
held in November 1966 by the Committee on Re-
actor Safety Technology of the European Nuclear
Energy Agency. The technical staff assisted the
AEC in the production of a document describing
the Water Reactor Safety Program, its bases, the
interrelationships of the various projects,
delineation of the problems that are to he solved,
the relationship of the program standards and
codes, and other factors.

da
16. Correlations of Fission Product Behavior

THE LIGHT BULB MODEL FOR RELEASE
OF FISSION PRODUCTS

C. E. Milier, Jr.

Experiments in various laboratories and in re-
actors have accumulated much data on release of
fisgion products under conditiens which simulate
reactor accidents. The effects of important vari-
ables on fission product behavior have generally
been recognized, but interpretations of the data
have been at best empirical. The “light bulb”
meodel of fission product release preseanted below
seems fo explain in a simple fashion most, if not
all, of the available release data.

Fonda,’ in studies of the incandescent lamp,
observed that tungsten filaments lost weight (at
a given temperature) more rtapidly in a vacuum
than in the presence of a nonreactive gas. He
related this to Langmuir’s® theory that heat loss
from incandescent wires in gases is controlled by
conduction of heat through a stationary gas film
around the wire., Fonda proposed that evaporation
of material from the filament in a ncareactive gas
is controlled by diffusion through a similar statione-
ary gas layer, If it is assumed that a heated speci-
men of reactor fuel has such a nonreactive gaseous
boundary layer and that diffusion (by Fick’s law)
through this layer controls the rate of release of
fission products, then the theories of Langmuir
and of Fonda may be applied in a straightforward
manner.

If the rate of evaporation of species 4 is con-
trolled by Fick’s law diffusion of 4 through a
boundary layer of species B,

Jo d(:A ()
A AB"‘C'}';’

 

la R Fonda, Phys. Rev, 31, 260 (1928).
25, Langmuir, Phys. Rev. 34, 401 (1912),

145

where

2

J , = molar diffusion flux (moles em™* sec™ 1),

D = binary diffusion coefficient in which sub-

48
scripts A and B denote the two diffusing
species (em?/sec),

¢ = concentration {moles/cm ®),

z = distance (cm).
The concentration of fission products at the sur-
face of the melt is given by Henry’s law,

P=kX=k"P°X, (D

where

P = partial pressure of fission product (atm),
k = Henry’s law constant,
X = mole fraction of fission product in reactor
fuel,
k” = temperature~-independent portion of Henry’s
law constant,
PY - vapor pressure of pure fission product —
temperature~dependent portion of Henry’s
law constant (atm).

yvields Eg. (3), which
expresses fractional release of fission products as

The complete derivation®

 

fr=1—exp <—2.264
(/M , + 1/M )V 2ATY *kp Pt
- 10-—5 ;_{f A ‘/1 & - A\pA , (3)
p_B[(o"A + O‘B)/2] ud
where
fr = fraction released,
M, = molecalar  weight for fission product
(g/mole),
M = molecular weight of inert gas (g/mole),

 

¢, B Milter, Jr., The Light Buld Model of Fission
Product Release from Reactor Fuels, QORNIL-4060, in
preparation.
Tobie 16.1. Comparison of Calcuiated ond Mecsured Release of Fission Products from UJ, During Groin Growth®

 

 

 

Fraction Released

 

 

 

 

Temperature Time

- a » 4 1
Run {OC + SDOC) (hr) Calc. Meas. Calc. Meas, Calc. Meas. Calc. Meas. Calcg. Meas. Calc. Meas. Calc. Meas.

132 137 140 106 144

 

TeO, Te Cs Cs  StO 89g,  Bal Ba  Ru Ru  Ce0 ce VO, U
D/94 2000 2.1 90.37 0.18 0.1 o011 0.09°  0.09 0.09 0.05  0.04 0.01  0.02 0.02  0.03
D/64 2000 4.5  0.50 0.59  0.14 6.12  0.16 0,14 0.16 0.17  0.08 0,12 0.04 9.05  0.05 9.04
D/88 2050 1.5 0.41 9.66  0.12 0.21  ©0.13 0.19  9.12 9.20  0.67 0.22  0.03 0.08  0.04 {10
D/95 2150 1.5  0.46 0.35  0.12 0.06  0.21 0.15 .18 0.1 9,18 9.15  0.0a 0.66  0.08
D/65 2200 2.0 0.59 4.68  0.09 ¢.11  9.49 0.52 0.36 0.54  0.58 0.63  0.21 0.25  0.23 9. 26
D/66 2200 5.0 0.94® 0.94  0.25 0,33  0.85 o83 0.74°  0.74 078 0.78 0.47%  0.47  0.45°  0.45

91

 

a) e measured values in this table are caken from D. Davies, G. Long, and W. F. Stanaway, The Emission of volatile Fission Products from Uranium
Dioxide, AERE-R-4342, p, 11, (1953).

b, . , ‘
These values were oreset o caiculate Oor k /G
147

A - surface area of melt (cm?),
T = temperature of melt (°K),
t = time during which the specimen is molten
(sec),
p, = pressure of gas (atm),
1 = moles of melt,
o, =collision diameter of fission product
molecule (A),
Uy, = collision diameter of inert gas molecule
(A),
5 = boundary layer thickness (cm).

The fission products are, of course, the dilute
solute in a solute-solvent mixture. Equation (3)
describes the behavior of the solute. During long
heating periods the reactor fuel may vaporize
appreciably. If this phenomenon is to be accounted
for, the u term in Eq. (3) must be expressed as a
function of time. A similar release expression has
also been derived for the solvent, in which case

Raoult’s law,

P=rP°%, (4)
was used rather than Henry’s law. The corme-
sponding equation for the fraction of fuel vaporized
is given hy

u,. — U
fi‘ = —0—_._'

10 ‘_5 AT1/2 N ([5)
- ou, , !
M M pB[((J + o )/2] ’

where the subscript 1 refers to the fuel material.
The model has been tested with data from several
sources and satisfactorily explains the observed
behavior of fission products and fuel. One such
set was presented by Davies, Long, and Stanaway?
ont the emission of fission products on postirradia-
tion heating of UO,. The results of the compari-
son of the calculated and measured values for UO,
and several fission products are given in Table
16.1, A detailed description of this application
of the model is given elsewhere.® The model
describes the dependence of the fraction released
on the atmosphere, that is, its composition and
pressure; the solvent, that is, its surface area
and amount; and the chemical form of the fission
products, the temperature, and time at temperature.

= 2.264

 

 

4D. Davies, G. Long, and W. F. Stanaway, The Emis-
sion of Volatile Fission Products from Uranium Dioxide,
AFERE-R-4342 {(19563),

EFFECT OF CONTAINMENT SYSTEM SIZE
ON FISSION PRODUCT BEHAVIOR

G. M. Watson R. B. Perez
M. H. Fontana

Simple mathematical models to aid in the de-
termination of effect of containment size on rate
and extent of deposition of icdine from the gaseous
phase have been postulated. '
of systems (those with and without condensing
steam) have been considered.

In the absence of steam the principal assump-
tions made were: (1) homogeneous mixing within
the gas, (2) boundary gas layer enveloping: all
surfaces, (3) diffusion through this boundary layer
as the rate-limiting process, and (4) icdine present
in the molecular form with combined forms such as
methyl iodide absent. Mathematical relations in
terms of mass transfer coefficients, surface-to-
volume ratios of containers, and surface-character-
ization parameters were developed for cases with
and without desorption from partially covered sur-
faces.

In the presence of condensing steam,
bulk flow of steam toward the walls,
cases, it has been assumed that the iodine flux,
which consists of diffusive and bulk f{low com-
ponents, may be approximated by the bulk flow
component alone. Furthermore, it has been as-
sumed that the solubility of iodine in steam con-
densate is high enough to permit the indine and
the steam to condense together with the same

Two general classes

thers is
In such

composition as the gas phase. As a result of
these assumptions, relations have
been derived for the concentration of iodine as a
function of time in terms of the surface-to-volume

mathematical

ratios, condensing steam fluxes, and steam con-
centrations.

The case of molecular iedine-methyl iodide~
condensing steam has been considered. It was
assumed that the deposition of methyl iodide iz
limited by its smaller solubility in steam con-
densate,

Pevelopment of the model for predicting mass
transfer coefficients using film theory and boundary
layer analysis for cases in which no steam was
present was based on rough approximations cou~
ceming the flow velocity structure within the
containment shell; flows were assumed to be in-
The model shows
that observable mass transfer coeftficients can be

duced by aatural convection.
predicted if the temperature differences causing
the flow are very small (0.001°K). Both the
velocities and the temperature differences are too
small to be observed using currently available
instrumentation. The model also predicts size- and
pressure-scaling effects which appear to be sub-
stantiated by limited available data.

Some correlations of the mathematical expres-
sions with experimental data have been performed.
Results of iodine-air experiments in the CMF,
CRI, and NSPP have been compared with the simple
theoretical expressions with moderate success.
The results of the correlations of experiments
with diy atmospheres are shown elsewhere.® The
values of the mass transfer coefficient and of the
asymptotic concentration were obtained empirically
for the Nuclear Safety Pilot Plant (NSPP) from
results of experiments with iodine in dry air.

 

SG. M. Watseon, R. B. Perez, and M. H. Fontana,
Effects of Containment System Size on Fission Product
Rzhavior, ORNI.-4033 (in press).

CRNL-DWG 65-9731A

    
 
   
  

 

 

 

HK —_—T
(NITIAL SLOPE DETERMINES k,
A . :
0.5 3
Q.2
o Te—e
L o
S
...PREDICTED FROM .
NSPP DATA -
0.05 : : : '
- —— — —_
CMF EXPERIMENTAL DATA AND LINE
PREDICTED FROM NSPP DATA ALONE
0.02 ! ‘ -
cot | i et |
0 40 80 120 160 200 240 280
TIME {(hr)
Fig. 16.1. CMF Experimental lodine Concentration

and Values Predicted from NSPP Data,

148

Corresponding values for the Containment Mockup
Facility (CMF) and for the Containment Research
Installation (CRI) were obtained by scaling with
size and pressure the NSPP parameters in a man-
ner prescribed by relationships obtained from the
model.  For the asymptotic concentrations, the
surface-chemistiy effects were assumed equal in
all three systems.

An example of the correlation of data obtained
in the absence of steam in the NSPPP and in the
CMF is shown in Fig, 16.1. We can conclude from
the results® that the model predicts the general
behavior of the concentration-time curve in the
absence of steam. It appears quite successful
in the extrapolation of mass transfer coefficients
as indicated by the agreement between the pre
dicted and experimental initial slopes. The neces-
sity for additional research on the surface chem-
istry of containment materials is apparent from
the relatively poor agreement of the asymptotic
concertrations using the simple assumption of
equal surface behavior in all three facilities.

ORNL-DWG €5 -5430R

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

100 ‘ ——
' ¢ + L
g0 | / / / /
T < / | 7
/‘% | /
o 70| i
wl
I._.
Q
L’j /
3 60
g ° / / ]
O
.
& /
5 50 /
- /
Q + a _
= a0 / / /  m CONDENSATE ACCUMULATION |
Q : o [ODINE ACCUMULATION
& / O NEPTUNIUM  ACCUMULATION
& . ] o BARIUM ACCUMULATION |
/1 // + IODINE CALCULATED
’ y;// |
10
0
0 2 4 6 8 10 12 14
TIME {(hr)
Fig. 16.2. Accumulation of Condensate and of Fis-

sion Products in Condensate vs Time, NSPF Run 8.
The simple assumptions of the condensing steam
model have been tested using the experimental
data on an NSPP experiment which provides infor-
mation on both iodine and water condensate col-
lection as well as system pressure and tempera-
ture as a function of time, Figure 16.2 shows a
comparison of the calculated and experimental
values of the iodine collected., The agreement
appears to be quite satisfactory.

Based on a very limited number of tests, it
appears that the behavior of jodine in containment
systems differing in size by two orders of magni-
tude may be correlated with moderate success
utilizing simple mathematical relationships.

CHEMICAL EQUILIBRIUM STUDIES OF
ORGANIC-IODIDE FORMATION UNDER
NUCLEAR REACTOR ACCIDENT CONDITIONS

R. H. Bames® J. F. Kircher®
C. W. Townley®

The presence of CH,I in nuclear reactor environ-
ments poses a potential hazard because of the dif-
ficulties involved in removing this compound from
gases using conventional trapping techniques.
To gain insight into the chemical processes lead-
ing to CH,I formation, a study was performed of
calculated equilibrium concentrations of CH,I
and other important species for a range of condi-
tions typical of reactor-accident systems. A
report of this work has been issued,’

The results of this study indicated that chemi-
cal systems containing iodine and simple com-
pounds such as COZ, HQO, C2H4, and CH, would
be expected to generate CH, L Such materials are
found as trace pollutants in the atmosphere, even
if some of them are thermodynamically unstable®-®

 

6Bat’telle Memorial Institute; work performed under
subcontract. This summary, prepared by W. E. Brown-
ing, Jr., is based on reports by the listed authors.

"R. H. Barnes, J. F. Kircher, and C. W. Townley,
Chemical-Equilibrium Studies of Organic-lodide Forma=
tion Under Nuclear Reactor Accident Conditions, BMI-
1781 (Aug. 15, 1966).

,c. R Junge, Air Chemistry and Radiocactivity, p.
355, Academic, New York, 1963.

9A. P. Altshuller and T. A. Bellar, J. Air Pollution
Control Assoc, 13, 81 (1963).

149

ORNL~-DWG 67~ 790

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

._? T
Hz
CeHg
__.9 j
HI
T [ — CHxl
o .
@ Iz
<
£ /
t
Z -5
o
2
= —-17
=
w
O
=
8 -9
w0
&
3
- 21 e
_23 ______________________________ e e .
—-25
300 o200 700 900 oo 1300 1500
TEMPERATURE (°K)
Fig. 16.3. Species Concentrations as a Function of
Temperature for the Chemical Equilibrium Systems Con-

taining C2H4, H,, HI, |2, and CH;l. Baosed on total
concentrations equivalent ta 1 x 10-8 g-mole per liter

of H2, 1x 10-° g-mole per liter of C2H and 1% 10~19

4°
g-mole per liter of |2.

and may exist only in transient conditions, Ac-
equilibrium calculations were made
the presence of several possibly
The calculated equilibrium

cordingly,
postulating
significant pollutants,
composition in the presence of ethylene is shown
in Fig. 16.3, which shows that at 300 to 500°K a
major fraction of the iodine appears as CH_L

These chemical equilibrium calculations indicate
that there are realistic conditions under which
CH,I could be generated if sufficient reaction
time were available. Chemical kinetic calculations
are now being made for promising reactions and
conditions identified in the equilibrium studies
in otder to assess rates of formation of methyl
iodide.
THE ADEQUACY OF SCALEUP IN
EXPERIMENTS ON FISSION PRODUCT
BEHAVYIOR IN REACTOR ACCIDENTS

C. E. Milles, Jr. W. E. Browning, Jr.

We concluded in a previous report!® that scaling
of experiments on fission product release and
transport in the U.S. Nuclear Safety Program does
not seem adequate; a very large extrapolation
will be necessary to compare presently available
data to that obtained from the projected Loss-of-
Fluid Test (LOFT).!! In a subsequent report'?
we propose experiments at 1 and 10% of the size
of LOFT (based on the mass of fuel) in order to
fill the gap in scaling.

Intermediate-scale experiments to study fission
product release and transport should be designed
so that they (1) generate fission products with
maximum practical realism as a function of various
simulated loss-of-coolant accident environments
and heatup rates and (2) simulate core geometry
sufficiently to achieve proper temperature profiles
and heatup rates, to allow natural movement of
core components during the heatup cycle, and to
provide a realistic escape path for fission products
through remaining core materials.

The characteristics
capabilities should be incorporated into the experi-

following eneral and

ments:

1. The experimenis should be performed in-pile
on multipin fuel subassemblies to give maximum
realism to the time-temperature-position pro-
files.

2. Where possible the experiments should be de-
signed to accommodate water, liquid metal,
and gas-cooled fuel systems.

3. The released fission products should be trans-
ported through a representative primary system
in which the transport rates result from realis-
tic thermal gradients and natural steam flow.

 

10¢. %, Miller, Jr., and W. E. Browaing, Jjr., The
Adequacy of Scale-Up in FExperiments on Fission
Product Behavior in Reactor Accidents, Part I. An
Analysis of Scaleup in the U.S. Nuclear Safety Program,
ORNL-3901 (July 1966).

U R Wilson et al., An Engineering Test FProgram
to Investigate a Loss-of-Coolant Accident, IDO-17049
(October 1964).

12 & Miller, Jr., and W. E. Browning, Jr., The
Adeguacy of Scale-Up in Experiments on Fission Prod-
uct Behavior in Reactor Accidents, Part II., Recom-
mended Additional Nuclear Safety Scale-Up Experis
ments, ORNIL-4021 (December 1966).

150

4, Parameters of interest in the experiments are
(1) maximum operating temperature (the experi-
ment should be capable of operating at initial
conditions similar to those of real reactor
operation), (2) core environment (steam, steam-
hydrogen, steam-air), (3) cladding and core
materials, and (4) burnup.

5. Measurements of interest are: (1) temperatures
of fuel and cladding, (2) observation of geo-
metrical changes, (3) permanent and temporary
fission product platecut (on fuel materials,
cladding, and ozxidized cladding in the un-
melted part of the fuel subassembly, and on
the walls of the primary, etc.) as a function
of time and temperature in the primary system,
(4) extent of fuel oxidation, fuel melting, and
fuel melting point lowering via eutectic forma-
tion, (5) extent and rate of metal-water reac-
tion, (6) size, (7)
transport as a function of flow (diffusion or

particle fission product
forced), and (8) physical and chemical form of
radioactive aerosols which remain in the gas
phase as a function of time.

ORNL-DWG G6—-T704R3

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

   

 

 

 

  
  
 

 

 

 

 

10
: Cs
10 Ere 7_;Q:r,{EROPOSED)— ,,,,1;_ —_— —n
. T OFTORCSE —— == o T T ——— ]
2 [~ {(PROPOSED) — .. LT I _ ]
— e - - o
[
C_J) 100 priee— T s e
o F—PBFORCRI.—T——
[0 (PROPOSED) ... ...
L L - e . _
= i U CRIL e _
g . {PROPOSED)
flj 10 T ——-QRNL —
o - .
wr ORNL —®
™~ """ MULTIRPIN — ~ ~7~ o
o - e . - e
L(; — ® NSPP
o \__.; CR,[;T’:_ S —’:_*** T
2 | e T D
I A — ORNL MULTIPIN 7 [
© p——— ORNL — - - 70?'?7 - —
“ TREAT — = ORNL
. ®
1073 —---- —-——— ORNL TREAT,
— v - ORR,
CRR 2 - —
100°% L
RELEASE TRANSFORT BEHAVIOR IN BEHAVIOR IN
FROM FUEL IN PRIMARY CONTAINMENT GAS CLEANING
SYSTEM SYSTEM SYSTEM
i ’
¥
AREAS OF INVESTIGATION
Fig. 16,4. Comparison of Size of Planned Nucleear

Safety Experiments with Regard to Various Stages of
Fission Piodust Behavior (Including the Proposed Ex.

periments).
The proposed experiments on the release and
transport of fission products from UO, are de-
scribed in some detail elsewhere,'? They are
scaled at 1 and 10% of the size of the LOFT ex-
periment based upon the mass of fuel in the LOFT
The fuel in these experiments ig to be
melted by nuclear self-heating in a reactor {facility
using a driver core. The suggested facilities are
the Power Burst Facility'® for the 1% scale ex-
periment using approximately one LOFT {fuel ele-
ment, and the LOFT itself for the 10% scale ex-
using approximately five LOFT fuel

core,

periment
elements,

Alternative experiments at the same scale using
a nonnuclear method of heating are proposed to
fill the gap in scaling if for reasons of expense or

scheduling the nuclear experiments cannot be
performed. The 1% scale experiment is proposed
for the Containment Research Installation,* and

the 10% scale experiment is proposed for the
Containment Systems Experiment.'® An electrical
method of heating the fuel is proposed, which is
presently under development in an AEC-sponsored
program. '

The capability of performing any of the proposed
experiments appears to be near at hand. All four
facilities are under construction, and the capability

151

of electrical heating is being developed. How-
ever, three of the four possible experiments involve
facilities which are already scheduled for experi-
mental programs. These are the LOFT, CSE, and
CRI facilities. The addition of these proposed
experiments would cause delays in programs which
are very much needed in other phases of nuclear
safety work. The remaining facility, the Power
Burst Facility, which is still to be constructed,
seems the most promising.

With the addition of these experiments to the
AEC Nuclear Safety Program, the comparison of
size of planned nuclear safety experiments will
be as shown in Fig. 16.4. The accomplishment
of these experiments will produce a scaling range
which should cover all possible mechanisms of
fission product behavior and make extrapolation
of data from one experiment to the other more
reliable,

 

13]:1, and R. 5. Kern (eds.), Preliminary
Power Burst Facility, 1DO-

Feinauer
Safety Analysis Report,
17060 (revised June 1965).

146G, W. Parker and W. J. Martin, *‘“The Containment
Research Installation,’’ Nucl. Safety Program Semiann.
Progr. Rept. June 30, 1965, ORNL-3843, pp. 92..07,

135, J. Rogers, Program for Containment Systems
Experiment, HW-83607 (September 1064).
17. Nuclear Safety Tests in Major Facilities

FISSION PRODUCTS FROM FUELS UNDER
REACTOR-TRANSIENT CONDITIONS

G. W. Paiker R. A. Lorenz
J. G. Wilhelm

ORNL fission product telease experiments are per-
formed in the TREAT reactor to study the release
and transport of fission products from fuel during re-
actor transients in which fuel is heated and melts
rapidly. In the cument series of experiments, vari-
ous components of the equipment are designed to
simulate the core, pressure vessel, containment ves-
sel, and the filter and charcoal cleanup system of a
typical large pressurized- or boiling-water power
reactor. Experiments 9, 107, 11Z, and 12Z were
recently completed and repoited in detail else-
where, 2 making a total of six experiments in which
stainless-steel- or Zircaley-2-clad UO, fuel spec-
imens were melted underwater by transient heating.
The letter Z indicates experiments in which
Zircaloy-2 cladding was used. FEach of these ex-
periments used 32 g of 10.7% enriched UO, sin-
tered into pellets 0.400 in. in diameter, with 0.020-
in.-thick cladding. Heat input to the fuel by in-
ternal fissioning during the transient was approx-
imately 520 calories per gram of UOZ; this treat-
ment heated the UO, well above its melting point.

In each experiment a reactor accident was sim-
ulated by first preheating the fuel autoclave to
about 120°C, executing the transieat which melted
th and allowing the transient-
generated steam aerosol to leave the fuel auto-
Steamm was condensed and collected in
water traps, and noncondensable gases passed
through a series of filter papers and charcoal-

<

fuel specimen,

clave.

10n assignment from Karlstuhe Center for Nuclear
Research and Development, Karlsruhe, West Germany.

2G. W. Parker, 2. A, Lorenz, and J. G. Wilhelm, Nucl.
Safsty Program Semiann. Progr. Rept. Dec. 31, 1966
{in preparation).

152

loaded papers into a gas collection tank. In sim-
ulation of accident after-heat, the fuel autoclave
was then heated electrically to abeut 300°C for
1 hr to boil out any remaining water.

We wished to determine the maximum fission
product release; so in addition to using the two
different cladding materials, the rate of steam
release from the melting region was varied.
experiments 7 and 87, ® only about 5% of the water
surrounding the fuel specimen boiled out of the
fuel autoclaves in the first minute following the

In

transient. In experiments 9 and 10Z, approximately
75% of the water boiled out in the first minute,
and essentially all of the water boiled out of the
11Z and 127 fuel autoclaves in 1 min. These two
latest experiments also explored the effect of
pressure during melting by enclosing the fuel and
water in sealed primary vessels. Experiment 117
used a 300-psi rupture disk to release the steam,
and experiment 12Z used a 2500-psi mupture disk
along with 500 psi of helium as preliminary pres-
surization. Examination of these two experiments
is in progress.

The fuel and cladding melted completely in
experiments 7, 8Z, 9, and 10Z, except for portions
of the metal end caps. The melted residue from
experiments 7 and 9, which used stainless steel
cladding, appeared to be foamy and more porous
than the residue from experiments 8Z and 10Z.
In Fig. 17.1 the firont half of the crucible in ex-
periment 10Z has been removed to reveal some
of the nonporous solidified fuel and cladding
around the sample holder pedestal and some ma-
terial splattered onto the crucible and flux monitor
capsule.

Approximately 24 and 16% of the stainless steel
cladding reacted with steam to form hydrogen in

 

3. W. Parker, R. A. Lorenz, and J. G. Wilhelm, Nucl.
Safety Program Semiann. Progr. Rept. Junc 30, 1965,
ORNL-3843, pp. 39—67.
experiments 7 and 9. Forty-one and forty-nine
percent of the Zircaloy-2 cladding reacted in ex-
periments 8Z and 10Z. These results all agree
well with those found by workers at Argonne
National Laboratory in metal-water reaction studies
in TREAT.?

Fission product release and transport was sim-
ilar in all four experiments except that transport
of the wvolatile elements tellurium, cesium, and
iodine out of the fuel autoclave in experiments
G and 10Z (6 to 19%) was about twice that in
experiments 7 and 87 (2 to 7%). The larger release

5

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 17.1. Opened Crucible from TREAT Experiment
10Z, Showing Upper End Cap, Flux Moniter Capsule,
and Part of Nonporous Fuel-Cladding Residue Around
Sample Holder and on Crucible Wall.

 

153

in 9 and 10Z is attributed to the faster steam
release. About 1% of the barinm and strontium,
0.3% of the ruthenium, and less than 0.1% of the
cerium, zirconium, and UO, were carried out of
the fuel autoclave with the released steam in each
of the four experiments. The condensation process
was highly efficient in trapping fission products
and UO2 in these experiments; a decontamination
factor of about 10 for nonvolatile materials was
observed. No significant effect of cladding ma-
terial was evident.

In experiments 9 and 10Z, distilled-water rinses
of the fuel autoclave walls contained 33 to 44%
of the total cesium and iodine; this behavior sug-
gested rapid formation of nonvolatile water-soluble
compounds, possibly cesium hydroxide and various
metal iodides.

The amount of unreactive or penetrating iodine
was only a small fraction of the total in all four
experiments. Sixty to eighty percent of the total
1311 was released from the melied fuel specimens,
but only 0.0006 to 0.005% was found on the char-
coal-loaded papers and in the gas collection tank.
This iodine was characterized as unteactive ot
penetrating, based on its poor sorbability in the
bed of 27 charcoal-loaded papers.

SIMULATED LOSS-OF-COOLANT EXPERIMENTS
IN THE OAK RIDGE RESEARCH REACTOR

B. F. Roberts
R. J. Davis

C. E. Miller, Jr.
R. P. Shields

The simulated loss-of-coolant experiments, con-
ducted in previously described facilities® at the
ORR, are designed to provide information on re-
lease and transport of fission products in reactor
accidents. It is intended that the information ob-
tained can be used to predict fission product be-
havior under conditions beyond those tested so
that hypothetical accidents can be more realis-
tically evaluated.

The interpretation of data from previous experi-
ments on fission product release and behavior has
been the main activity during a period when major

 

*R. C. Liimataimen and F. J. Testa, Chem. Eng. Div.
Semiann. Progr. Rept. July—Dec. 1965, ANL-7125, pp.-
170--78.

5W. E. Browning, Jr., ef al., Nucl. Safely Program
Semiann. Progr. Repi. June 30, 1965, ORNL-3843,
p. 156.
constiuction work has been under way on the re-
actor facility.

A major construction program has been carried
on during the past year involving the reactor fa-
cility in which the in-pile fission product release
experiments are performed. The modifications are
in line with the emphasis on fission product trans-
port. They include the construction and instal-
lation of a simulated reactor containment vessel,
the installation of on-line analyzers for the de-
termination of water, hydrogen, oxygen, and carbon
oxides in the sweep gas, and of increased capa-
bility for temperatiure measurement and control in
the experiment system. A schematic diagram of
the new system is shown in Fig. 17.2. The avail-
ability of this modified facility will allow the
complete study of realistic fission product release,
transport, and behavior in a vessel.

The considerable literature has been examined
in a study of the extent and the rate of sorption
of iodine on a variety of surfaces and possible

components of reactor systems. The practical re-

154

sults of this study, from nuclear safety consider-

ations, are as follows: -
1.

tably clean steel, clean stainless steel, and soiled

steel with oxide coatings) can be interpreted very

well by mechanisms involving monolayer adsorp-

Sorption of iodine on several materials (no-

tion with dissociation.

2. Absoiption of iodine in water or in aqueous
solutions can be treated by use of models involv-
ing diffusion through boundary layers. Such mech-
anisms for iodine deposition will be important
when surfaces are covered with water films during
steam condensation within a containment vessel.

3. Sorption of iodine on stainless steel and
some other metals covered with an oxide layer
(e.g., as-treceived stainless steel) seems to in-
volve complicated mechanisms and to follow rel- -
atively complex kinetics. Since stainless steel
is likely to be an abundant material in containment
systems, it will be necessary to study the sorp-
tion-desorption process as a function of oxide
thickness and temperature,

CRNL~LR-DWG 56274 R

 

 

 

  

 

 

 
 

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 HV-902
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PRESSURE REGULATOR VALVE
PRESSURE INDICATOR
PRESSURE SWITCH
DIFFERENTIAL PRESSURE GAGE
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FURNACE

 

Fig. 17.2. ORNL In-Pile Fuel Destruction Experiment: Fiow Diagram.
IGNITION OF CHARCOAL ADSORBERS

C. E. Miller, Jr. R. P. Shields

Charcoal adsorbers are used in many present and
proposed reactor safety systems to remove iodine
from the containment system atmosphere. The High
Flux Isotope Reactor uses such a safety system.
Such charcoal adsorbers will ignite and burn if
temperatures and air flows are sufficiently high.
Reactor accidents would discharge large quan-
tities of fission products whose radioactive decay
could produce serious overheating (especially in
local hot spots) of the adsorber beds. Such ac-
cidents might also permit air to enter the contain-
ment system and the adsorber bed. Accordingly,
we have initiated a program of tests both in the
laboratory and in-pile to establish the effect of
fission products and of irradiation on the ignition
temperature of charcoal adsorbers.

L.aboratory studies have been conducted primar-
ily with Barnebey-Cheney type KE (BC-KE) and
Mine Safety Appliances (MSA) No. 85851 char-
coals, both of which have been widely used in
adsorbers. The data from small-scale investiga-
tions of these materials show that ignition tem-
perafures vary somewhat for various lots of the
same charcoal; initial ignition temperatures dif-
fered by 30°C for two lots of MSA charcoal, while
subsequent ignition temperatures differed by 15
to 20°C. Ignition temperatures seem independent
of apparatus material (glass or metals such as
stainless steel) and of tubing size over the small
range (0.5 and 0.7 in. in diameter) studied. Initial
ignition temperature was unaffected by change in
air flow rate from 20 to 40 fpm, but subsequent
ignition occurred at temperatures 10 to 15°C lower
at the higher flow rates. Ignition temperatures
were lowered measurably (about 6°C) when moist
air was substituted for dry air at the same veloc-
ity.  The ignition temperatures for the BC-KE
charcoals were increased by addition of iodine.

A most important finding in this study is that
charcoal from adsorbers which had been in service
for one year on the NS “Savannah’’ showed igni-
tion temperatures 150 to 200°C lower than unex-
posed samples of the same charcoal. The reason
for this difference is not yet known.

155

In-pile tests have been conducted in an experi-
mental unit (see Fig. 17.3) designed for use in
the fuel melting facility in the ORR. This facility
and the first in-pile ignition test of this series
were described in a previous report.® A total of
100 ignitions have been performed with ten ignition
cycles per position. Countrol of the position of the
U0, cylinder (from which fission products were
emitted to the charcoal adsorber) permitted study
at three levels of fission product concentration in
the charcoal; iodine concentrations cormesponded
to 4.3, 5.0, and 7.0 w per square inch of charcoal
surface. These energy release rates are greater
by a factor of 8 than those expected in HFIR,
LOFT, or NPR adsotbets.

Ignition temperatures for the in-pile experiments
differed in several regards from those observed
in the laboratory experiments. Typical experiments
with & BC-KE charcoal showed that initial igni-
tion, which took place after the iodines were at
equilibrium, occurred at 336°C; this value is not
significantly different from the value 341°C which
is the average of ten fast ignitions in the same
apparatus out-of-pile. However, subseguent igni-
tions in the in-pile assembly occurred at tempera-
tures up to 40°C higher. This increase in ignition
temperature (which is quite unlike the laboratory
behavior) persisted even after the unit was re-
tracted to stop fission product (and jodine) gen-
eration. When an in-pile ignition experiment was
resumed (in the retracted position) after a 45-day
lapse following an equipment malfunction, the
ignition temperature remained at some 20°C above
its laboratory value, even though virtually all
iodine activity had decayed. Some factor in addi-
tion to iodine and the short-lived fission products
seems respensible for a part of this temperature
increase.

Future in-pile experiments will be performed to
study the ignition temperature behavior of high-
ignition charcoals and impregnated charcoals, In
these experiments a controller will be used to
reproducibly control the heating rate of the ad-
sorber. Future out-of-pile experiments will include
the further development of a standard ignition ap-
paratus and the study of ignition characteristics of
various charcoals.
156

ORNL-DWG €6-8047R

GAS EXIT

SOLENOID VALvVE - BYPASS TO GAS EXIT

CHARCQAL TRAP

GAS INLET TQ IGNITION
TUBE

 

CHARCOAL TRAP ...
(GAS EXIT)

- "GLASS-WCOL
BUFFER

 

T AR i

SOLENOID VALVE- ..
- ELECTRIC HEATER

CHARCOAL .

 

 

 

 

THERMOCOUPLE (2)

*As—in. STAINLESS -
STEEL-SHEATHED

CHRCMEL-ALUMEL

THERMOCOJPLES .

 

 

 

 

 

 

 

 

CHARCOAL IGNITION
UNIT

 

 

 

STAINLESS STEEL
SCREEN FILTER

- HEATER

ALUMINUM COOLER

FISSION CHAMBER
(GAS EXIT) — - - —..

Y
o
i
i

#

- Tho, PLUG
" TWO UQ, CYLINDERS

FISSION CHAMBER
(REACTCR FURNACE)--- g

T 7 =ThO, HOLDER

 

Fig. 17.3. Experimental Eguipment for In-Pile Charcoal Ignition Experiment IGR-2,
FISSION PRCDUCTS FROM ZIRCALOY-CLAD
HIGH-BURNUP UOQ,

G. E. Creek
R. A. Lorenz

W. J. Martin
G. W. Parker

The previous report® gave results of an experi-
ment performed in the Containment Mockup Facility
(CMF) with stainless-steel-clad UO, irradiated to
a burnup of 1000 Mwd/ton. Data obtained in a
similar experiment (run 4-11) with Zircaloy-clad
UC, irradiated to a burnup of 7000 Mwd/ton are
reported below.

The conditions used in this experiment were
quite close to those prevailing in the previous
high-burnup experiment.® The total pressure in
the CMF, furnished by a mixture of air and steam,
was about 29 psig before heating the fuel. Heating
of the fuel was started with a mixtare of steam
and air flowing through the pressurized furnace
tube, but steam flow was discontinued before the
fuel temperature reached an estimated 2200°C be-
cause of steam condensation in the furnace tube.
Dry air flow continued during the balance of the
heating period (10 min total) and for 20 min there-
after. During most of the 20-min cooling period,
bumning of the specimen was obsetved at irregular
intervals; such behavior has been observed with
previous Zircaloy-clad UO, specimens that had
been melted and then cooled in air.

The distribution of iodine observed in this ex-
periment is compared in Table 17.1 with the data
obtained with the 1000 Mwd/ton burnup fuel® (run
10-29). The distribution of four other important
fission products is similarly compared in Table
17.2. 1In both tables the observed differences are
not large enough to be considered significant
without further, confirmatory, data. The high trans-
port of iodine observed in the 7000 Mwd/ton experi-
ment (21.6% as compared with 9% for the 1000
Mwd/ton experiment and 6 to 10% in simulant ex-
periments) was an unexpectedly large difference.
The difference in iodine collection on plateout
samples is due, at least in part, to a larger number
of painted carbon-steel samples in run 4-11 as
compared with run 10-29. There appears to be no
obvious explanation for the smaller fraction of

 

iodine in the condensate in the 7000 Mwd/ton
experiment.
5

(x, W. Parker et al., Reactor Chem. Div. Ann. Progr.
Rept. Dec. 31, 1965, ORNL-3913, p. 138.

157

Table 17.1.

Distribution (%) of lodine Released

from High-Burnup U02

 

 

Burnup
1000 7000
Mwd /ton Mwd /ton
Iodine released ~10G0 90.8
Iodine held in containment tank
Retained on tank walls 16.8 20.3
Collected on plateout samples 7.2 23.7
Collected in condensate 57.6 19.2
Total icdine retained 84.6 63.2
Iodine removed from tank after
aging by
Pressure release 2.9 7.7
Argon displacement 4.8 11.5
Air sweep 1.2 2.6
Total icdine transported 8.9 21.8
from tank
Retention of airborne iodine
from tank
On filters 0.35 2.1
On silver or copper screens 5.6 5.0
In charcoal cartridges 2.2 11.7
Todine in penestrating form 0.6 0.6

 

a . . . .
Based on the ratio of activity on the silver section of
the diffusion tube to that on the charcoal-lined section,

Data on airborne fission products in the contain-
ment tank as a function of time were also obtained
from analysis of gas samples taken at various
intervals. Comparison of the data with those for
the 1000 Mwd/ton experiment® shows little dif-
ference in airborne I, Cs, Mo, and Ru. The air-
borne tellurium values are lower for the 7000
Mwd /ton experiment, reflecting the alloying effect
of the zirconium cladding, while the barium and
strontium values are higher as the result of the
reducing effect of the zirconium. Both phenomena
have beea observed earlier.’

Fission products released from Zircaloy-clad
7000 Mwd /ton burnup fue!l in the CMF differed from
those released from stainless-steel-clad 1000

 

’G. W. Parker et al., Nucl. Safety Program Semiann.
Progr. Rept. Dec. 31, 1962, ORNL.-3401, p. 5.
158

Table 17.2. Distribution of Fission Products Refeased from High-Buraup |.102 (Runs 10-29 and 4-11)

 

 

Fission Product Burnup Level Furnace Tube Aerosol Total
Element (Mwd /tomn) and Duct Tank Condensate Filters Release
to Tank Walls from Fuel
Cesium 1000 3.0 15.1 43.6 .8 62.5
7000 8.6 13.8 6.2 1.5 30.0
Tellurium 1000 0.3 6.9 0.8 0.45 8.4
7000 0.7 1.4 0.04 0.112 .3
Ruthenium 1000 0.07 0.35 0.12 0.0006 0.54
7000 0.16 0.05 0.002 0.004 .22
Strontium 1000 0.01 0.04 (0.0003 0.0004 0,05
7000 0.05 0.11 0.015 0.18

Mwd/ton fuel in a way that could have been pre-
dicted from the difference in cladding materials.
No variations were observed that could be at-
tributed to the difference in burnup level of the
fuel specimens.

BEHAVIOR OF fiz AND Hi IN THE CONTAINMENT
RESEARCH INSTALLATION TANK
G. W. Parker G.
W. J. Martin

The Containment Research Installation (CRI)
has been constiucted at Oak Ridge National Lab-
oratory for the investigation of fission product
release, transport, and plateout as a function of
burnup, fuel and containment temperature, time
at temperature, and atmospheric and containment
sutface composition.

The primary objective of the initial experiments
in the CRI was to compare the deposition behavior
of molecular iodine (run 100) and HI (run 101) in
a stainless steel system under ambient conditions
of temperature and humidity and an internal air
pressure of 30 psig.

In run 100 an initial concentration of 2 mg/i'n3

1301 was intro-

of elemental iodine tagged with
duced into the CRI containment vessel by com-
pressing a thin-wall stainless steel tube which
contained an iodine-filled glass ampule. An air
stream passed through the heated tube for 12 min

to transport iodine to the containiment vessel.

 

SPhillipS Petrolenm Co., on assighment to Oak Ridge
National Laboratory.

0.0026

Over a period of 22 hr, approximately 98% of the
airborne iodine was deposited on the tank walls,
Depletion of the iodine occurred very rapidly dur-
ing the first 12 min and then more slowly for the
remainder of the experiment. The results are sum-
in Table 17.3.
retained by the absolute filter media increased
with time, while the fractions which passed such
filters did not. It appears that the filterable frac-
tion of the iodine has a rate of deposition on the

marized The f{raction of iodine

tank which differs from that of the other forms of
iodine present.

In run 101,
passing a hydrogen gas stream over a crushed

liydiogen iodide was formed by
capsule containing I, tagged with !°°T and com-
bining these components on a platinum catalyst
at 400°C. The product was formed and injected
into the CRI tank during a period of about 35 min.

The total weight of iodine used was again suf-
ficient to give an initial concentration of about
In this run, as well as in run 100, ap-
proximately 98% of the activity was deposited on
the containmeit vessel surfaces during the course
of the experiment. Table 17.4 suminarizes the
sampling data, which show that the rate of deposi-
tion of HI over the first 3 hr was faster than that
of I, in run 100. The slower deposition rate after
3 hr possibly indicates that the HI has changed
the adsorption characteristics of the stainless
steel surfaces, rendering them wmore inactive, or
that it reacted to give an unidentified form of
iodine which was only slightly reactive with the

2 mg/m>.

stainless steel containment vessel,
159

Table 17.3. Distribution of lodine Among CRI Gas Sompler Components in Molecular |2 Experiment (Run 100)

 

 

 

Time After Percent of
Completing Total Iodine ¥raction of Sample Iodine Activity Coellected by
I2 Injection Inventory Absolute Silver Charcoal
inte Tank Adrborne in Filters® Membranes” Cartridges©
{min) Tank
12 32.4 0.093 0.898 0.009
38 26.3 0.083 0.911 0.006
76 22.7 0.074 0.922 0.004
161 10.5 0.074 0.922 0.004
285 6.9 0.089 0.904 0.007
400 3.2 0.128 '0.860 0.012
710 1.0 0.417 0.539 0.044
832 0.77 0.506 G.434 0.060
1020 0.47 0.594 0.328 0.078
1155 0.42 0.59 0.328 0.080
1255 0.41 0.588 0.329 0.083

 

Note: See Table 17.4 for explanation of feotnotes.

Table 17.4, Distribution of lodine Among CRI Gas Sampler Components in Hl Experiment (CRI Run 101)

 

 

 

Time After Percent of
Completing Total Iodine Fraction of Sample Iodine Collected by
RI Injection Inventory Absolute Silver Charcoal
into Tank Alrborne in Filters® Membranes” Cartridges”
{min) Tank
16 12.0 0.16 0.83 0.01
40 8.5 0.07 0.84 0.09
76 6.1 0.05 0.48° 0.47
151 4.2 0.04 0.91 0.05
477 2.8 0.04 0.94 0.902
604 2.5 0.03 0.96 0.01
728 2.3 0.04 0.95 0.01
842 2.0 0.04 0.90 0.06
918 1.9 0.05 0.93 0.02
996 1.7 .04 0.94 0.02
1003 0.9 0.05 0.93 0.02

 

“Absolute filter media — Flanders filter 7H70A.
bSilver membrane filter — Flowtronics silver membrane, 5 L pore size, four or eight filters were used.
“Charcoal cartridge — 1/2 in. of unimpregnated charcoal and }Kz in. of impregnated charcoal.

dFaulty mounting of silver membranes apparently caused low HI retention on silver membranes and high charcoal
collection value.
18. Laboratory-Scale Supporting Studies

DEVELOPMENT OF FILTRATION
AND ADSORPTION TECHNOLOGY

K. E. Adams
Jack Truitt

J. S. Gill
W. D. Yuille!

This program is intended to advance the tech-
nology of removal of fission product gases and
colloidal dispersions by adsorption and filtration
techniques in order to increase the confidence
in the reliability of the effectiveness of various
air cleaning systems unader accident conditions.
As this confidence is achieved, a wider acceptance
of air cleaning systems as engineered safeguards
will be accomplished.

Results of these studies are reporied in detail
elsewhere. ?
niques have been reported previously.®'* Briefly,
a reproducible aerosol containing stainless steel
and oxides of uranium is used for these tests.
The test aerosol is prepared by striking an elec-
tric atc between electrodes consisting of UO,
(with thoriated tungsten) and a stainless steel
tube packed with UO,. Under diy conditions the
test aerosol (median size of primary paiticles
0.018 p) is easily filtered out of air streams by
any of the so-called ‘‘high-efficiency (absolute)”’

Prior results and experimental tech-

filters with efficiencies of removal in excess of
09 99,

1Visiting scientist from Great Britain.

’R. E. Adams et al., Nucl. Safety Program Ann. Progr.
Rept. Dec. 31, 1966 (to be issued).

w. E. Browning, Jr., et al., ““Removal of Particulate
Materials from Gases Under Reactor Accident Con-
ditions,?® Nucl. Safeity FProgram Semiann. Frogr. Rept.
June 30, 1965, ORNI.-3843, pp. 148-56.

‘R. E. Adams, J. 8. Gill, and W. E. Browning, Jr.,
““Removal of Particulate Materials from Gases Under
Reactor Accident Conditions,’ Nucl. Safety Program
Semiann. Progr. Rept. Dec. 31, 1965, ORNL-3915,
pp. 80-81,

160

Prior experiments at room temperature have shown
that a moist atmosphere may reduce the efficiency
of the same filter medium toward the same test
acrosol to as low as 23% under ~95% relative
humidity.

The effect of moisiure had not been anticipated
by reactor designers, aand, because of its impli-
cations in case of reactor accidents, recent re-
search efforts have been directed toward defining
the mechanism of moisture in reducing filtration
efficiency. The effect of moisture may be due
to changes in the properties of the acrosol, changes
in the {ilter media, or both.

A difference has been noted in the physical
characteristics of asrosols produced under humid
and under diy conditions. Election photomicro-
giraphs of an aerosol produced under high relative
humidity reveal that the particles are covered with
a thin film of unknown composition and that ag-
glomerates are in the form of chains (Fig. 18.1)
rather than clusters as observed under diy con-
ditions. Evidently the stability and filtration char-
acteristics of these two types of aercsols differ
in some respects.

Tests have established that activity profiles
obtained by filtration of wmoist and div aerosols
through fibrous-filter analyzers ate guite different.
In a drv atmosphere as much as 90% of the aerosol
passed through the fibrous-filter analyzer is re-
tained on the fiist fiber mat, whereas under high
humidity conditions a flat activity profile is ob-
tained as a result of more even distribution of
on the successive fiber mats. Inter-
pretation of results under wet conditions can be
taken to mean that the fibrous filter mat efficien-

particles

cies ate being reduced, or moisture is affecting
the aerosol in some way, making it more difficult
to filter, or both. In any case, a flat distribution
of particles through the filter pack would result.

To resolve the anomaleus filtration behavior

in the presence of woisture, a series of experi-
161

PEM—200~51421

 

Fig. 18.1. Appearance of Aerosol Formed in a High-Humidity Atmosphere. Note thin film covering the particles.
165,000, Reduced 6.5%.
ments was made in which an aerosol generated
in a high-humidity air atmosphere was sampled
by two filter packs. One pack was held at room
temperature, and the fibers were saturated with
water vapor from the high-humidity carrier gas.
To reduce the moisture content of the fibers, the
other pack was heated to 100°C under similar
conditions. In another set of experiments the
acrosol was generated in a dry air atmosphere and
once more sampled with two packs. One pack
was dry, and the other was wetted by pretreating
it with 90% relative humidity air immediately prior
to its use in an experiment. The conclusion
drawn from these experiments was that moisture
does not affect the efficiency of a fibrous filter
but that it influences the physical properties of
the acrosol, making it more difficult to filter.
However, not all the results from these experi-
ments were equally unambiguous. At aeirosol con-
centrations estimated to be at the level of 10°
nuclei/cm?, the activity profile obtained in the
presence of a wet atmosphere was not character-
istically flat. It is probable that at these higher
concentrations, agglomeration will occur regardless
Study of the effects

o
~

of the presence of moisture.
of moisture is continuing.

Construction of a recirculating aerosol facility
for laboratory aerosol studies has been proceeding
parallel to the moisture effects study just de-
scribed. This recirculating facility consists of
a 100-liter stainless steel vessel with associated
piping which will allow the behavior of an aerosol,
a scaled-down air to be
studied under a variety of accident conditions
ranging from temperatures of 25 to 120°C and
humidities from ~0 to ~100%.
aerosol facility has been installed, and shake-
down tests are in the final phase.

or cleaning system,

The recirculating

EXAMINATION OF PARTICULATE AEROSOLS
WITH THE FIBROUS-FILTER AMALYZER

M. D. Silveiman
Jack Truitt

V. E. Browning, Jr.
R. E. Adams

The fibrous-filter analyzer (FFA) is being de-
veloped for measuring the characteristics of ra-
dioactive azerosols in terms of their response to
filtration processes by determining their distri-
bution vs depth in a filter under carefully con-

trolled conditions. Moisture does not significantly

162

affect the peiformance of the FFA, although the
test aerosol itself was affected. The filtration
efficiency data agreed well with the theoretical
treatment of filtration developed by Torgeson. The
analyzer was calibrated against particles 150 to
1500 A in diameter, measured by electron micros-
copy.

A summary report on this project has been com-
pleted,® and only a brief review of the subject
material will be presented here.

The FFA is an “‘in situ’’ analytical device
which characterizes radioactive aerosols dynam-
ically by particle respoiise to the major processes
of filtration: diffusion, inteiception, and inertial
impaction. The concept of the FFA ocriginated
from a note by Sisefsky,® who determined the
penetration of ‘‘fallout’’ material in a commercial
filter by radioassay by peeling off layers of the
filters with pressure-sensitive cellophane tape.
By contrast, the fibrous-filter analyzer is made
from uniform-diameter Dacron fiber (which is formed
into a uniform web by a carding maching) into a
layered structure to facilitate separation of the
fiber bed into disctete layers for radioassay. The
test aerosol containing ®3Zn was produced by
using a Tesla coil to generate a spark between
two preirradiated zinc electrodes. A stream of
air passing over the electrodes carried the aerosol
through the system containing the filters. Elec-
tron micrographs of samples of the aerosol col-
lected on carbon-covered Millipore membrane filters
yielded information regarding the size of the par-
ticles. Depending on the experimental conditions,
acrosols have been prepared over the size range
50 to 10,000 A (0.05to 1 u).

The data were analyzed graphically by means
of the Chen equation’ to estimate single-fiber
efficiencies. These were compared with thecretical
fiber efficiencies calculated according to Torge-
who used an adaptation of Davies’?

son, ? inter-

ception and impaction theoty combined with a new

 

SM. D. Silverman et al., Characterization of Radio-
active Particulate Aerosols by the Fibrons-Filter Ana-
iyzer, ORNL-4047 (in press).

6]. Sisefsky, Nature 182, 1438 (1958).
7C. Chen, Chem. Rev. 55, 595 (1955).

L. Torgeson, **The Theoretical Collection Ef-
ficiency of Fibrous Filters Due to the Combined Effects
of Inertia, Diffusion, and Interception,’’ paper No.
J-1057, Applied Science Division, Litton Systems, Inc.,
St. Paul, Minn., 1963.

°c. N. Davies, Proc. Inst. Mech. Enges. (London) 1B,
185 (1952).

By
163

interception and diffusion theory. The Torgeson
theory was selected by Whitby!? as that which
agreed most closely with experimental data ac-
cumulated in numerous researches. A computer
program developed specifically for these FFA
calculations is given in Appendix II of the sum-
mary report, >

Calibration of the FFA was performed by com-
paring particle sizes estimated from electron pho-
tomicrographs of the inlet and outlet aerosols with
those calculated according to the Torgeson treat-
ment. The best correlations were observed in
dry experiments and at low velocities. Additional
calibrations will be performed at the University
of Minnesota under a subcontract.

DISTINGUISHING iODINE FORMS
AT HIGH TEMPERATURES AND HUMIDITIES

R. L.. Bennett
W. H. Hinds

R. E. Adams
Zell Combs

In the event of a nuclear accident, radioiodine
is the most hazardous fission product which may
be released. lodine may exist in elemental or in
chemically combined species in molecular form
or as aerosols dispersed on particulate material
from the fuel or structural members of the reactor
core. The various species exhibit different be-
havior toward removal, and information is needed
on their behavior in order to design adequate gas
cleaning systems. A program is in progress to
develop tools for distinguishing the various vapor
forms of iodine which occur in the laboratory and
in larger-scale experiments. The identifying de-
vices should be capable of remote operation and
pteferably should keep their effectiveness under
extreme conditions of temperature and humidity
that may occur in a reactor accident.

The analytical devices most commonly used for
radioiodine studies are composite diffusion tubes
and May packs. The adverse effects of high
humidity on the respoanse of diffusion tubes and
the application of selective desiccants and im-
pregnated charcoal linings in improving their per-
formance have been reported.'!''?  Extensive
tests of May packs under elevated temperatures
and high-humidity conditions, such as those ex-
pected in the LLOFT experiments, are in progress.

 

19, T. Whitby, ASHRAE J. 7(9), 5665 (1965).

The May pack is an assembly of filter materials
and adsotbent beds intended to separate iodine
forms of different reactivity or adsorption tend-
ency. Since high specificity is difficult to ob-
tain for wide ranges of temperature and humidity,
considerable testing is needed to establish the
range of reliability of the optimum components
of the pack. Initial emphasis has been placed
on a configuration suggested for the LOFT pro-
gram. This arrangement consists of a sequence
of three high-efficiency filters, eight silver screens,
five charcoal-impregnated filters, two 3/j‘—in, char-
coal beds, and finally one more high-efficiency
filter section. The initial filter section is in-
tended to remove particulate forms of iodine, the
silver screens remove elemental iodine, and the
charcoal filters retain the iodine species which
are not removed by the silver screens but which
are easily adsorbed by charcoal-impregnated filter
papers. The more penetrating forms, such as
methyl iodide, are adsorbed in the charcoal beds.
The last high-efficiency filter section is designed
to trap any charcoal particles which might be
dislodged from the beds.

Duplicate May packs were usually tested with
two associated diffusion tube assemblies,
at room temperature under dry conditions and the
other at the same conditions as the May packs.
The tests were made at 90°C with a superficial
face velocity of 10 fpm. Dry or 90% relative-
humidity air streams were used in tests with
elemental iodine and methyl iodide.

Results and discussion of 2 large number of
these tests have been reported. 13

one

Briefly, the effect of moisture on penetration
of methyl iodide into the pack is illustrated in
Fig. 18.2. Under dry conditions the CH,I was
about evenly distributed between the charcoal-
loaded filters and the first chawcoal bed, while
at 90% relative humidity most of the CH,I was
swept into the beds. The sharp separation of a

 

1R, E. Adams ef al., **Characterization and Behavior
of Varicus Forms of Radiociodine,*” Nucl. Safety Pro-
gram Semiann. Progr. Rept. Dec. 31, 1965, ORNL-3915,
pp. 10111,

2R, E. Adams, R. L. Bennelt, and W. E. Browning, Jr.,
Characterization of Volatile Forms of lodine at High
Relative Humidity by Composite Diffusion Tubes, ORNL-
3985 (August 1966}

13R. E. Adams ef al., “*Characterization, Control, and
Simulation of Fission Products Released Under LOFT
Conditions,?® Nucl. Safety Program Ann. Progr. Rept.
Dec. 31, 1966 (to be issued).
164

mixture of 16% CH.,I with eclemental iodine is
showa in Fig. 18.3. The elemental iodine de-
posited on the first two sections, and the CHEI
deposited on the charcoal beds.
of the CH,I into the second bed can be greatly
reduced by use of iodine-impregnated charceal.

The penetration

Several tests have revealed that the elemental

section is erratic and often large. Filter materials
investigated include Hollingsworth-Vose HV-70,
Millipore AP-20, Flanders F-700, Cambridge 1G,
Reeves Angel 934AH, and Zitex 5- to 10-p pore
membranes. It appears that separation of par-
ticulate iodine from elemental iodine by use of

a high-efficiency filter in the first section of the

iodine deposition on the first high-efficiency filter May pack is not reliable under the conditions

ORML--DWG ©65-8204

 

 

 

 

   

 

 

 

 

 

 

100 [ e e -
SAMPIE: CHjl RELATIVE HUMIDITY
TEMPERATURE: 90°C Y DRY
80 1 FLow: ~10 fpm BT g8 ~o0m —
. B0 I
=z
L
O
n o]
Ll
8. 40 R *\ e
20 — ——¢——§§§ —
O e e s NN R
HIGH - SILVER  CHARCOAL CHARCOAL CHARCOAL HIGH -
EFFICIENCY  SCREENS FILTER BED BED EFFICIENCY
FILTER PAPER FILTER
(3} {8) (5) (0.75 in.) (0.75 in.) (3)
HV-70 ACG/B PCB PCB HV 70

Fig. 18.2. Retention of Methyl lodide by May-Pack Compenents in Dry and High-Humidity Air Streams.

 

 

 

 

 

 

 

 

 

 

100 ORHL-DWS S6- 3205
80 o TEMPERATURE: 9Q°C B
V FLOW: ~10 fpm
% RELATIVE HUMIDITY: ~90%
pm 60 [ .- e e - _ -
=
i
L
Q. 40 _ . % -
20 .. f\‘é_‘_ - - e o
W 4 E \ e )
HIGH - SILVER CHARCOAL CHARCOAL CHARCOAL HIGH-
EFFICIENCY  SCREENS FILTER BED BED EFFICIENCY
FILTER PAPER FILTER
(3) (8) (5) (075 in.) (0.75 in.}) {3)
F-700 ACG/B PCB PCB F-700

Fig. 18.3. Retention of Mixed Methyl lodide and Elementcl lodine by May-Pack Components in a High-Relative-
Humidity Air Stream.
tested. Alternate pack configurations without the
initial section of filters are being studied. Both
silver screens and silver membrane filters were
found to be effective for removal of elemental
iodine from streams of 90% relative humidity. The
charcoal beds should be of the iodine-impregnated
type for maximum retention of methyl iodide under
moist conditions.

REACTIONS OF {ODINE YAPOR
WITH ORGANIC MATERIALS

RB. E. Adams
Ruth Slusher

R. L. Bennett
Zell Combs

When radioiodine is released into & closed en-
vironment, such as in a containment test facility,
it has been noted that a generally small, but pos-
sibly significant, fraction may appear in the form
of methyl iodide and other alkyl icdides.!* -In
an accident situation, methyl iodide may be formed
in the containment atmosphere by gas-phase re-
actions with organic contaminants, or on the var-
ious types of surfaces within the containment
vessal with subsequent desorption into the gas
phaze. This investigation, which is an extension
of some earlier efforts,?5'1% iz concerned with
the methods of formation of methyl iodides to
avoid, if possible, conditions or materials con-
ducive to its fermation in a reactor systewm.

Possible reactions beiween painted surfaces and
elemental iodine to produce methyl iodide have
been investigated.

Preliminary work has been done using gas chro-
matography, with electron capture detectors for
analysis of the reaction products. Calibration
of methyl iodide response has been made with
samples in the gas and in the liguid phase (cy-
clohexane).

The first experiments were performed using a
reaction vessel painted with two coats of Amercoat
64 primer and two coats of Amercoat 66 seal coat

 

e E. Adams et al., The Release and Adsorption
of Methvl Jodide in the HFIR Maximum Credible Ac-
cident, ORNL-TM-1291 (Oct. 1, 1965).

ISW. E. Browning, Jr., et al., *Reaction of Radio-
iodine WVapors with Organic Vapors,® Nucl. Safely Pro-
gram Semiana. Progr. Rept. June 30, 1965, ORNIL.-3343,
pp. 18791,

ey, E. Browning, Jr., et al., “Reaction of Radio-
iodine Vapors with Organic Vapors,? Nucl. Safety Fro-
gram Semiann. Progr. Rept. Dec. 31, 1965, ORNL.-3915,
pp. 99160,

165

and heated at 100°C. Periodic sampling gave
chtomatograms with about six major peaks, none
of which fell near the methyl iodide peak. About
1 mg of elemental iodine was placed in another
painted vessel and heated at 100°C. Chroma-
tographic analysis indicated the presence of methyl
iodide, with a total mass in the vapor phase of
about 107° mg. Additional studies will be made
with coating formulations of interest in the LOFT
program.  When a dual detection chromatograph
which is on order becomes available, it will be
used to examine the paint and reaction product
vapors in mote detail. These techniques will
also be employed to investigate the reaction of
iodine with trace organic components in the gas
phase.

BEHAVIOR OF FISSION PRODUCTS
IN GAS-LIQUID SYSTEMS

R. E. Adams B. A. Soldano
W. T. Ward

Removal of fission product dispersions from
containment atmospheres by application of liguid
systems has been proposed as an engineered safe-
guard. Examples of these systems are pressure
suppression pools and containment sprays. An
experimental program has been initiated to study
both the chemical and physical aspects of such
gas-liquid systems. A study of the adsorption
of gases, or particles in gases, by a liquid in a
spray system involves a situation wherein there
exists a large amount of gas and a relatively small
amount of liguid under highly dynamic conditions.
On the other hand, a study of fission product
trapping in a pressuare suppréssion pool represents
a situation in which there is a very large amount
of water under relatively static conditions in con-
tact with a small amount of gas. Since the two
systems involve the two extremes of the gas-
liguid spectrum, the study has been divided into
two parts. This dualistic approach will hopefully
permit extrapolation of information to intermediate
conditions and will allow one to {ix experimental
conditions such that a chemical-physical descrip-
tion of our experiments becomes feasible.

An experimental study of the efficacy of water
sprays in the removal of released fission products
in reactor containment vessels requires a knowl-
edge of the hydrodynamics of these systems as
well as an understanding of their kinematic be-
havior. Since single liquid drops under highly
dynamic conditions constilute a primary element
of the spray itself, we propose to study the be-
havior of liquid drops suspended in a low-velocity
wind 1.7  Such a tupnel can be used to
simulate the gas-liquid environment accompanying
a fission release accident. Some of the pertinent
variables are drop size, height of fall, time of
contact, composition of both gas and liquid phases,
pressure, temperature, and the volume of tunnel

tunne

gases,
The advantages of a wind tunnel in such dy-
namic studies aze, in part, as follows:

1. The time of contact of each drop with the gas
streaim can be widely varied.

2. Drops suspended by the gas stream in the tun-
nel can be directly observed and photographed
so that shape and oscillation factors can be
properly accounted for.

Random convection effects are eliminated.

4. Close temperature control can be achieved in
the working area.

5. Homogeneity of the gas mixtures and therefore
reproducibility of results is, in principle, at-
tainable with such a probe.

The second part of this program involves a study
of the behavior of fission products in a pressure
suppression pool. General Electric Company is
incorporating such pools in their design of com-
mercial power reactors and, for this reason, is
conducting research into the behavior of suppies-
sion pools during and subsequent to a reactor
accident. Negotiations are under way with General
Electric, San Jose, for a subcoitract under which
they would perform theoretical ana laboratory in-
vestigations of the effectiveness of pressure sup-
pression pools for fission product trapping.

At present, engineering designs have been com-
pleted on both the wind tunnel and the supporting
drop collection equipment. It is estimated that
approximately 75% of the paris have been fabri-
cated.

Prior to initiation of the wind tunnel! investi-
gation, studies were undertaken to determine the
efficiency of various solutions in removing methyl
iodide from air. The experimental procedure con-

 

1. H. Garner and R. Kendrick, Trans. Inst. Chem.
Engrs. (London) 37, 155—61 (1959).

166

sists in bubbling air containing methyl iodide vapor
through a 1 1/2—1'n.—diam glass column containing 800
ml of the solution at approximately 25°C (depth
of solution = 28 in.) for approximately 2 hr, fol-
lowed by a “‘clean’ air sparge of from 16 to 20 hr.
The air entering the bottom of the column is
dispersed through a porous glass disk. The air
leaving the column is passed through two or three
beds of iodine-impregnated activated charcoal to
remove the methyl iodide that is not captured by
the solution. lodine-131 tracer is used; the ra-
dioactivity of the charcoal and solution indicates
the iodine distribution. All the activity on the
charcoal was found to be in the first bed.

A comparison of the efficiencies of the various
solutions tested to date is given in Table 18.1.
Tests of other solutions are planned, as well as
tests to determine the effect of increasing the
temperature.

Table 18.1.
for Methyl lodide Remava! from Air

€fficiency of Scrubbing Solutions

 

Amount

of Activity

 

Solution Concentration Retained by
Solution (%
Distilled water 0
Sodium hydroxide 0.01 M 0
Hydrogen peroxide 15 wt % 0
Iodic acid a1 m 0
Sodium acetate 1.2 M Q
Ammonium hydroxide 0.5 M 16
Potassium iodide 1M 25
Sodium thiosulfate 0.01 M 59.9
0.05 & 87.3
0.10M 93.5
0.25M 26.8
Hydrazine 27 wt % 99.8
Salution A® 0.1 wt % 84.7
1% 99.6
5% 99.92
12%, 99.99.+

 

%The identity of solution A is being withheld pending
patent svaluation by the USAEC.
HIGH-TEMPERATURE BEHAVIOR
OF GAS-BORNE FISSION PRODUCTS.
TELLURIUM DIOXIDE

M. D. Silverman A. P Malinauskas

Recent experimental studies'® indicate an en-
hanced volatility of a rather broad class of metal
oxides in the presence of water vapor. This en-
hancement is believed to be the result of the
generalized reaction

oxide(s) + H,0(g) = hydroxide(g) , (1)

in which the hydroxide so formed is stable only
at high temperatures. The present investigations
are directed toward a study of this reaction.

The research has been initiated using tellurium
dioxide, TeQ,, as the compound under investi-
gation. Although reaction (1) has yet to be un-
equivocally established in this case, the increase
in the apparent vapor pressure, due to water vapor,
has been experimentally demonstrated over the
temperature range 600 to 700°C. /1% Qur imme-
diate objective is to verily these data and to
extend the temperature range.

The transport method (also called the “transfec®’
or ““trauspiration’ method) has been chosen for
In brief, the pro-
cedure involves saturating a suitable carrier gas
with the material under study in one section of
the apparatus and allowing this substance to be
transported and collected in another region. A
detailed description of the experimental and the-
oretical aspects can be found elsewhere, '8—%0

Since the program has only recently been ini-
tiated, much of the work has been concerned thus
far with the design, construction, and testing of
the apparatus itself. Data obtained from two of
the preliminary runs, however, rather dramatically
demonstrate the effect of water vapor on the vol-
atility of TeO,  These data are presented in
Table 18.2.

nse in the cument research.

 

%0, lemser and H. G. Wendlandt, Advan. Inor. Ra-
diocchem. 5, 215--58 {1963).

190, Glemser and R, v. Haeseler, Naturwissenschaften
47, 467 {1960) O. Glemser, R. v. Haeseler, and A.
Muller, Z. Anorg. Allgem. Chem. 329, 51 (1964); O.
Glemser, A. Muller, and H. Schwarzkopf, Nafturwissen-
schaften 52, 129 (1965).

2%0). Glemser and R. v. Haeseler, Z. dnorg. Allgem.
Chem. 316, 168 (1962).

167

Toble 18.2. Effect of Water Vapor on the Volatility
of TeO, (Preliminary Results) ~

 

Sample temperature (°C) 653 661

Duration of experiment (hr) 5 26

Oxygen carrier gas 7.00x 107*  7.85% 1074
flow rate {moles/min)

H,O transported (g) 19.92 0

Te02 transported (g) 0.0245 0.0016

Apparent vapor pres- 8.72x 1072 0.61x 1072

sure of T602 {torrs)

 

THE CASCADE IMPACTOR AS A TOOL
FOR THE 5TUDY OF SIZE DISTRIBUTION
OF FISSION PRODUCT AEROSOLS

G. W. Parker H. Buchholz??

Cascade 22 are instruments used to

separate aerosols into fractions of a discrete range
of particle size. They are frequently used for
analyzing radioactive aerosols to collect a range
of size groups which may later be examined for

impactors

their radioactivity. Several thecretical and ex-
perimental studies of cascade impactors have been
reported. 23727 Impactors are limited to relatively
large particles, usually above 0.5 g, and, in order
to make this instrument genuinely useful in nu-
clear safety research, it is necessary to extend
their range to particles below 0.1 u. The basic
theory of these devices is quite simple. When
a gas jet carrying particles is directed toward a
surtace, all particles having sufficient inertia in
the first stape will leave their stream lines . and
settle on the sutface. Smaller particles will re-
main within the jet stream. In the cext stage the

 

2IVisitimg scientist on assipnment from Hahn-Meitoer
Institute, Nuclear Research, Reriin.

22K, R. May, J. Sci. instr. 22, 187~93 (1945),

23]. C. Couchman, Use of Cascade Impactors for
Analyzing Airborne Particles of High Specific Gravity,
CONEF-650407, pp. 11631203 (1965).

‘4. L Mitchell and J. M. Pilcher, Design and Cali-
bration of an Improved Cascade Impactor for Bize Ana-
lysis of Aerosols, TID-7551, pp. 6784 (April 1958).

25, Mercer, M. I. Tillery, and C. W. Ballew, 4
Cascade Impactor Operating at Low Volumeiric Flow
Rates, LF-5 {December 1962).

25}. J. Cohen and I, N. Mortan, Theoretical Consid-
erations, Design, and Evaluation of a Cascade Impactor,
UCRI-14440, Rev. 1 (June 1966).

27A. R. McFarland and H. W. Zerller, Study of a Large-

Volume Impactor for High-Altitude Aerosol Collection,
TID-18624 (April 1963).
gas passes through smaller holes, and the jet is
accelerated to a higher velocity. The probability
of smaller particles settling is thus increased.

If the pressure within the cascade impactor is
lowered until the particle diameter is comparable
to the mean free path of the gas molecules, there
are fewer collisions between particles and gas

molecules. Particles having small inertia are able

168

under reduced pressure to leave the jet. By this
slip effect, described by the Cunningham correction,
the cascade impactor becomes more efficient for
separating smaller particles. Using an equation
based on May’s theory,?? we calculated the size
of particles that have a 50% probability of de-
positing on each stage of the Andersen sampler.
The results for the case where the inlet gas flow

ORNL-DWG 87-708

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

    

 

 

 

 

 

 

 

 

 

 

 

40

2a ‘h\ -
. \ p:’l?f)omm Hg
1
-~ 40 e S
=
e
=
w
O
Q.
ul
o 4 —
W
O
>._
’__
S 2pe
=
<
m
O
n
o {
3
O
0
<1
2
< 04 -
>
<J
T
0
G2 “eemne
©
r
<1
n.
| 04
1.
o
ot
w
}_
Ll
=
< 0.04
o Q=100cc/sec

e 7= g/cc
———n =6 g/cc }
o2 —m—— ’ —_—
o0 L
o { 2 3
STAGE NO,

Fig. 18.4.
Particle Density.,

Calculated Porticle Distribution in the Andersen Sampler as a Function of Internal Fressure and
rate is 100 e¢m?/sec and the density is 1 or 6
g/cm? are displayed in Fig. 18.4. It is clear from
this graph that pressures less than 40 mm Hg will
allow the cascade impactor range to be extended
to particles smaller than 0.1 g

The original Andersen sampler was modified for
subatmospheric operation. Each stage of the sam-
pler is connected to a separate manometer to allow
measurement of differences in internal pressure.
The modified. apparatus will be calibrated by col-
lecting particles directly on electron microscope
grids at each stage.

REACTION OF MOLECULAR IODINE
AND OF METHYL {ODIDE
WITH SODIUM THIOSULFATE SPRAYS

G. W. Parker
W. J. Martin

G. E. Creek
N. R. Horton

Addition of chemicals to water used in reactor
containment misting sprays to aid in the rapid
scavenging of radioiodine from steam-filled reactor
coitainment shells has been proposed in many
recent applications for power reactor construction
permits presented to the Atomic Energy Commis-
sion. Pure water sprays do.not provide the de-
sired iodine removal rate.

The removal of heat in most latge reactor de-
signs is effected primarily by ait-recirculation
cooling units, and the removal of inorganic vapor
forms of jodine (I, and HI) is the major require-
ment of the sprays. Griffiths *® described a method
for predicting the rate of iodine removal from
gases by sprays and presented most of the pa-
rameters for making such calculations in a com-
prehensive report which was summarized in Nuclear
Safety.?? The most useful experimental exami-
nation of Griffiths’ theory was performed by Tay-
lor, % using glass columns in which he varied
both the liquid flow rate and the gas velocity.
From these experiments the mass transfer process
through the liquid film was confirmed. It is clear
that elemental iodine, as it is absorbed by a

 

28y Griffiths, The Removal of Iodine from the At-
mosphere by Sprays, British Report AHSB(S)R-45 (De-
cember 1962). '

29y Griffiths, Nucl. Safety 6(2),
1964—65).

30p . Taylor, Chem. Eng. Sci. 10(1/2), 68~80
(1959).

186—94 (Winter

169

liguid drop containing reducing material, is chemi-
cally altered to the form of iodide by the reducing
agent. Thus, there is a sharp reduction in the
vapor pressute of iodine near the liquid surface,
and, since the total gquantity of iodine available
from a reactor is almost negligible (by comparison
with the quantity of reducing agent in the spray),
there is essentially no change in efficiency as
the drop continues to fall. The iodine present
as HI is even more soluble in the droplet, and
it should be removed at a faster rate than [,.

In order to test the spray removal concept under
conditions used in fission product transport studies
and in order to be able to assess the limitations
of the process for the nonreactive forms of iodine,
we decided to carmry out spray experiments in the
Containment Mockup Facility (CMF) with charac-
teristic forms of radioiodine. The design of the
spray system allowed very small gquantities of
liquid to be used, and therefore the liquid-to-gas
ratio could be adjusted to reactor containment pro-
portions.

The spray system chosen for test contained six
atomizing nozzles, each of which delivered 0.1
liter/min at 80 psig. Since the test containment
was small (180 liters), the sprays were generally
operated for only 15 to 30 sec. Operation fot
30 sec produced a liguid volume of 0.3 liter or
1/600 of the containment tank volume. A con-
centration of Na,§,0, of 0.1 mole/liter (1.6%)
was used in these tests.

Test with Molecular lodine (1,)

The conditions selected for the molecular io-
dine test were typical of those used in previous
CMF steam-air runs. The initial pressure was
about 35 psig, of which about 20 psig was due
to steam. The tank atmosphere temperature was
about 130°C at the injection time, and the cooling
rate without sprays was about 30°%hr. Since the
natural removal rate for molecular iodine in the
CMEF (fuz = 30 min) was fairly well known, the
effect of thiosulfate sprays on the removal rate
could be readily calculated. We found that ad-
dition of sodium thiosulfate spray at a rate of
1/300 of the containment tank volume per minute
decreased the half-life of disappearance of Izlfrom
the tank atmosphere to 3.9 sec. This value agreed
quite well with the calculated value of 2.6 sec
for the half-time of removal of 1, by 200-u droplets
of 0.1 mNa,S,0,.
Tests with Methyl lodide (CH3I)

In two experiments conducted in a similar man-
ner to the molecular iodine experiment described
above, methyl iodide injected at similar concen-
trations (2 mg/m>) was first exposed to steam
and to the condensation process and then to
sodium thiosulfate spray. The results were en-
couraging, but the rate of removal of CH,I is
only a fraction of the corresponding iodine rate.
However, the CH ,I removal rate is of sufficient
magnitude to warrant its consideration in safety
The calculated half-times
for depletion of CH.I were approximately 2 hr in
one run and 4 hr in the other.

analysis calculations.

Conclusions

A model for removing reactive iodine by thio-
sulfate sprays was tested in the stainless steel
CME tank.
culated and the experimental iodine removal rates
was found.

Tests of methyl iodide removal with Na,S 0,
sprays gave encouraging results but showed much
slower removal rates than the corresponding io-
dine rates. More efficient scavenging agents will
be required for methyl iodide removal.

Excellent agreement between the cal-

STUDIES OF CSE-TYPE FISSION
PRODUCT SIMULATION

G. W. Parker R. A. Lortenz
N. J. Horton

Fission product aerosols will be simulated in
experiments to be performed in the Containment
Systems Experiment (CSE) at Hanford using a dif-
ferent technique from that used at ORNL.. Rogers3!
pointed out that, in the 30,000-ft? CSE contain-
ment tank, use of irradiated fuel to furnish real-
istic fission product levels is impractical. Neither
is it feasible to use simulated high-burnup fuel
pellets of the type used in the CMF?®? aad in

g, J. Rogers, Program for Containment System FEx-
periment, HW-83607 (September 1964).

326y, W. Parker et al., **Simulation of High-Buinup

U0, Fuel in the Containment Mockup Facility,*’ Nucl.
Safety Program Semiann. Progr. Rept. Dec. 31, 1964,
ORNL.-3776, pp. 70--74,

170

the NSPP.3% The simulation technique devised
for the CSE experiments, described by Hilliard
and McCormack,** iavolves vaporization of suit-
able quantities of fission product elements con-
taining radioactive tracers and passing the vapor
over molten unirradiated UO, before it enters the
containment vessel. In order to determine how
well the aerosols produced by this teclinique imi-

tate those produced by overheated high-burnup

fuel, it will be necessary to make direct com-
parisons under similar conditions. We plan to
do this in the CMF, where experiments with

high-burnup fuel have already been performed, and
perhaps later in the Containment Research Instal-
lation (CRI).

We have completed the design, construction, and
preliminary testing of equipment for performing
CSE-type simulation experiments either in the CMF
or the CRI (the fuel-melting cans of these sys-
tems are interchangeable). It was necessary to
modify the experimental arrangement described
by Hilliard and McCormack®* rather extensively
in order to adapt it to the CMF-CRI fuel meltdown
arrangement, but the differences relate mainly to
methods of getting the vaporized materials into
the pressurized meltdown furnace. In addition,
we chose to provide a steam-air environment in
the vicinity of the molten UO, and in the con-
tainment tank, both at 30 1b total pressure, rather
than air, because our recent experimeits with
high-burnup fuel were all carried out using a steam-
air atmosphere.

As shown by Fig. 18.5, two ribbon heating units
are inserted through the glass envelope, which
is fitted to the end of the quartz meltdown tube
by means of a tapered joint. A platinum ribbon
which can be operated at a temperature between
1400 and 160C°C in the furnace tube atmosphere
will be used to vaporize tellurium (in the form
of TeO,), cesium (intioduced as Cs,CO,), and
ruthenium metal. Cesium will probably be vapor-
ized as the metal but will quickly be reoxidized
in the furnace atmosphere. Ruthenium metal will
undoubtedly be converted to a volatile oxide, RuO,

 

3. p. Parsly et al,, “*Transport Behavior of Fission
Products in the Nuclear Safety Pilot Plant,’”’ Nucl.
Safety Program Semiann. Progr. Rept. Dec. 31, 1965,
ORNI1.-3915, pp. 44--51.

34R. K. Hilliard and J. D. McCormack, **Simulation
in the Containment Systems Experiment,”’ pp. 588602
in International Symposium on Fission Product Relsase
and Transport Under Accident Condifions, Oak Ridge,
Tennessee, April 57, 1965, CONIT-650407.
171

. PHOTO 84154

 

Fig. 18.5. Glass Envelope Used in CSE Fiss}ion Product Simulation Tests, Showifig Platinum (Lower) and
Tungsten (Upper) Ribbon Filaument Heating Elements,
or RuO,. The other heater, which has a tantalum
ribbon (1/8 in. wide and 5‘2 in. long, V shaped),
operates in a helium atmosphere in a small glass
envelope having a 5”1 c-in.-diam hole through which
helium cariying the vaporized material flows into
the furnace tube. A mixture of BaCO, and finely
divided zirconium metal is placed on the tantalum
ribbon to increase the volatility of barium by re-
ducing the oxide to the more volatile metal.

Iodine in the form of I, is introduced through
a side arm attached to the outside pressurized
tube extension. A glass capsule containing the
I, is inserted in a Teflon-lined side arm, and it
is crushed while a stream of air flows through
the tube to carry the iodine into the inner (fur-
nace) tube.

Finally, steam is introduced through a ball joint
at the end of the glass envelope. Water supplied
under pressure at a carefully controlled rate is
converted to steam by a miniature water vaporizer
located in the outer pressurized shell, quite close
to the ball-joint entrance to the furnace tube.

Preliminary testing has centered on achieving
satisfactory volatilization of barium, the least
volatile species expected to be used in these ex-
periments. We found that a temperature of 1800°C
was needed to volatize half the barium activity
from a BaCO -Zr mixture in a flowing helium at-
mosphere.

RETENTION OF RADIOACTIVE METHYL
IODIDE BY IMPREGNATED CHARCOALS

R. E. Adams J. D. Dake??
R. D. Ackley J. M. Gimbel?®
F. V. Hensley

Methy! iodide, which is more difficult to trap
than elemental iodine, may be generated in re-
actor accidents. It readily penetrates beds of
the common types of activated charcoal at ambient
temperatures except when the relative humidity
is low. However, certain specially impregnated
(iodized) charcoals have been observed to have the
capability of effectively trapping radoactive methyl
iodide from air streams of fairly high relative hu-
midity at temperatures as high as 115°F.37:38
These charcoals, which are impregnated with one

3°Co—op student, University of Tennessce.

3'5Co-op student, Drexel Institute of Technology.

172

or more iodine-containing substances, appear to
possess this unusual capability as the result of
an isotopic exchainge mechanism. ‘To obtain in-
formation pertaining to the applicability of im-
pregnated charcoals under various reactor accident
situations is the objective of this work, which
has been reported in more detail elsewhere. 3940

A rather large number of screening tests on
various laboratory-impregnated charcoals and on
various types of commercially impregnated char-
coal were performed. The manner of conducting
these tests is illustrated in Fig. 18.6. The tests
are made at ambient temperature and pressure
and usually at around 70% relative humidity. Four
commercial products were cbserved to be effective
for CH,'*'l trapping. They arc BC-727 (from
Barnebey-Cheney), MSA-85851 and MSA-24207 (from
Mine Safety Appliances Company), and G601 (from
North American Casbon, Inc.). Results from these
room-temperatuie tests for the four commercial
charcoals are given in Table 18.3. A number of
laboratory-impregnated charcoals also gave prom-
ising results, although at present none of them
are regarded to have ainy especial advantage over
the commercial charcoals mentioned.

The conditions corresponding to the earlier
tests>® and those discussed above are less severe
than the conditions which have been postulated
for the atmosphere in a reactor containment ves-
sel in which stean is released, causing consider-
able elevation of temperature, pressure, and hu-
midity. Consequently, tests have also been made
under the more severe conditions characteristic
of steam-air systems, Methy!l iodide labeled with
CH313II is employed. Typical results are shown
in Fig. 18.7, where the deleterious effect of very
high relative humidity is displayed. The use of
charcoal laboratory impregnated with triethylene-
diamine was prompted by results of United King-
dom researchers.*’ To the extent they have been
tested, the other commercial charcoals identified

37R. E. Adams et al., The Release and Adsorption of
Methy!] lIodide in the HFIR Maximum Credible Accident,
ORNL-TM-1291, pp. 2426, 40 (Oct. 1, 1965).

38,
Progr.

*9R.
Progr.

D. Ackley et al., Nucl. Safety Program Semiann.
Rept. Dec. 31, 19265, ORNIL.-3915, pp. 61.-80,

Y. Adams et al., Nucl. Safety Program Ann.
Rept. Dec. 31, 1966 (to be issued).
40

K. E. Adams, R. D. Ackley, and W. E. Browning, Jr.,
Removal of Radioactive Methy! Iodide from Steam-Air
Systems, ORNL.-4040 (in press).

‘IR, D. Collins (letter), Nucleonics 23(9), 7 (1965).
173

ORNL-DW5 66-7CHB

 

(}B— OFFERENTIAL PRESSURE TRANSMITTER

   
 

 

 

ACTIVATED CHARCOAL
8E0S BEING TESTED

 

 

 

 

STEAM DEHUMIDIFIER

 

 

  

VARIABLE OF«‘EF"CEg HOCD

EXHAUST

 

ACTIVATED CHARCDAL

AR BACKUP BEDS

-] i
FLOW CONTROL VALVE
FLOWMETER

AIR-CH4I

Fig. 18.6. Simplified Drawing of Setup for Investigation of Removal of Radioactive Methyl lodide from Flowing
Steam-Air by Impregnated Charcoals.

Table 18.3. Radioactive Methyl lodide Removal Tests on Commercially Impregnated Charcoals ot 25°C

Charcoal bed diameter: 1 in.

Charcoal bed depths: 0.5, 0.5, and 1 in. in series or 1 and 1 in.
Air velocity (superficial): 40 fpm

Duration of air flow measured from start of CH3I injection: 6 hr

Duration of CHSI injection: 2 hr (1st 2 of above 6 hr)

 

gy 131
Amount of CH 1 CH, " 7'I Removal

 

 

Relative . . Efficiency (%) for

Charcoal, Mesh Size Humidity Injected Relative Bed Depth of:

%) to Amount :
of Charcoal (mg/g) 0.5 in. 1 in. 2 in.
Unimpregnated activated charcoal, 6 x 16 70 1.4 4.6 8.9 17.3
MSA-B5851 lot No. 23, 8--14 (11)® 70 1.4 56.3 82.3 97.1
MSA-85851 1ot No. 53, 8-14 68 1.4 52.0 79.2 97.0
MSA-85851 No. 93066, 8--14 (3) 70 1.0 c 88.0 98.7
MSA -242077 (2) 72 1.2 52.8 81.2 97.5
BC-727, 8~14 (6) 69 1.4 64.8 88.7 98.9
G-601, 12 x 16 (2) 78 0.8 < 85.9 98.1

 

®Number in parentheses denotes results are averages for that number of tests.
BNo mesh size furnished (appears to be 8~14).
“Test beds were 1 and 1 in. rather than 0.5, 0.5, and 1 in.
above behave similarly under these conditions
to the charcoals for which results are shown. In
addition to the tests at room temperatuie and at
approximately 280°F, a series was also conducted
at ~212°F.

According to the various results obtained and
subject to certain qualifications, a number of com-
available

mercially impregnated charcoals are

highly effective for trapping radioactive methyl

ORNL-DWG 66— 7637

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

100 i ——pr——
| J
9 - e - ‘ -
80 — - N
|
70 ‘ -
5
Z 80
g |
o
'“E IMPREGNATED CHARCOAL , 2-in. DEPTH !
3 5C ® MSH B585
<>1 o BC-727
5 A 5% TRIETHYLENEDIAMINE ON PCB,
= 6 x 16 mesh :
w40 —— . - -
e CONDITIONS .
= STEAM-AIR AT SUPERFICIAL VELOCITY
=, OF 27 TO 45 fpm
r 30 STEAM--AIR FLOW CONTIMUED FOR AT
© LEAST 3 hr AFTER COMPLETION OF
CHsI INJECTION
AROUND 3.5 mg CHzI INTRODUCED :
20 p-- PER g OF CHARCOAL. ‘{ -
| |
10 | . o —
O .................... .
30 40 50 60 70 80 90 100

RELATIVE HUMIDITY (%}

174

iodide from flowing air and steam-air over a wide
range of conditions including 70 to 300°F and 14
to 60 psia. The qualifications are (1) that the
samples tested are representative of the commer-
(2) that the charcoal has not been
example, by severe weathering or
by poisoning from adsorbed foreign substances
such as oil vapor; and (3) that the prevailing
relative humidity in the charcoal does not greatly
exceed 90%.

cial material,

damaged, for

Fig.

moval

18.7.

of Radiocactive

Effect of Ralative Humidity on the Re-
Methyl lodide

Charceals at Temperatures and Pressures Around 280°F

by lmpregnated

and 60 psia.
AUTHOR(s)

Baucarella, A, L., and A, L.
Sutton

Barton, C. J., and
V. B. Cottrell

Brunton, G, D.

Burns, J. H., and
E. K. Gordon

Cantor, 8., D. G, Hill, and
W. T. Ward

Carroll, R. M., and
Q. Sisman

Davis, ®. J., T. H. Mauney,
and J. R. Hart

De Bruin, H. J., G. M.
Watson, and C. M., Blood

Fuller, E. L., Jr.,
H. F. Holmes, and
C. H. Secoy

Holmes, H. F., E. L.
Fuller, Jr., and C. 8.
Secoy

Jenks, G. H.

Keilholtz, G. W.

Keilholtz, G. W., J. E.
Lee, Jr., and R. E.
Mocore

Malinauskas, A. P.

Publications

JOURNAL ARTICLES
TITLE

Anodic Film Growth on Zirconium at Tem-
peratures from 200° to 300°C

Fission Product Release and Transport

Under Accident Conditions
The Crystal Structure of LiUFS

Refinement of the Crystal Structure of
leBeF4

Density of Molten ThF4: Increase of
Density on Melting

Fission-Gas Release During Fissioning in

UO2

Corrosion of Zircaloy 2 by Hydrogen Per-

oxide at Elevated Temperature

Cation Self-Diffusion and Electrical Con-
ductivity in Polycrystalline Beryllium
Oxide

Gravimetric Adsorption Studies of Thorium
Oxide. TI. Water Adsorption at 25.00°C

Heats of Immersion in the Thorium Oxide—
Water System. II. Net Differential Heats
of Adsorption

Prediction of Radiation Effects on Reactor

Water and Solutions

Release and Transport of Fission-Product
Iodine and Its Removal from Reactor-

Containment Systems

Irradiation Damage to Sintered Beryllium
Oxide as a Function of Fast-Neutron
Dose and Flux at 110, 650, and 1100°C

Thermal Transpiration. Rotational Relaxa-
tion Numbers for Nitrogen and Carbon

Dioxide

175

FUBLICATION

Electrochem. Technol. 4, 117 (1966)

Nucl. Safety F(2), 203 (1966)

Acta Cryst. 21(5), 814 (1966)
Acta Cryst. 20, 135 (1966)

Inorg. Nucl. Chem, Letters 2, 15
(1966)
Nucl. Appl. 2, 142 (1966)

J. Electrochem. Soc, 113, 1222 (1966)

J. Appl. Phys. 37, 4543 (1966)

J. Phys. Chem. 70, 1633 (1966)

J. Phys. Chem. 70, 436 (1566)

Trans. Am. Nucl. Soc. 9(2), 382
(1966)

Nucl, Safety 7(1), 72 (1965)

Nucl, Sci. Eng. 26, 329 (1966)

J. Chem. Phys. 44(3), 1196 (1966)
AUTHOR(s)

Malinauskas, A, P.

Marshall, W. L., and
E. V. Jones

Marshall, W. L., and
Ruth Slusher

Perez, R. B.

Quist, A. S., and W. L.
Marshall

Reagan, P. E., J. G.

Morgan, and Q. Sisman

Sisman, O.
Soldano, B. A., and P. B.
Bien

Thoma, R. E., H. A.
Friedman, and R. A.

Penneman

Thoma, R. E.

Thoma, R. E., H. Insley,
and G. M. Hebert

Thoma, R. E., and R. H.

Karraker

‘Thoma, R. K., and G. D.

RBrunton

Friedman, H. A., and R. E.

Thoma

Griess, J. C., and J. L.
English

Hitch, B. ¥., R. G. Ross,
and H. F. McDuffie

176

TITLE

(Gaseous Diffusion — the Systems He-Kr,
Ar-Kr, and Kr-Xe

Second Dissociation Constant of Sulfuric
Acid from 25 to 350° Evaluated from
Solubilities of Calcium Sulfate in

Sulfuric Acid Solutions

Thermodynamics of Calcium Sulfate
Dihvdrate in Aqueous Sodium Chloride
Solutions, 0--110°

A Dynamic Method for In-Pile Fission-Gas

Release Studies

Electrical Conductances of Aqueous Solu-
tions at High Temperatures and Pres-
sures. III. The Conductances of Po-
tassium Bisulfate Solutions from 0 to

700° and at Pressures to 4000 Bars

Performance of Pyrolytic Carbon Coated
Uranium Oxide Particles During Irradia-

tion at High Temperatures
Preface — Fission-Gas Release Symposium

Osmotic Behaviour of Agueous Salt Solu-

tions at Elevated Temperatures. Pari IV

Isomorphous Complex Fluorides of Tri-,

Tetra-, and Pentavalent Uranium

Selected Topics in High Temperature

Chemistry (book review)

The Sodium Fluoride-.L.anthanide Tri-

fiuoride Systems

The Sodium Fluoride--Scandium Tri-

filuoride System

Equilibrium Dimorphism of Lanthanide

Trifluorides

REPORTS ISSUED

Chemical Stability of Refractory Ceramics
in MSRE Fuel

Materials Compatibility and Corrosion
Studies for the Argonne Advanced Re-

search Reactor

Tests of Various Pariicle Filters for
Removal of Qil Mists and Hydrocarbon
Vapor

PUBLICATION

J. Chem. Phys. A5, 4704 (1966)

J. Phys. Chem. 70, 4028 (1966)

J. Phys. Chem. 70, 4015 (1966)

Nucl. Appl. 2, 151 (1966)

J. Phys. Chem. 70, 3714 (1966)

Trans, Am. Nucl. Soc. 9(1), 2829
(1966)

Nucl, Appl. 2, 116 (1966)
J. Chem. Soc. (A) 1964, 1825

J. Am. Chem. Soc. 88, 2046
(1966)

J. Am. Ceram. Soc. 49, 292 (1966)

Inorg. Chem. 5, 1222 (1966)

Inorg, Chem. 5(11), 1933 (1966)

Inorg. Chem. 5(11), 1937 (1966)

ORNL-TM-1406 (January 1966)

ORNI.-4034 (November 1966)

ORNL-TM-1623 (September 1966)
AUTHOR(s)

Jenks, G. H., H. C.
Savage, and E. G.

Bohlmann
Jenks, G. H., H. C.
Savage, and E. G,

Bohlmann

Keilholtz, G. W.

Kelly, M. J.

McQuilkin, F. R., D. R.
Cuneo, J. W. Prados,
E. L. Long, Jr., and
J- H. Coobs

Miller, C. E., Jr., and
W. E. Browning, Jr.

Morgan, J. G., M. F.
Osborne, and E. L.
Long, Jr.

Morgan, J. G., P. E.
Reagan, and E. L.
Long, Jr.

Nicely, V. A., and R. J.
Davis

Osborne, M. F., E. L.
L.ong, Jr., and
J. G. Morgan

Redman, J. D.

Reed, 5. A.

Rutherford, J. L., J. P.
Blakely, and L.. G.

Overholser

Savage, H. W., E. L.
Compere, W. R. Huntley,
B. Fleischer, R. E.
MacPherson, and A.
Taboada

177

TITLE

NASA Tungsten Reactor Radiation
Chemistry Studies — Final Report

NASA Tungsten Reactor Radiation
Chemistry Studies. I. Experiment
Design

Filters, Sorbents, and Air-Cleaning Sys-
tems as Engineered Safeguards in

Nuclear Installations

An Analytical Approach to Waterlogging
Fajlure

An Irradiation Test of AVR Production
Fuel Spheres in the Qak Ridge Research

Reactor

The Adequacy of Scale~Up in Experiments
on Fission Product Behavior in Reactor
Accidents., PartI. An Analysis of
Scale-Up in the U.S. Nuclear Safely

Program

The Adequacy of Scale<Up in Expetiments
on Fission Product Behavior in Reactor
Accidents, Part [1. Recommended Ad-
ditional Nuclear Safety Scale~Up Experi-

ments

Postirradiation Examination of EGCR

Prototype Capsule 03A-6

Evaluation and Irradiation Effects Studies
on Pyrolytic Carbon Coated Fuel

Particles

Some Electrical Measurements on Anodic
Films on Zirconium

Postirradiation Examination of High Bumup
EGCR Prototype Fuel Capsules

A Literature Review of Mass Spectrometric

Themmochemical Technique — Supplement I

Corrosion of Carbon and Alloy Steels in

Water and Seawater

Oxidation of Unfueled and Fueled Graphite
Spheres by Steam

SNAP-8 Corrosion Program Summary Report

FUBLICATION

ORNL~-TM-1630 (Qctober 1969)

ORNL-~TM-1403 (March 1966)

ORNL-NSIC-13 (October 1966)

ORNL-3867 (December 1966)

ORNL-TM-1512 (June 1666)

ORNL-39901 (July 1056)

ORNL-4021 (December 1966)

ORNL-TM-1378 (February 1966)

ORNI.-3923 (March 1966)

ORNL-TM-1435 (I?ecember 1965)

ORNL-TM-1511 (June 1966)

ORNL-~TM-939 (July 1966)

ORNL-TM-1612 {October 1966)

ORNL-3947 (May 1966)

ORNL-3898 (December 1965)
AUTHOR(s)

Ackley, R. D., R. E.
Adams, and W. E.
Browning, Ji.

Adams, R. E., J. S. Gill,
W. D. Yuille, L. F.
Parsly, W. E. Browning,
Jr., and C. E. Guthrie

Baes, C. F., Jr.

Bennett, R, L., R. E.
Adams, and W. E.

Browning, Jr.

Carroll, R. M., and

0. Sisman

Keilholtz, G. W., J. E.
Lee, Jr., and R. K. Moore

Blakely, J. P., and L. G,

Overholser

Keilholtz, G. W., R. E.
Moore, M. ¥. Osbeorne,
B. W. Wieland, and A. F.
Zulliger

Park. ¢, G. W., G. E. Creek,

and A. Ferreli

Savr-ge, H. C,, J. M. Raker,
M. J. Kelly, and E. L.

Compere

Silverman, M. D., J. Tritt,
W. E. Browning, Jr., L. F.

Franzen, and R. E. Adams

Watson, G. M., R. B. Perez,
and M. H. Fontana

178

BOOKS AND PROCEEDINGS

TITLE

Rewmoval of Radiocactive Methy! Iodide from

Steam-Air Systems

Performance of Filter System Under

Accident Conditions

The Chemistry and Thermodynamics of

Molten Salt Reactor Fluoride Solutions

Characterization of Volatile Iodine [Forins

In-Pile Properties of Reactor Fuels by
Oscillating Technigques

Properties of Magnesium, Aluminum, and
Beryvllium Oxide Compacts Irradiated
to Fast-Neutron Doses Greater than 1021
neutrons/cm? at 150, 800, and 1100°C

Oxidation of AT]J Graphite by Low Concen-
trations of Water Vapor and Carbon Dioxide

in Helium

Techniques for Irradiating High Tempera-

ture Materials in a Steep Flux Gradient

Retention of Methy! Iodide by Impregnated

Carbons Under Ambient Conditions

An Assembly for Irradiation of Molten
Fluoride Fuel to High Burmups in the
Cak Ridge Resecarch Reactor

Characterization of Radicactive Particu-
late Aerosols by the Fibrous Filter
Analyzer

Effects of Containment System Size on

Fission Product Behavior

PUBLICATION

Proc. 9th AEC Air Cleaning Conf.,
Boston, Mass., Sept. 1416, 1966,
CONF-~660904 (1966)

Proc. 9th AEC Air Cleaning Conf.,
Boston, Mass., Sept. 1416, 19686,
CONF-660904 (1966)

SM-66/60 in Themodynamics, vol. 1,
IAEA, Vienna, 1966

Proc. 9th AEC Air Cleaning Conft.,
Boston, Mass,, Sept. 14—16, 1966,
CONF-660004 (1966)

Proc. Intern. Symp. Capsule Irradiation
Experiments, Pleasanton, Calif.,
May 3—5, 1966, TID=7697
(September 1966)

Proc. Conf, Nuclear Applications of
Non-Fissionable Ceramics,
Washington, D. C., May 9—11, 1966,
ed, by Alvin Boltax and J. H.
Handwerk, Interstate, Danville, I11.
(1966)

Carbon 3, 269.-75 (1966)

Proc. Intem. Symp. Capsule [rradiation
Experiments, Pleasanton, Calif.,
May 3-5, 1966, TID-7697 (September
1966)

Proc. 9th AEC Air Cleaning Conf.,
Boston, Mass.,, Sept. 1416, 1966,
CONF-660904 (1966)

Proc. Intem. Symp. Capsule Irradia-
tion Experinents, Pleasanton, Calit.,
May 3--5, 1966, TID-76G7 (September
1966)

Proc., 9th AEC Air Cleaning Conf.,
RBoston, Mass., Sept. 14—16, 1966,
CONF-660904 (1966)

Proc. 9th AFEC Air Cleaning Conf.,
Boston, Mass., Sept. 14—16, 1966,
CONF-660934 (1966)
AUTHOR(s)

Thoma, R. E.

Weaver, Clayton F.

Compere, K, L., and E. G.

Rohlmann

Keilholtz, G. W., and
C. C. Webster

Thoma, R. £., M. R.
Benmnett, and J. W.

Ullmann

179

TITLE

The Rare Earth Halides

Complex Compounds in the Sodium

Fluoride—~Rare Earth Trifluoride Systems

THESIS

Kinetics of Formation of Xenon Fluorides

PATENTS

Method of Removing Hydrogen from Liquid
Alkali Metals

Method for Analyzing Inert Gas for

Presence of Oxygen or Water Vapor

Method for Processing Aluminum-Contain-

ing Nuclear Fuels

PUBLICATION

Progress in the Science and Technol=
ogy of the Rare Earths, vol. 11,
pp. 90122, Pergamon, New York,
1966

Rare Earth Res. III, 561--70 (1966)

Thesis submitted in partial fulfillment
of the requirements for the FPh. D.
degree, University of Califomia,
Berkeley, 1066

U. 5. Pat. 3,243,280, Mar. 29, 1966

U.S. Pat. 3,262,756, July 26, 1966

U.S. Pat. 3,273,973, Sept. 20, 1965
Papers Presented at Scientific and Technical Meetings

AUTHOR(s)

Ackley, R. D, R. E.
Adams, and W. E.

Browning, Jr.

Adams, R. E., J. S. Gil},
W. D. Yuille, W. E.
Browning, Jr., L. F.
Parsly, and C. E. Guthrie
Apple, R. F., J. M. Dale,
F. L. Whiting, A. S.
Meyer, and C. F. Baes, ]Jr.

Bennett, R. L., R. K.
Adams, and W. E. Browning,
Jr.

Bennett, R, L., R. E. Adams,
and W. E. Browning, Jr.

Bien, P. B.

Blankenship, F. F.

Blood, C. M., and
I.. G. Overholser

Bopp, C. D., and W. W.

Parkinson, Jr.

Brunton, G. D.

Cantor, S,

TITLE

Removal of Radioactive Methyl Iodide from

Steam-Air Systems

Performance of Filter Systems Under Ac-

cident Conditions

Determination of Oxide in Highly Radio-

active Fused Fluoride Salts

Characterization of Volatile Iodine Forms

Characterization of Volatile Forms of
Radiciodine Under High Humidity

Conditions

The Corresponding States of Aqueous Salt

Solutions

Chemical Separations in Molten Fluorides

Compatibility of Pyrolytic-Carbon Coated
Fuel Particles with Water Vapor

Radiation-Induced Codimerization of

Ether-Unsaturate Mixtures

The Crystal Structure of LiUFs

Predicting Density, Specific Heat and
Thermal Conductivity of Fluoride Melts

Constant Volume Heat Capacities of

Molten Salts

180

FLACE PRESENTED

9th AEC Air Cleaning Conference,
Boston, Mass., Sept. 1316, 1366

gth AEC Air Cleaning Conference,
Hoston, Mass., Sept. 1316, 1966 “

Analytical Chemical Society, Phoenix,
Ariz., Jan. 1621, 1966

Oth AEC Air Cleaning Conference,
Boston, Mass., Sept. 1316, 1966

American Chemical Society, New
York, Sept. 1116, 1666

American Chemical Society, New
York, Sept. 11--16, 1966

2nd Intern. Thorium Fuel Cycle
Symposium, Gatlinburg, Tenn., -
May 3—6, 1966

Libby-Cockcroft Graphite Chem,
Meeting, Harwell, England, Apr.
25--27, 1966 =

6th Annual Contractors Meeting,
Process Radiation Development
Program, Washington, D. C,,
Sept. 2627, 1966

American Crystallographic Assoc,,
Austin, Tex., Feb. 28—Mar. 2, 1966

American Chemical Society, New
York, Sept. 1116, 1966

American Chemical Society, SW
Regional Meeting, Albuguerque,
N. Mex., Nov. 3(¢--Dec. 2, 1966
AUTHOR(s)

Carroll, R. M., and

. Sisman

Coobs, J. H., and
J- G. Morgan

Field, P. E., and
J. H. Shaffer

Grimes, W. R.

Jenks, G. H.

Keilholtz, G. W., R. E.
Moore, M. F. Csbormne,
B. W. Wieland, and A. F.
Zulliger

Keilholtz, G. W., J. E. Lee,

Jr., and R. E. Moore

Keilholtz, G. W., R. E,

Moore, and M. F. Osbome

Marshall, W. L.

Marshall, W. 1.., and Ruth

Hlusher

Parker, G. W., R. A, Lorenz,
and J. G. Withelm

181

TITLE

In-Pile Properties of Reactor Fuels by

Oscillating Techniques

Coated Particle Fuels Development at
Oak Ridge Wational Laboratory

The Solubilities of Hydrogen Fluoride and

Deuterium Fluoride in Molten Fluorides

Molten #luorides as Nuclear Reactor

Fuels

Molten Fluorides as Fuels and Coolants

in the Molten Salt Reactor Experiment

Prediciion of Radiation Effects on Reactor

Water and Solutions

Technigues {or [rradiating High Tempera-

ture Materials in a Steep Flux Gradient

Properties of Magnesium, Alumioum, and
Beryllium Oxide Compacts Irradiated to
Fast-Neutron Doses Greater than 1021
neutrons/cm2 at 150, 800, and 1100°C

Properties of the Refractory Metal
Carbides of Titanium, Zirconium,
Tantalum, Niobium, and Tungsten
Irradisted to Fast Neutron Doses Greaier
than 1041 neutr.f)ns/(:m2 (Greater than 1
Mev)

Aqueous Solutions at High Temperatures

and Pressures

Thermosdynamics of Gypsum in Agueous

Sodium Chloride Solutions, Dw110°C

Solubilities of Calcium Sulfate in Sea
Salt Solutions (o QOODC —

Concentration Limits for Saline Water in

Temperature

General

Chemical Factors Affecting the Distribution
of Fission Products from o, Fuel
Melted Under Water PDuring Transient

Accidentis

PLACE PRESENTED

Intemational Symposium on Capsule
Irradiation Experiments, Pleasanton,
Calif., May 3-5, 1966

11th AEC Coated-Particle Working
Group Meeting, Los Alamos, ‘
N. Mex., June 12, 1966

American Chemical Society, New
York, Sept. 1116, 1966

EUCHEM Conference on Molten Salt
Chemistry, Ulvik, Norway, May
10-13, 19566

American Chemical Society, New
York, Sept. 1116, 1966

American Nuclear Society, Pittsburgh,
Pa., Det 31~Nov. 4, 1966

Intemational . Symposiun on Capsule
Irradiation Experiments, Pleasanton,
Calif., May 35, 1966

Joint American Nuclear Bociety and
American Ceramic Society Meeting,
Washington, . C., May 9~-11, 1966

American Ceramic Society, Pacific
Coast Regional Meeting, Portland,
Ore., October 19586

Princeton University, Chemistry
Seminar, Feb. 10, 1966

New York University, Chemistry
Seminar, Feb. 11, 1966

Northwesterm University, Chemistry
Seminar, Feb. 24, 1966

American Chemical Society,
Pittshurgh, Pa., Mar. 2831, 1966

American Chemical Society, SW

Regional Meeting, Albuquergue,
N. Mex., Nov. 30-Dec. 2, 1966

American Chemical Society, New
York, Sept. 1116, 1966
AUTHORS(s)

Parker, G. W.

Parker, G. W., G. E.

Creek, and A. Ferreli

Parkinson, W. W., Jr., and
W. C. Sears

Quist, A. S., and W. I..
Marshall

Quist, A. S.

Reagan, P. E., J. G. Morgan,

and O. Sisman

Romberger, K. A., C. F.
Baes, and H. H. Stone

Rutherford, J. L., J. P.
Blakely, and L. G.

QOverholser

Savage, H. C., J. M. Baker,
M. J. Kelly, and E. L.

Compere

Sears, D. R., and J. H.

Bums

Secoy, C. H.

Silverman, M. D.,
J- Truitt, R. E. Adams,

and L[.. F. Franzen

Strehlow, R. A.

Sweeton, F. H.,, C. F.
Baes, and R. W. Ray

Thoma, R, E.

182

TITLE

Chemical Factors in the Removal of Radio-

iodine by Reactor Filter Systems

Retention of Methyl Iodide by Impregnated

Carbons Under Ambient Conditions

The Effect of Radiation on the Olefinic

Groups in Polybutadiene

Electrical Conductances of Agueous
Sodium Chloride Solutions to 800°C and
4000 Bars

Electrical Conductances of Aqueous Solu-
tions to 800°C and 4000 Atmospheres

Performance of Pyrolytic Carbon Coated
Uranium Oxide Particles During Irradia-
tion at High Temperatures

Phase Equilibrium Studies in the UO2-
Zr02 System

Oxidation of Unfueled and Fueled Graphite
Spheres by Steam

An Assembly for Irradiation of Molten
Fluoride Fuel to High Bumups in the
Oak Ridge Research Reactor

The Crystal Structure of Beta-l KLHF4

Electro-Kinetic Transport at Phase

Boundaries

Adsorption of Water on Refractory Oxides

Characterization of Radicactive Particu-
late Aeroscls by the Fibrous Filter
Analyzer

Lithium Fluocoride Heats of Sublimation

The Solubility of Ferrosoferric Oxide in
Aqueous Solutions at Elevated Tempera-

tures

Corrosion Behavior of the MSRE in Zero-

and L.ow<Power Tests

PLACE PRESENTED

American Chemical Society, New
York, Sept. 1116, 1966

9th AEC Air Cleaning Conference,
Boston, Mass., Sept. 14--16, 1966

American Chemical Society,
Pittsburgh, Pa., Mar, 22--31, 1966

American Chemical Society, New
York, Sept. 1116, 1966

University of Nebraska, Dept, of
Chemistry Seminar, Linceln, Neb.,
Dec. 14, 1966

American Nuclear Society, Denver,

Colo., June 20--23, 1966

American Chemical Society,
Pittsburgh, Pa., Mar. 22--31, 1966

[.ibby-Cockcroft Graphite Chemistry
Meeting, Harwell, England, Apr.
25.--27, 1966

Inteimational Symposium on Capsule
Irradiation Experiments, Castlewocod
Country Club, Pleasanton, Calif.,
May 3-5, 1966

American Crystallographic Association,
Austin, Tex., Feb. 28—Mar. 2, 1966

ORAU Traveling Lecture Program,
University of Toledo, Mar. 9, 1966

ORAU Traveling Lecture Program,
Hendrix College, Conway, Ark,.,
Apr. 15, 1966

Oth AEC Air Cleaning Conference,
Boston, Mass., Sept. 1416, 1966

Bendix Time-of-Flight Symposium,
Cincinnati, Ohio, October 1966
American Chemical Society, SE

Regional Meeting, Iouisville, Ky.,
Oct. 27-29, 1966

15th AEC Corrosion Symposium, Qak
Ridge, Tenn., May 23--25, 1966
AUTHOR(s)

Thoma, R. E.

Thoma, R. E., B. G. Ross,
and C. F. Weaver

Thoma, R. E., H. Insley,
H. A. Friedman, and
G. M. Hebert

Watson, G. M., R. B.
Parez, and M, H. Fontana

Waison, G. M.

183

TITLE

Molten Salt Reactor Technology

Production of Lithium Fluoride Crystals
with Selected Isotopic Ratios of
Lithium

The Lithium--Beryllium—Zirconium Flueride

System in Molten Salt Reactor Technology
Effects of Containment System Size on
Fission Product Behavior
Defect=Trap Theory of Fission Gas

Release

Recoil Range of Fission Fragments in

Graphitic Systems

Gas-Graphite Reactions

Carbon Deposition Studies

Migration of Actinides in Graphite and

Pyrocarbons

Effects of Containment Size on Fission

Product Behavior

PLACE PRESENTED

ORAU Traveling Lecture Program,
University of Texas, Nov., 7, 1966

ORAU Traveling Lecture Program,
Notth Texss State College, Nov. 9,
1966

ORAU Traveling Lecture Program,
Texas Technological University,
Nov. 10, 1966

Intetnational Conference on Crystal
Growth, Boston, Mass., June 20.-24,
1966

American Chemical Society,

New York, Sept. 11-16, 1966

Oth AEC Air Cleaning Conference,
Boston, Mass., Sept. 13—-16, 1966

Reactor Chemistry Information Ex-
change Meeting, General Atomic,
San Diego, Calif., Nov, 8~9, 1966

Reactor Chemistry Information Ex-
change Meeting, General Atomic,
San Diego, Calif., Nov. 8-9, 1966

Reactor Chemistry Information Ex-
change Meeting, General Atemic,
San Diego, Calif., Nov. 8~9, 1966

Reactor Chemistry Information Ex-
change Meeting, General Atomic,
San Diego, Calif., Nov. 8--9, 1966

Reacter Chemistry Information Ex-
change Meeting, General Atomic,
San Diego, Calif., Nov. 8--9, 19606

Reactor Chemistry Information Ex-
change Meeting, General Atomic,
San Diego, Calif., Nov, 8—9, 1966
2~

6~

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185

ORNL-4076
UC-4 — Chemistry

INTERNAL. DISTRIBUTION

Biology Library

Central Research Library

Laboratory Shift Supervisor

ORNL Y-12 Technical Library
(Document Reference Section)

LLaboratory Records Department

l.aboratory Records, ORNL. R.C.

. Larson

. Weinberg

. MacPherson
. Boyd

. Bruce

. Culler

. Jordon

. Snell

TIT Aamo=Em

. Briggs

. Cottrell

. Fraas

. Kraus

. Lane

. Miller

. Rosenthal

. Trauger

. Whitman

. Beall

. Billington

. Ferguson
Frye, Jr.

. Kelley

. Taylor

. Bettis

. Bredig

. Corbin

. Cunningham
. Crawford

. Haubenreich
. Kasten

. Kastenbaum
. King

. Lyon

. Parker

. J. Skinner

ZAAMZOVVCCIIMMAZTCT00OIZIPCR>EAPIP>O>ETTOL >0
WZEPAZIMAPAIATIMMOTEC > P> TE®EC

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159.

. White

. Williams

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. Bohlmann

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. Baker

. Bamberger
Barton

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. Bennett

. Bien

. Blood

. Bopp

raunstein

E-O0O0O0TAAON0OC-SOP>OD N TOIME O

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. L
. B
. M
. D

B
. E

. Browning, Jr.
G. D. Brunton
H. Buchholz

S. Cantor

R. M. Carroll
Zell Combs

. Compere

. Creek

. Cuneo
Davis

. Doss

. English

. Evans, I
. Fairchild
. Fontana

. Fried

. Friedman
. Fuller, Jr.
Gadiyar
Gill

. Gilpatrick
. Griess, Jr.
. Hebert

Hess

VoOErr-ImIVIrA-To060Mm
ZZTOAOYVr>PITrWr>"0mr
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O_L—'UIern(‘):Dm'fl§>0msfiz’£®1‘loi——%§ommm1_@rfi

Holmes

. Jenks
. Keilholtz

Kelly

. Keyser

Kirslis

. lorenz
. Malinauskas

. Marshall, Jr.

Martin

. Mauney

. Mesmer

. Miller, Jr.
. Moore

. Morgan

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Myron

. Neumann

. Osborne

. Overholser
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Quist

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Shor

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186

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234,

0. Sisman

Ruth Slusher

B. A. Soldano

H. H. Stone

R. A. Strehlow

B. J. Sturm

F. H. Sweeton

R. E. Thoima, Jr.

J. Truitt

W. T. Ward

G. C. Warlick

C. F. Weaver

J. F. Winesette

L. B. Yeatis

W. D. Yuille

Leo Brewer (consultant)
J. W. Cobble (consultant)
R. W. Dayton {consultant)
P. H. Emmett (consultant)
H. S. Frank (consultant)
N. Hackerman (consultant)
D. G. Hill (consultant)

H. Insley (consultant)

E. V. Jones (consultant)
T. N. McYay (consultant)
G. Mamantov (consultant)
J. L. Margrave (consultant)
E. A. Mason (consultant)
R. F. Newton {consultant)
R. B. Perez (consultant)
J. E. Ricci (consultant)
Howard Reiss (consultant)

G. Scatchard (consultant)
D. A. Shirley (consultant)
H. Steinfink (consultant)
R. C. Vogel {consultant

T. F. Young (consultant)

EXTERNAL DISTRIBUTION

D. F. Bunch, Health Physics Branch, AEC, Washington
Paul E. Field, Dpt. of Chemistry, Virginia Polytechnic Institute

Research and Development Div., AEC, QRO
Reactor Div., AEC, ORO

Asst. General Manager for Research and Development, AEC, Washingten
Division of Research, AEC, Washington
Division of Isotopes Development, Washington
Asst. General Manager for Reactors, AEC, Washington
243.
244,
245.
246.
247.

248-510.

s

187 )

T W

Division of Reactor Development and Technology, AEC, Washington
Space Nuclear Propulsion Office, AEC, Washington

J. A. Swartout, 270 Park Ave., New York, N.Y.

Milton Shaw, AEC, Washington

W. W. Grigorieff, Assistant to the Executive Director, Oak Ridge
Associated Universities

Given distribution as shown in TID-4500 under Chemistry category
(25 copies ~ CF3TI)