: leil'ifli- ADEENT *J'J.i':;felfl"c i‘fi_h\i BITHYAELT S | i 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. 3 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. iii U iv 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.) 11 13 14 15 17 18 19 22 24 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~ 26 26 27 32 34 36 38 vi 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. 39 41 45 46 48 49 50 51 53 Vil 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. viii 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. 70 72 74 74 75 77 78 83 84 ix 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 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 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 £00 ; T . © | et | | | | 2 - e e - Ty T T ‘ % o | e e | I e | ! A i € 600 |— Wl e - 1= ) | ~ !E | i ; o o~ | e N -~ | ! w | \ | | 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/ -« ° 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 w0 bl bbb b 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. 1) =z =T & s o = o 2l > E oo i . ] uZJ‘iJJ - - = Z ! ?E:IJ € 5 5 =2 T e - o Il.l L fi =z 0. O > I éo 100 & e Ll Eow [ ¥ B . Bo o = Zw g = il PR | - O W "._) < boao — Gura WL 2 by e - = ra | I < a4 2 a2 a. 8—) O K e o 6O - —--———':LJE'-!;{;--—S- o ?-qr- = & O - z o] £z L o ‘so}/ oo — - W0 2 Loger @] = e} - - . 1 o 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 — < © ; 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 OFF-GAS HV-B08 (METERING VALVE ). HV-902 - PS-801 CONTAINMENT HV-806 Ly ey ®) s N VESSEL., \ V7 —r]q.] ,,,,,,,,,,,,,,,,,,,,,,, N OUICK \ \‘ | HV-807 ( P . ' ! PS-802 ) DISCONNECTY 4 wr/ HV-BC3 v . ORRPOOL conTANMENT o A ’ \ SECONDARY j \J-eo! Y dadS ] THERMOCGUPLE S AN— =g B Y ACCESS : R e NS 801 ( T il Ll S - o FERRULE SEAL-._ j i -— == jL‘—__ — — e FILTER COMTAIN- ------ MENT EVACUATION | TUBE e ‘ JUNCTION \TUBE______,,, - | GAS | ‘ HY NORMAL ¢ NEEDLE VALVE POOL WALL =~ LIQUID - NITROGEN- COOLED CHARCOAL TRAP QUTSIOE EOTTLE RACK - OFF-GAS HOOD — . FLEXIBLE BELLOWS HOSE- -, ‘ TN | ffil{fm_f_ ~Lem S Y A N 4 "Q"'"""W‘W“’WML ‘:fi\ DIFFUSION ~ o - | \ HV—BOSN GAS INLET | OFF-GAS - % 80X SPRAY NOZZLE FOR WASH SOLUTION Yo | WASH SOLUTIGN } RECOVERY TUBE—‘—__, 7‘7* \ 'CONTAINMENT VESSEL “CONTAINMENT SECONDARY PRESSURE REGULATOR VALVE PRESSURE INDICATOR PRESSURE SWITCH DIFFERENTIAL PRESSURE GAGE FLOW INDICATOR 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 - > ) 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 ScaleEAPIP>O>ETTOL >0 WZEPAZIMAPAIATIMMOTEC > P> TE®EC 114. 115. 116. 117. 118. 119. 120. 121. 122. 123. 124. 125. 126. 127. 128. 129. 130. 131. 132. 133. 134, 135. 135. 137. 138. 139. 140. 141. 142. 143. 144. 145. 146. 147. 148. 149. 150. 151. 152. 153. 154. 155. 156. 157. 158. 159. . White . Williams . Grimes . Bohlmann . McDuffie . Watson . Blankenship . Secoy . Ackley . Adams . Bacarella . Baes . Baker . Baker . Bamberger Barton . Baumann . Bennett . Bien . Blood . Bopp raunstein E-O0O0O0TAAON0OC-SOP>OD N TOIME O . C .C . R .G . F . M . F . H .D . E . L . F E M . E - J .D . 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 160. 161. 162. 163. 164. 165. 166. 167. 168. 169. 170. 171. 172. 173. 174. 175. 176. 177. 178. 179. 180. 181. 182. 183. 184. 185. 186. 187. 188. 189. 190. 191. 192. 193. 194. 195. 196. 197. 235. 236. 237. 238. 239. 240. 241. 242. TP M- O-IATIPNOVE TP EOrIO-EOCINRAZZI> VO DT OO >OZ =TT 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 . Moulten . Morgan Myron . Neumann . Osborne . Overholser . Parker . Parkinson, Jr. Quist . Reagan . Redman . Reed . Richardson . Roberts . Robertson . Romberger . Savage . Savolainen . Sears . Shoffer . Shields Shor . Silverman 186 198. 199. 200. 201. 202. 203. 204. 205. 206. 207. 208, 209. 210. 211, 212. 213. 214. 215, 216. 217. 218. 219. 220. 221. 222. 223. 224, 225. 226. 227. 228. 225. 230. 231. 232. 233. 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)