LOCKHEED MARTIN ENEI AESEARCH LIBHARIES il 345 05159761 ORNL-3913 UC.4 — Chemistry TID-4500 (47th ed.) Contract No, W-7405-eng-26 REACTOR CHEMISTRY DIVISION ANNUAL PROGRESS REPORT For Period Ending December 31, 1965 Director W. R. Grimes Associate Directors E. G. Bohlmann H. F. McDuffie G. M. Watson Senior Scientific Advisors F. F. Blankenship C. H. Secoy MARCH 1966 OAK RIDGE NATIONAL LABORATORY Dak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the 1. 5. ATOMIC ENERGY COMMISSION LOGCKHEED MARTIN ENERGY RESEARCH LIBRAR UAADIRALT 1l 4 445k 0515976 1 i i [ i Contents PART |, MOLTEN-SALT REACTORS 1. Phase Equilibrium ond Crystallographic Studies LIQUID-LIQUID IMMISCIBILITY IN THE SYSTEM L:’F—Ber-ZrF4 H. A. Friedman and R. F. TROMG ..t e ee et bbb ey s b s aae e e e et nns 3 Composition~temperature limits of liquid-liquid immiscibility in the system Lil“~'BeF?—ZrF4 have been partially established by application of high-temperature centrifugation. S0DIUM FLUORIDE-SCANDIUM FLUORIDE PHASE EQUILIBRIA R. H. Karraker and R. E. Thoma...........c.oceiccevinann e e ettt ee e ah et eees e aeeeeenaeeians ceiei i ianeeraaei e nean s e e 4 Results of a completed investigation of the high-temperature reactions of Na¥ and S(:F3 show that the system is characterized by two eutectic and two peritectic reactions which occur in association with the intermediate phases 31‘~Ta3['?~-Sr:F3 and NaF-SCFs.- COMPOSITIONAL VARIABILITY IN SODIUM FLUCRIDE -LANTHANIDE TRIFLUORIDE COMPLEX COMPOUNDS R. E. Thoma, H. Insley, and G. M. Hebert i e e e er e e e e e e et e s e e et enemceen 6 Monotonic expansion of the compesition limits of the complex NaF»l’.,-nF3 crystalline phases has been ascribed te a reduction in the polarizability of the lanthanide ions with increasing atomic number accompanied by a cor- responding increase in free space within the crysial lattices, PHASE EQUILIBRIUM STUDIES IN THE U02~Zr02 SYSTEM K. A. Romberger, H. H. Stone, and C. F. Baes, Jr. i e 8 The U02-2r02 phase diagram has been revised in accord with the measured low solubilities of Z’rO2 in cubic UO2 and of UO,2 in tetragonal ZrO:! and monoclinic ZrOZ. These solubilities are, respectively, 0.4, 1, and 0,15 mole % at the eutectic temperature of 1116°C. THE CRYSTAL STRUCTURE OF LIUF5 G D BrifI O s et sarae s ar s e e b a e etaeeeMeeeseitieasereeiseesieaereteaeeseseemeeeaetieaeeaaneneiiaas 10 An x-ray structure analysis has ascertained the formula of this compound, previcusly designated 7LiF‘—6UF4. In it are found UE‘8 polyhedra which are quite similar to those in crystalline UF4. THE CRYSTAL STRUCTURE OF Li3A!F6 J- H. Burns and A. C. TenniSSBIl .o e e 12 - + This compound crystallizes with A!.F‘S3 octahedra joined together by 1L.i ions. REFINEMENT OF THE CRYSTAL STRUCTURE OF (NH jj MnF The previously determined structure of (NH4}2MnF5 was refined by least squares. HIGH-TEMPERATURE X-RAY STUDIES G. D. Brunton, D. R. Sears, and J. Hl. BUINS i e 16 The furnace attachment for the x-ray diffractometer has heen used to study phase transformations in rare-earth trifluorides, the U(')Q-'-Zr(i);Z system, and LisAIFG. i1l iv CRYSTALLOGRAPHIC DATA ON NEW COMPOUNDS J. H. Burns, D. R. Sears, and G. I BrURLON oot e et ettt ettt et et e e aae e 17 Unit-cell dimensions and space groups have been determined for Na3SCF6, BI-KLaF4, Rb}”aFé, Li4UF LiU4F17, and NaBiF4. THE CRYSTAL STRUCTURE OF I\e'chZréF_.’,r J. H. Burns, R. D. Ellison, and H. AL L evy o e e e e e e e e it e 17 Six formula weights of NaF and six of Z’rF4 comprise a structure which has fluorine-bridged zirconium octahedra containing a seventh fluorine atom within; a seventh sodium atom is also present in the structure and has 12 fluorine neighbors. 2. Chemical Studies of Molten Salts OXIDE CHEMISTRY OF LfF«BeFZ-ZrF4 MELTS C. F. Baes, Jr., and B. F. HitCh o e e e e e e e 20 Measurements of the solubility of ZrO2 in simulated MSRE fuel salt and flush salt mixtures are being continued by means of an improved technique of oxide analysis. THERMODYNAMICS OF MOLTEN L.iF-BeF2 SOLUTIONS O R C T Vet £ PRSP 21 Enthalpies and free energies of formations, electrode potentials, and activity coefficients are estimated for various solutes in 2Lil~'-13el-""_j3 from existing chemical and thermochemical data. VAPOR PRESSURES OF FLUORIDE MELTS S. Cantor, D. 8. Hsu, and W. T Ward e et et e e 24 Rodebush-Dixon and boiling-point techniques were used to measure vapor pressures of the LiF-BeF2 system, the MSRE fuel solvent, and Ber. A transpiration apparatus was constructed and tested with pure LiF. VISCOSITIES OF MOLTEN FLUORIDES B Cantor and W. L W ard oo e e e e s e ey e et e e et i 27 Studies on the LiF-BeF2 system were extended. The compound NaBF4 was found to be fluid. ESTIMATING DENSITIES OF MOLTEN FLUORIDE MIXTURES Bt I O L st it et e e e e eee et et e eeets e ereeetaee et eeea e e ee et ee e ettt eee bt e en et e et ae e e e enianeras 27 Examination of newer data of molar volumes resulted in revised values for use in estimating densities of molten fluorides., ESTIMATING SPECIFIC HEATS AND THERMAL CONDUCTIVITIES OF FUSED FLUORIDES B A O i e it it ittt e e e e et e e e e e e e e et e e e et ee s e oo e e —eeeeee e e eea e eeeneeeneeetnnn et et et et a et a e e 29 By modifying the Dulong-Petit rule to equal 8 cal (OK)'"1 (gTam-atom)_l, recasonable agreement with experi- mental data was obtained. A semitheoretical method used to estimate thermal conductivities agreed with experi- mental results; the method depends on estimating sonic velocity, which itself can be calculated from thermo- dynamic parameters. SOLUBILITY OF DF AND HF IN LiFvBeF2 (66-34 MOLE %) P. E. Field and J. H. SRaffer e e e e e 33 Solubilities of DF and HF were determined for the temperature interval 500—700°C to provide a basis for estimating the solubility of tritium fluoride in the blanket of a thermonuclear breeder reactor. 3. Chemical Separatien and Irradiatien Behavior IN-PILE MOLTEN-SALT IRRADIATION ASSEMBLY H. C. Savage, M. J. Kelly, E. L.. Compere, J. M. Baker, and E. G. Bohlmann ..., 34 An in-pile molten-salt irradiation experiment is being constructed for operation in bearm hole HN-1 of the ORR. EVAPORATIVE-DISTILLATION STUDIES ON MOLTEN-SALT FUEL COMPONENTS M o BBLLY oottt et e e et e e e 35 Mass-~rate variations with temperature for distillation of "LiF and MSRE fuel solvent and the volatility of NdF | relative to that of MSRE fuel solvent have been determined experimentally. EFFECTIVE ACTIVITY COEFFICIENTS BY EVAPORATIVE DISTILLATION OF MOLTEN SALTS M. J. Kelly Activity coefficients of BeF2 and ZrF4 have been obtained from data on vacuum distillation of these com- pounds from molten LAF solvent. REMOVAL OF IODIDE FROM LiF-BeFZ MELTS B. F. Freasier, C. F. Baes, Jr., and H. H. Btone .. e e e e, e 38 Tedide was readily removed from molten LiF-Bel", mixtures by HF-H sparging; this process should prove a valuable means of removing the precursor of 135ye (1351) from molfen—fluondc reactor fuels, REMOVAL OF RARE EARTHS FROM MOLTEN FLUORIDES BY EXTRACTION INTO MOLTEN METALS J. H. Shaffer, F. A. Doss, W. K. R. Finnell, W. P. Teichert, and W. R. Grimes. . ..., 40 Rare-carth fluorides have been removed from solution in a fluoride mixture which simulates the barren fuel solvent for a molten-salt breeder reactor by reduction and extraction into molten bismuth. REMOVAL OF PROTACTINIUM FROM MOLTEN FLUORIDES BY OXIDE PRECIPITATION J. H. Shaffer, F. A. Dossg, W. K. R. Finnell, W. P, Teichert, and W. R. Grimes ... ........cooiuiiiimmimiiineeaan 41 Previous studies were extended to demonstrate that protactinium could be precipitated as its oxide by the addition of ZrO? to a fluoride solvent known to have Zr()2 as a stable oxide phase. REMOVAL OF PROTACTINIUM FROM MOLTEN FLUORIDES BY REDUCTION PROCESSES J. H. Shaffer, F. A. Doss, W. K. R. Finnell, W. P. Teichert, and W. R. Grimes In the reference-design MSBR, protactinium can be removed from solution in the proposed fluoride blanket mixture by reduction with thorium or with alloys of thorium in lead or bismuth. SOLUBILITY OF THORIUM IN MOLTEN LLEAD J- H. Shaffer, F. A. Doss, W. K. R. Finnell, and W. P, T eiChert et e, 43 The solubility of thorium in lead over the temperature interval 400-600°C was determined in order to support studies of the extraction of protactinium from molten fluorides. PROTACTINIUM STUDIES IN THE HiGH-ALPHA MOLTEN-SALT LABORATORY e e BB O it iitie it et L L b s L bttt et e eot £k e e e f e e £at te ettt et e eun e esfn e et tn ettt et tneettn e stntntarnsrtnarrrn e bares 44 Seven glove boxes have been interconnected and equipped for studies of protactinium recovery from molten- fluoride breeder-blanket mixtures. SEGREGATION ON FREEZING LiCI-KC] EUTECTIC MELTS CONTAINING SOLUBLE SOLUTES H. A. Friedman and F. F. Blankenship e e e et te et s e et et e s ettt e e e n e e e e e eanennns 45 The feasibility of purifying salt melts by freezing with rapid stirring to facilitate diffusion of rejected solute from the freezing front was explored. 4. Direct Support for MSRE PREPARATION AND LOADING OF MSRE FLUORIDES J. H. Shaffer, W. K. R, Finnell, F. A. Doss, and W. P. Teichert ..o e 47 The production of all the component fluoride mixtures for the MSRE fuel was completed; these mixtures were added to the MSRE as required during the precritical and critical testing phases of MSRE operation. CHEMICAL BEHAVIOR OF FLUORIDES DURING MSRE OPERATION Results of chemical tests with MSRE fuel and coolant salts, obtained during the precritical test, the zero- power test, and the initial period of the full-power test, indicated that the reactor salts are of excellent purity and that no appreciable corrosion of the interior of the reactor has occurred. vi MEASUREMENT OF DENSITIES OF MOLTEN SALTS B. J. Sturm and R. E, TROMA i i e e e e e e 50 Density of the MSRE fuel salt was determined with molten mixtures using an electrical probe to measure volume of the salt. The fuel-salt density (in g/cm”) is described by the expression d = 2.848 — 7.693 x 10~ " ¢ (°C). PART 1l. AQUEOUS REACTORS 5. Corrasion and Chemica! Behavior in Reactor Environments MECHANISM OF ANODIC FILM GROWTH ON ZIRCONIUM AT ELEVATED TEMPERATURES Al L. Bacarella and Al L SUEEON et e st e ekt e a e e e et e r b 55 Considerations of experimental data for zirconium oxidation and of theory provided evidence that the pre- exponential term in the anodic-film-growth equation can be interpreted in terms of paraineters of fundamental sig- nificance. MECHANISM OF RADIATION CORROSION OF ZIRCONIUM AND ZIRCALOY-2 R. J. Davis and G. H. Jems o e ettt e e et e et 57 The previously reported radiation effect on the postirradiation corrosion was confirmed, and exploratory work toward new methods of evaluating film protective properties is in progress. EFFECTS OF REACTOR OPERATION ON HFIR COOLANT .................................................................................................................................................................. 58 Final evaluations were completed of the probable concentrations of excess oxidant needed to stabilize HNO3 in the HFIR coolant-moderator and of the expected steady-state concentration of the decomposition products of water. NASA TUNGSTEN REACTOR RADIATION CHEMISTRY STUDIES G. H. Jenks, H. C. Savage, and E, G. BohImanm ... .o e e et et e e 58 Designs, equipment, and procedures are being developed for experiments to test the effects of irradiation on loss of cadmium from CdSO4 solution under conditions of interest in the NASA Tungsten Water-Moderated Reactor. CORROSION SUPPORT FOR VARIOUS PROJECTS J. C. Griess, J. L. English, L. 1.. Fairchild, and P. D. Neumanm .. .................ccccooiiiiiriiioiiieie i 60 Corrosion studies were conducted on materials to be used in the High Flux Isotope, Advanced Test, and Argonne Advanced Research reactors; some studies of Hastelloy N and nickel! for use in gas-phase fluidized+bed fuel-element process equipment were also carried out. 6. Chemistry of High-Temperature Aqueous Solutions ELECTRICAL CONDUCTANCE MEASUREMENTS OF AQUEQOUS SODIUM CHLORIDE SOLUTIONS TO 800°C AND 4000 BARS A. S. Quist, W. Jennings, Jr., and W. L. Marshall e et 63 The electrical conductances of aqueous sodium chloride sclutions from 0.001 tc 0.1 m were measured at tem- peratures from 100 to 800°C; sodium chloride solutions still behaved as moderately strong electrolytes at high temperatures and densities, AQUEOUS SOLUBILITY OF MAGNETITE AT ELEVATED TEMPERATURES F. H. Sweeton, R. W. Ray, and C. F. Baes, Jr. i e ittt ter e et tv et a e tee e et 64 The solubility of Feao4 is being studied as a function of pH at temperatures between 150 and 260°C. yii SOLUBILITIES OF CALCIUM HYDROXIDE AND SATURATION BEHAVIOR OF CALCIUM HYDROXIDE-CALCIUM CARBONATE MIXTURES IN AQUEOUS SODIUM NITRATE SOLUTIONS FROM 0.5 TO 350°C L. B. Yeatts, Jr., and W. L. Marshall ..., et et e e a e nn et aaen e e et n e e e e e 65 The solubilities of calcium hydroxide and of calcium hydroxide—calcium carbonate mixtures in water and aqueous sodium nitrate solutions were determined and used hoth to test further the applicability of an extended Debye-Hickel equatlion and to obtain thermodynamic functions. 7. interaction of Water with Particulate Solids SURFACE CHEMISTRY OF THORIA C. H. Secoy Heats of Immersion and Adsorption H. F. Holmes, E. L. Fuller, Jr., and J. E. SHUCKEY .ottt e e e 68 Net differential heats of adsorption for several high- and low=-surface-area thoria samples have been calculated from calorimetrically determined hezats of immersion. Water Vapor Adsorption and Desorption E. L. Fuller, Jr., and H. Fu HoOe S, i et ir e e e et r e et ie et e e e e rr e e eart b eaeeeereeetanrterareneees 69 Careful determination of adsorption and desorption isotherms for water vapor on thoria has revealed the complex details of both.the chemisorption and physical adsorption processes, Infrared Spectra of Adscrbed Species on Thoria Co 80 BShoup, Jre oo e e ettt teeanwaseee e et eeiiEedieseeemeeeeeraseaeaseeeetsaeaeeeattensean e aaenetn e enain e 70 The infrared spectra of a thin thorium oxide disk as functions of pretreatment conditions substantiate the com-~ plex nature of the chemical water-adsorption process, Elvctrokinetic Phenomena at the Thorium Oxide~Aguecus Solution Interface C. 8. Shoup, JIr., and H. F. HOIMES ..o oo e e 71 An electrical study of the interface between thoria and agueous solutions of several electrolvtes has disclosed surface conductances two or three orders of magnitude greater than those predicted by classical theory. GAS EVOLUTION FROM SOL-GEL URANIUM-THORIUM OXIDE FUELS D, N. Hess and B. A, Soldano........ccoiiiiiiii ettt A e eeet e eeen e eLhee e e eeentaeeeeee e e e e reinrae s aaeeneat s 72 The gas-adsorption capacity of thorium sol-gel is shown to be reduced by exposure to heated mixtures of COZ and HZO’ PART . GAS-COOLED REACTORS 8. Diffusion Processes TRANSPORT PROPERTIES OF GASES Thermal Transpiretion. Rotational Relaxation Numbers for Nitrogen and Carbon Dioxide A I, M A A UE KA it ettt et ettt e e e eeee ettt e e et eeeeeeeeen te ettt eeeeaseeeet A teeaeteniaaea e ea it tae i eataen e teearen e raraa e ehe s 77 The analysis of thermal transpiration data in accordance with the dusty-gas model appears teo provide a simple method for the investigation of inelastic collisions. Goseous Diffusion in Noble Gas Systems A . M AU S A S ittt i et ttieetie ertit et e et et eee et e e eeet ettt ea et teaeentt ettt tetneer At e e nrar e te e eeentnrn e ee e enaanne s - 77 Preliminary analysis of diffusion data of the systems He-Kr, Ar-Kr, and Xe«Kr reaffirms the feasibility of em- vloying viscosity measurements to derive diffusion coefficients at high temperatures. Gaseous Diffusion in Porous Media A. P. Malinauskas, F. A. Mason, and R. B, Evans HI. ... et e et e et h e ettt anee e iieee e eataeeaaaneareas 78 The additivity of diffugsive and viscous fluxes has been found to be theoretically justified; as a result, the previous treatment of gas transport in the presence of a pressure gradient has been improved. viii SOLID-STATE TRANSPORT PROCESSES IN GRAPHITIC SYSTEMS Recoil Phenomena R. B. Evans 1III, J. L. Rutherford, and R. B. Perez. ... e 79 The recoil range for light fission fragments in General Electric pyrolytic carbon is greater than the range for the heavy fragments (13 (land11 (), and the straggling factor to range value, /R, is 0.126. Actinide Diffusion R. B. Evans III, J. L. Rutherford, and F. L. Carlsen, Jr. ... e 80 Actinide diffusion at low concentrations in General Electric pyrocarbon was found not to be concentration- dependent as in experiments with High Temperature Materials pyrocarbon, but preliminary constant-potential data indicate high coefficients and low activation energies. Self-Diffusion R S T - R o £ 1) L U ORI 81 Equipment and technigues are being developed to study the diffusion of 14¢ in pyvrolytic carbens and various graphites. 9. Reactions of Reactor Components with Oxidizing Gases I.. G. Overholser REACTIVITY OF ATJ GRAPHITE WITH LOW CONCENTRATIONS OF OXIDIZING AND REDUCING GASES P S =1 =033 O OO 83 Experimental studies were made of the reactivity of AT]J graphite with low concentrations of water vapor, carbon dioxide, methane, and carbon monoxide. COMPATIBILITY OF PYROLYTIC-CARBON-COATED FUEL PARTICLES WITH WATER VAPOR e M. Blood o ettt e ettt en et aa e e 85 Reaction-rate, coating-failure, and surface-area data were obtained for seven batches of pyrolytic-carbon- coated fuel particles during and after exposure at 1000°C to partial pressures of water vapor of 4.5, 45, and 570 torrs. COMPATIBILITY OF METALS WITH LOW CONCENTRATIONS OF CARBON MONOXIDE Jo B BaK T e e e U 86 The effect of carbon monoxide at 250 to 300 ppm in He on molybdenum, gold-plated stainless steel, and mild steel at 450 to 850°C has been evaluated. 10. lrradiation Behavior of High-Temperature Fuel Material Oscar Sisman and J. G. Morgan FISSION-GAS RELEASE FROM PYROLYTIC-CARBON-COATED FUEL PARTICLES P. E. Reagan, J. W. Gooch, J. G. Morgan, T. W. Fulton, and C. D. Baumann...................co, 88 Fission-gas release rates from pyrolytic-carbon-coated UO2 fuel particles were low even at elevated tem- peratures after high burnups. POSTIRRADIATION EXAMINATION OF COATED FUEL PARTICLES P. E. Reagan and E. L. L ong, Tl i e e e e e e e e 90 Several pyrolytic-carbon-coated UO2 particles have shown very little radiation damage at temperatures up t{o 1600°C and burnups to 25 at, % of the heavy metal. POSTIRRADIATION TESTING OF COATED FUEL PARTICLES M. T. Morgan, R. L.. Towns, and C. D. BaUmMaNTI ..o e e et e a e 94 Metallic fission product release studies of pyrolytic-carbon-coated fuel particles show possible effects of burnup, coating density, and fuel composition on the fission product release rates. ix POSTIRRADIATION EXAMINATION OF FUEL.ED GRAPHITE SPHERES D. R. Cuneo, J. G. Morgan, H., E. Robertson, C. D. Baumann, and E. L. Long, Jr. .....nnni. 95 Postirradiation examination of AVR-type spheres fueled with coated particles led to the conclusion that, after 10% burnup, the spheres had retained acceptable structural integrity. POSTIRRADIATION EXAMINATIOM OF EGCR FUEL ELLEMENT PROTOTYPE CAPSULES M. F. Osborne, E. L. Long, Jr., H. E. Robertson, and J. G. MOFZAam ... ..ottt et a e eeeneas a8 An EGCR prototype capsule that contained productien~-run UO2 pellets was examined after irradiation fo a purnup of 10,000 Mwd/metric ton UOQ, and we found no major detrimental changes. 11. Fission-Gas Release During Fissioning of UO, R. M. Carroll, Oscar Sisman, T. W. Fulton, R. B. Perez, and G. M. Watson EXPERIMENTAL .. e e e e eeer e aas 99 The medified in-pile assembly has been used te study fission-gas release rates while the fission rate (at constant temperature) or the temperature {at constant fission rate) was varied sinusoidally in a carefully contrelled wWay. A T HE M AT AL MO D E L e it ittt tttaaaae s ee i aaee e trtrs neeaaaaae eesastatsterarnesea e s e eenaaraanessnsrebes sanaeaaaanes 100 The mathematical model of a defect-trapping theory of fission-gas behavior has been refined and tested in a preliminary way with our experimental data. 12. Miscellaneous Studies for Solid-Fueled Reactors EQUILIBRIUM STUDIES IN THE SYSTEM ThOZ-UOZ-Uflg Lo O. Gilpatrick and Cu He SOy i i e e b 103 Althcugh the general features of the phase diagram for this system in equilibrium with air from 700 to 1600°C tad been reported previously, several details have begen either confirmed or slightly revised by sdditional experi- ments with improved techniques. BEHAVIOR OF REFRACTORY-METAL CARBIDES UNDER IRRADIATION . W. Keilholtz, R. . Moore, and M. F. Osborne o i s e s 104 The effects of fast neutrons on monocarbides of Ti, Zr, Nb, Ta, and W are being investigated at temperatures from 100 to 1400°C in order to evaluate their potential use in nuclear systems requiring extremely high power densities, EFFECTS OF FAST-MEUTRON IRRADIATION ON OXIDES G. W. Keitholtz, R, E. Moore, and M. F. Os8bBOME e T 105 Trradiation effects on sintered MgQ, AIEO%, and BeQ were investigated over the temperature interval 100~ 1100°C and the fast-neutron dose range of 0.2 to 5.1 X 104! peutrons/cm’ in srder te determine realistic conditions for uze of these oxides in nuclear systewms and to establish mechavisms of neulron damage. ANNEALIMG OF IRRADIATION-INDUCED THERMAL CONDUCTIVITY CHANGES OF CERAMICS C. D. Boppoooooeeie e e e e 109 The annealing of the neutron-induced thermal conductivity change was measured in nine ceramic materials after an irradiation dose uwp to 2 ¥ 1017 fast neutrons/cmz. PART IV. OTHER ORNL PROGRAMS 13. Chemical Support for Saline Water Program THERMODYNAMICS OF GYPSUM IN AQUEOQUS SODIUM CHLORIDE SOLUTIONS W. L. Marshall and Ruth SIuS her e et et e ettt e e e s 113 From an extensive study of the phase behavior of gypsum in aqueous sodium chloride and of those solutions cosaturated with sodium chloride or a double salt, NaZSO4-5CaSO4-3H20, thermodynamic functions were calculated both for a standard state and at high ionic strengths. THE OSMOTIC BEHAVIOR OF SIMULLATED SEA-SALT SOLUTIONS AT 123°C P. B. Bien and B. A. Soldano .. ... e e e 115 Varying the nature of the multivalent ionic components of seawater gave rise to significant changes in the osmotic behavior of saline solutions at elevated temperature. ALUMINUM- AND TITANIUM-ALLQY CORROSION IN SALINE WATERS AT ELEVATED TEMPERATURES E. G. Bohlmann, J. C. Griess, F. A. Posey, and J. F. Winesette e 117 Continuing studies of the corresion of aluminum and titanium alloys in saline water have demonstrated the superiority of 5454 aluminum and substantial inverse temperature dependence of the titanium pitting potential. CHEMISTRY OF SCALE CONTROL E. L. Compere and J. E. Savolaiflen . 119 Considerations of chemical factors related to the economic control of scale deposition in sSeawater distillation equipment have included the possibility of carbon dioxide addition to prevent alkaline scale formation, the tend- encies of certain elements to form ion pairs in seawater, and the rate factors in therinal-precipitation processes for calcium sulfate. 14. Effects of Radiation on Grganic Materials W. W. Parkinson and QOscar Sisman EFFECT OF RADIATION ON POLYMERS W. W. Parkinson, W. K. Kirkland, and R. M. K eV S O .o e 121 The tensile properties of polytetrafluoroethylene are retained to doses in excess of 2 x 107 rads in a vacuum; elongation at break shows a maximum at dosages near 3 X 10* rads whether irradiated in air or vacuumnl, RADIATION-INDUCED REACTIONS OF HYDROCARBONS R. M. Keyser and W. W, Parkini s O o e e 121 Gamma radiation of saturated solutions of ammonia in n-hexane and in hexene=1 produced less than the de- tectable limit (about 0.2 molecule per 100 ev absorbed) of amines. ADDITION REACTIONS OF FURAN DERIVATIVES C. D. Bopp, W. D, Burch, and W. W, Parkins Om o e e e e 123 Scolutions of dihydrofuran and cyclohexene in saturated furan derivatives were irradiated in a survey of reactions utilizing isotopic-decay radiation, and it was found that yields of adducts or dimers and trimers approached 10 molecules per 100 ev. 15. Fivoride Studies for Other ORNL Programs THE CHEMISTRY OF CHROMIUM IN THE FLLUORIDE VOLATIILITY PROCESS B. J. Sturm and R. E. T aomia . o i e e e e e e et e e ans 125 Fluorides and oxyfluorides of chromium in the oxidation states Cr(III) to Cr(VI) were synthesized and partially characterized; free energy of formation for solid Cr02F2 was found to be 189 kcal/mole at 298°K. xi PREPARATION OF LiF SINGLE CRYSTALS BY THE MODIFIED STCCKRBARGER METHOD 12, (5. IR085 AN R B, oM. it et et ettt e e et ra e e et b ann e as b reras s n et e a e e e e n e e na et aren 126 A large (270-g) single crystal of Li¥, grown in a platinum-lined capsule, was found to contain a lower con- centration of heavy-metal contaminants than is currently detectable by activation analysis, that is, <1.86 pph. 16. Chemical Support for the Controiled Thermonuclear Research Frogram R. A, Strehlow VACUUM ANALYSIS IN AN EXPERIMENTAL PLASMA DEVICE B A B0 Lottt irii et oo oeed e et en ettt nee e n e et et teek e et e e e e A eaeha e A eRs e nra bt e e e ea s 127 A relatively simple mass analyzer installed on DCX-2 and its counterpart on a much simpler similar vacuum system are being used to define the gaseous environmenti in this experimental plasma device, INTERACTION OF TRITIUM WITH THERMONUCLEAR-REACTOR MATERIALS Bt B Al L S ert i ivren ettt et ee e e et e e et e n b e et oo aa e e e eeata i aeaaeaan ety e e 128 Literature information and 2 simple diffusion model were used to estimate the steady-«state heldup of tritium in the atom-bombarded wall of a proposed thermonuclear reactor. HYDROGEN SURCHARGING OF MOLYBDENUM IN A GLOW DISCHARGE D M RICRAP S OG0l s o iee et v e te s ettt e et et e e ettt et ettt eeeee e et nen e eea et ee e n e e et et an aean e 129 Molvbdenum at elevated temperatures picks up relatively little hydrogen when pombarded with protons in a glow discharge; bombardment al lower temperatures of molybdenum whose surface is contaminated leads to veclusion of large guantities of the gas. MEASUREMENT OF GAS LOAD FROM SOURCE OF ELECTROMAGNETIC SEPARATOR B2 Al LB LMW oot ettt et oo e e4eeis et eeeeeeee e eeeeeeeestaeeeneeed ettt e e tneanan et en e e e aeetn s 129 The flow rate of chlerinating agent (0014) (rom the ion source of a calutron has been measured during operation, and the data have been applied in design of a differential pumping system for the scurce of a new electromagnetic separator. APPEARANCE-POTENTIAL MEASUREMENTS FROM TIME-OF-FLIGHT MASS SPECTROMETRY Jo D REUMATL. o o e et e e e v e e hus e e r e e e s e a e 130 The time-of-flight mass spectrometer with a retarding-potential circuit has been calibrated with several per- manent gases and shown to be capable of precize evaluation of appearance potentials with source pressures ag low as 5 » 1077 torr. PART V. NUCLEAR SAFETY 17. Nuclear Safety Tests in Major Facilities FISSION PRODUCTS FROM FUELS UNDER REACTOR-TRANSIENT CONDITIONS G. W. Parker, R, A. Lorenz and J. G. Wilhelm o e e 135 Fission product release from UO2 in stainless steel or Zircaloy cladding has been successfully determined during exposure o transients in TREAT of specimens in water or high-pressure steam. FISSION PRODUCTS FROM SIMULATED LOSS-OF.-COCLANT ACCIDENTS IN ORR W. E. Browning, Jr., C. E. Miller, Jr., W. H. Montgomery, B. F. Roberts, R. P. Shields, 0. W. Thomas, A. F. Roemer, and J. G. WIlhelm .. 137 Effects of atmosphere, cladding material, fuel burnup, maxzimum fuel temperature, and seresol aging on be- havior of fission products are being investigated in in-pile experiments. X11 FI5510N PRODUCTS FROM HIGH-BURNUPF U02 G. W. Parker, W. M. Martin, G, E. Creek, R. A, Lorenz, and C. J. Barton................ooiii i 138 Behavior of fission products from UO2 previously irradiated to 1000 Mwd/ton and melted in the Containment Mockup Facility was similar to that observed in similar experiments with simulated high-burnup fuels. THE CONTAINMENT RESEARCH INSTALL ATION G, W, Parker and W. Jo Martinl o e e et e e e e e et e e e e e 139 The Containment Research Installation, an enlarged and improved version of the Containment Mockup Facility, is nearly complete, and component testing is expected to begin early in 1966, ANALYSIS OF PLANS FOR SCALE-UP IN NUCIL.EAR SAFETY PRUGRAM C. E. Miller, Jr., and W E. BrowWmiiig, J o i ettt e ettt e e e ettt e r e e s 141 Qur analysis of experiments planned in the U.S. to study the behavior of fission products following a reactor accident suggests fhat additional intermediate experiments, 1% and 10% of the size of the LOFT, are necessary. 18. Laboratory-Scale Supporting Studies RETENTION OF RADIOIODINE AND METHYL IODIDE BY ACTIVATED CARBONS W. E. Browning, Jr., R. D. Ackley, J. E. Attrill, G. W. Parker, G. E. Creek, F. V. Hensley, R. E. Adams, J. D. Dake, D. C. Evans, and A. F ermelh oo e e e e e 142 It has been found that activated carbon impregnated with inactive 12712 or K127I removes iodine activity as CHSISII from gas streams under a variety of conditions. IDENTITY OF VAPOR FORMS OF RADIOIODINE W. E. Browning, Jr., R. E. Adams, R. D. Ackley, J. E. Attrill, J. D. Dake, and D. 1. Ford ..... ...l 144 Gas chromatography with simultaneous electron-capture and radiation detection has been utilized in determining the identity of various iodine vapor forms encountered and in ascertaining the purity of elemental iodine and methyl iodide socurces employed in laboratory iodine-behavior studies. DEVELOPMENT OF METHODS FOR DISTINGUISHING AND MEASURING GAS-BORNE FORMS OF FISSION PRODUCTS W. E. Browaing, Jr., R. E. Adams, R. D. Ackley, R. L. Bennett, M. D. Silverman, J. Truitt, W. H. Hinds, A. F. Roemer, Jr., B. A. Cameron, J. D. Dake, and D. T, Ford. ... e, 144 Characterization devices such as composite diffusion tubes, fibrous-filter analyzers, gas chromatographs, and May packs are being examined and modified for application in high-humidity experiments designed to study the be- havior of the various forms of gas-borne fission products under reactor accident conditions. REMOVAL OF PARTICULATE MATERIALS FROM GASES UNDER ACCIDENT CONDITIONS W. E. Browning, Jr., R. E, Adams, J. S. Gill, and G. L. KoChanny ... e e 145 l.aboratory investigations are being continued of the filtration efficiency of high-efficiency filter media for particulate aerosols which simulate those expected to be generated during reactor accidents. IGNITION OF CHARCOAL ADSORBERS BY FISSION PRODUCT DECAY HEAT W. E. Browning, Jr., C. E. Miller, Jr., B. F. Roberts, and R. P, Shields ... 146 Preliminary results of a study to determine the effects of iodine-decay heat on the ignition temperature of charcoal adsorbers show that the ignition temperature is not greatly affected by a decay-heat load greater than those expected at various power reactors. P B L LA T LN S e ettt ettt bttt ettt 147 PAPERS PRESENTED AT SCIENTIFIC AND TECHNICAL MEETINGS ... 153 Part | Molten-Salt Reactors 1. Phase Equilibrium and Crystallographic Studies LIQUID-LIQUID IMMISCIBILITY IN THE SYSTEM LiF-BeF -ZiF H. A. Friedman R. E. Thoma Beryllium fluoride systems have been of interest because they provide “‘weakened models™! of Si0, The weakened model contains cations and anions of the same radius as those in the silicate system, but with half the ionic charge. In MgO-S10,, a liquid immiscibility region is found; in the weakened model LiF-BeF, this has not been found, although metastable glasses, mutually in- soluble, are encountered at 60 to 90 mole % BeF, , where the melts are highly viscous. Our recent work on the Lil-BeF -ZiF system” reveals that stable liquid-liguid regions do exist in this ternary, although they do not extend stably to the LiF-Bel binary. systems. Silicate and BeF glagses are structurally anal- ogous and are composed of networks of bridging ions, oxide and fluoride ions respectively. Other cations, particularly the low velent ones, if added to these glasses, break the bridges and are referred to as modifiers; their effect on viscosity is quite pronounced. The possible role of ZrF, as a net- work modifier or network former has not previously been examined. Our liquid-liquid immiscibility studies were made with a new high-temperature centrifuge that devel- oped a centrifugal force of 530 x ¢ and gave pood layer separation in sealed metal capsules. Results 1V. M. Goldschmidt, *'Geochemische Verteilungs- gesetze der Flemente: VIII, Untersuchungen uber Bau und Eigenschaften von Krystallen®® (Geochemical Dis- tribution L.aws of the Elements: VIII, Researches on Structure and Properties of Crystals), Skrifter Norske Videnskaps-Akad, Oslo, It Mat.-Naturv. KI. 1926, No. 8, pp. 7—156 {1927); Ceram. Abstr. 6(7), 308 (1927). 2Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL.~3789, p. 3. were based on chemical analyses of portions of quenched samples obtained after 1-hr periods at elevated temperature in the centrifuge. In pre- liminary trials with PbO-B O, glasses, good agree- ment was obtained with the work of Geller and Bunting.” lsothermal tie lines at 550, 650, and 700°C for the immiscibility region in LiF-Bel, - ZrF , are shown in Fig. i.1. The possibility of stable liquid-liquid immis- cibility in the LiF-BeF, system was given special attention. This was in part prompted by results on the chemical activity of BeF, as a function of composition in the LiF-BeF, binary, as obtained from HF-H, 0 equilibria,” where there was an indication that immiscibility might occur for 80 mole % BeF , at temperatures above 700°C. Ten samples representing six binary composi- tions in the range from 80 to 34 mcle % Bel, were centrifuged and examined; no phase separation was found. Further, visual observations of an 80 mole % BeF | melt were made under helium in a glove box and furmace assembly by Cantor and Ward.’ Again the liquid was single phase, although on the initial heating to 700°C a lower viscous layer and a thinner, easily stirred, upper layer persisted; only after heating to 850°C did the melt become homogeneous. The melt remained homogeneous through subsequent temperature cycles ifrom room temperature to 850°C. The conclusion was reached that there is no stable immiscibility above the liguidus in the LiF-BeF , hinary. 3R. F. Geller and E. N. Buonting, J. Res. Nail, Bur. Std. 18, 585 (1937). *A. L. Mathews and C. F. Baes, Oxide Chemistry and Thermodynamics of Molten Lithium Fluoride—Beryilium Fluoride by Egquilibration with Gaseous Water—Hydrogen Fluoride Mixtures, ORNL-TM-1129, p. 104 (May 7, 1965). 5, Cantor and W. T. Ward, MSRP Semiann. Progrc Rept. Aug. 31, 1965, ORNL-3872 (in press). Zrfa 3LIF-4ZrF, J N Z 2 LIQUIDS ~_ 2LIF-ZrF, /0 N \/ LiF Fig. 1.1. Liquid-Ligquid Immiscibility in the System LiF-BeF,-ZrF,. (c) 700°C isotherm. SODIUM FLUORIDE-SCANDIUM FLUORIDE PHASE EQUILIBRIA R. H. Karraker® R. E. Thoma One of the most important factors which influence the number and types of intermediate compounds SORINS research participant, Memphis State University, Memphis, Tenn. ORNL-DWS 66 - 9614 BeF, (a) 550°C isotherm; (b) 650°C isotherm; formed between pairs of highly ionic compounds such as the alkali fluorides and group III fluorides is the relative size of the cations.” Size effects are of diminishing significance as bonding becomes less ionic or when unusual ligand effects, such as those induced by the lanthanide and actinide contractions, become of importance. In the course ’R. E. Thoma, Inorg. Chem. 1, 220 (1962). | | 1300 - —1 Nt/ ludt =146 | {°C} Hoo - 900 N MPERATUR e 1 = 5 Naf + 9 Lui 700 b N e e d e 500 300 L—— ’ (C‘)J { ( NaY/Sc o et 2EE e b ‘Naf/m“: 248 — T ORNL—-DWG 66 ~962 e e W’ JF}} = E] . | —dhal = “} | o JI’_J By AR | Fig. 1.2. Comparison of the Systems {a) NaF-LuF3, (b) NaF-ScFj3, and {c} NaF-AlF4. of the recently completed investigation of the sodium fluoride—lanthanide trifluoride systems (see *#Compositional Variability in Sodium Fluoride— Lanthanide Fluoride Complex Compounds,” this report), it became evident that within the series cation charge density was beginning to obscure the effect of size. It was therefore of great interest to obtain ioformation concerning the comparative phase behavior among the following group of sys- tems, which are ranked below in order of increasing difference of the tripositive cation to alkali cation radius ratio: T\IaJF«LaF3 0.923 NaF-BiF3 1.054 NaF-Lqu 1.156 ]\I::flf"-InI*“3 1.210 NaF-8clk 3 1.256 K.F-LaFS 1.254 NaF-AIFS 2.178 Investigation of the system NalF-ScF,, reported previously in preliminary form,® has now been completed. The equilibrium phase diagram of the system is compared with those of NaF-LuF, and Nab-AlF, in Fig. 1.2. Equilibrium reactions of NaF and SLFJ bear closer resemblance to those of NaF and AlF, than to NaF and LuF the 5o despite fact that the scandium ion radius (D.78) is 8Rrea(:tor Chemn, Div. 1965, ORNL-378%9. Ann. Progr. Rept. Jan. 31, nearer to Lu®? (0.848) than to A1** (0.45). Similar to the Nal-AlF , system, molten NaF-ScF mix- tures form a cryolite-like phase, 3NaF.Sck which, like cryolite, is dimorphic. Crystals of each form are isomorphous with their cryolite analogs. The phase 3NaF -ScF, undergoes a strongly exothermal solid-state inversion when cooled below 680°C and inverts to poorly crystallized and highly twinned material. Single crystals of the low- temperature form of 3Nal - ScF , were obtained by growth from NaF-ScF, mixtures of 70-30 mole % composition. X-ray diffraction analysis of these single crystals indicated that they are of mono- clinic symmetry (see “*Crystallographic Data on New Compounds,*’ this report) and are isomorphous with cryolite. A high-temperature foem of cryolite is reported” to occur as a face-centered cubic phase. Preliminary results of high-temperature x-ray diffractometric experiments with 3NalF . ScF indicate that the high-temperature forms of 3Nal - 5c¢F, and 3NaF - AlF | are isomorphous. Whereas the intermediate phase in the 3Naf - AlF -AlF , subsystem is of 5:3 composition, only a 1:1 phase is formed in the 3NaF -ScF -5cF, subsystem. The compound NaF - 5ck | melts incon- gruently to ScF, and liquid at 660°C. X-ray dif- fraction data obtained from single crystals of NaF - 5¢F, revealed it to be of hexagonal ' sym- metry (see ““Crystallographic Data on New Com- pounds,”’ this report). 9E, G. Steward and H. P. Rooksby, Acta Cryst. A, 48 (1953), Table 1.1. Inveriont Equilibria in the System NaF-ScF Compositicn Invariant (mole % SCFS) Temperature (°C) L refers to liquid Type of Equilibrium 3 Phase Reaction at Specified Temperature 17 800 Eutectic L. == NaF 4 @ -3NaF - S(:F3 25 885 Congruent melting L &= a-3NaF -Sch point 35 680 Inversion of 3NaF -ScF 4-3NaF -ScF ; === f3-3NaF -ScF peritectic 38 650 Eutectic L= B~3NaF . S(:F3 + NaF . ScF3 42 660 Peritectic We believe that the greater resemblance of the system NaF-SCF3 to NaF-AlF, rather than to NaF-LuF3 arises as a result of the lanthanide contraction, producing the Lu®" ion of extra- ordinarily high charge density. Quantitative rela- tionships of the combined effects of ion size and charge density have been developed by Dietzel.1? They require accurate information of interionic distances, which for these complex fluorides must await determination of the crystal structures. The composition and temperatures of invariant equilibrium points in the system Na\F-ScF3 are listed in Table 1.1, COMPOSITIONAL VYARIABILITY IN SODIUM FLUORIDE-LANTHANIDE TRIFLUORIDE COMPLEX COMPOUNDS R. E. Thoma H. Insley (. M. Hebert Investigation of the sodium fluoride—~lanthanide trifluoride systems'! has been completed. A salient feature of the NaF-LnF , binary systems is the unusual behavioral sequence produced by the compositional variability of the fluorite-like phases. This behavior shows the remarkable effect of cation size and polarizability in lanthanide systems. As shown in Table 1.2, the cubic phase is not formed at all in the systems NaF-LaF and IOA, Dietzel, Z, Elektrochem, 48, 9 (1942), 11I'\’e-ac:tor Chem. Div. Ann. Frogr. Rept. Jan. 31, 1965, ORNL-3789, p. 13. L+ SCF3 == I. + NalF -SCF3 NaF~CeF3, is formed by Nal' and PrFS, and is extended to an increasingly broad composition range throughout the rest of the system sequence. A trend also develops toward broadening the com- position limits and extension of the temperature range through which the cubic and orthorhombic phases are stable. The trends toward increasing number, compositional variability, and stability of the NaF-LnF, crystal phases appear to be dis- tinct, not only with respect to the crystal chem- istry of the complex compounds, but also in their modes of crystallization. No satisfactory theory has as yet been developed to explain the quanti- tative aspects of compositional variability of NaF-Lnk , phases. Qualitatively, the trends are believed to indicate an increase in the lattice energies of the crystalline phases resulting from the lanthanide contraction and a corresponding decrease in polarizability. The decrease in polar- izability is probably the most influential factor in enabling compositional variability of the phases to be extended with increasing atomic number of the lanthanide. As represented by the Lorentz-Lorenz equation, molar refraction approximates the volume occupied by the constituent ions in crystalline phases. A measure of the crystal free space can therefore be estimated from the difference between molar volume, as computed from unit-cell dimen- sions, and the molar refraction. As the atomic number increases, small increases in the free space fraction are noted for the lanthanide tri- fluorides and for the hexagonal NaF - LnF , phases. A more pronounced increase in free space fraction is found for the fluorite-like NaF-LnF3 phases, Table 1.2. Optical and X-Ray Diffraction Data for NuF-LnF3 Crystalline Phoses A. Hexagonal NoF * Lnf-’3 Ln ag (A) T(ay) cq (A) T(cy) N N ¢ o La 6.157 C.006 3.822 0.008 1.486 1.500 Ce 6.131 0.006 3.776 0.004 1.493 1.514 Pr 6.123 0.604 3.743 0.002 1.494 1.516 Nd 6.100 0.002 3.711 0.002 1.493 1.515 Pm (6.056) (3.670) (1.492) (1.515) Sm 6.051 0.61¢ 3.640 G.007 1.492 1.516 Eu 6.044 0.603 3.613 0.003 1.492 1.516 Gd 6.020 0.003 3.001 0.008 1.483 1.507 Th 6.003 0.004 3.580 0.002 1.486 1.506 Dy 5.085 0.004 3.554 0.003 1.486 1.510 Ho 5.851 0.601 3.528 0.002 1.486 1.510 Er 5.959 0.002 3.514 0.002 1.482 1.504 Tm 5.953 0.6032 3.494 0.002 1.476 1.496 Yb 5.929 0.002 3.471 0.002 1.482 1.504 Lau 5.912 0.003 3.458 0.003 1.484 1.506 (Y5 5.967 0.002 3.523 0.002 1.464 1.486 8. Cubic NqF-LnF3 Phases ot Composition Limits NaF-Rich Composition Limit SNal - QLnFS Composition Limit b Compositioa X-Ray Density ~ X-Ray Density (mole % LnFS) R. L 2o (4) (g/ce) R.L g (a) (g/cc) Pr 55.5 1.512 5.710 1.526 5,720 5.768 Nd 55.0 1.506 5.670 4.431 1.524 5.685 £.962 Pm (54.5) (1.5C03 (5.630) (4.620) (1.522) (5.5655) (6.131D Sm 53.5 1.485 5,605 4,702 1.520 5.627 5.315 Eu 52.5 1.486 5.575 4.809 av 1.519 5.616 6.392 Gd 51.5 1.470 5.550 4,978 1.502 5.594 6.620 Th 0.0 1.472 5.535 5.051 1.504 5.565 6.772 Dy 48.5 1.462 5.505 5,205 1.504 5.547 £6.939 Ho 47.0 1.458 5.490 5.286 1.504 5.5125 7.093 r 45.0 1.440 5.475 5.387 1.493 5.514 7.263 Tr 43.5 1.427 5.460 5.4&6 1.494 5,493 7.336 i 41.5 1.415 5.444G 5.611 1.488 5.480 7.510 La 29,0 1.4G2 5.425 5.6938 1.482 5. 463 7.580 . Orthorhomhic SNoF - ‘?Lan Kefractive Index Lattice Constants (A) Lo . SO . ) N, N, N a, b, cq Dy 1.514 5.547 27.74 4.804 Ho 1,510 5.525 27.63 4.784 Fr 1.504 1.506 1.506 5.514 27.57 4.775 I'm 1.501 5.493 27.47 4.757 Yh 1.482 1.494 1.495 5.480 27.40 4.746 L 1.480 1.496 1.487 5.463 27.32 4,731 Table 1.2 {(continued) Refractive Index Lattice Constants (A) Ln Symmetry Density Ref. NgorNa NworN,y a, .!b0 o (g/ec) La Hexagonal 1.597 1.603 7.186 7.352 5.936 a Ce Hexagonal 1.607 1.613 7.112 7.279 6.157 b Pr Hexagonal 1.614 1.618 7.075 7.238 6.14 C Nd Hexagonal 1.621 1.628 7.030 7.200 6.500 a Sm Orthorhombic 1.577 1.608 6.669 7.059 4,405 6.643 a Eu Orthorhombic 1.572 1.600 6.622 7.019 4.396 6.793 d Gd Orthorhombic 1.570 1.600 6.571 6.985 4.393 7.056 e Tb Orthorhombic 1.570 1.600 6.513 6.949 4.384 7.236 e Dy Orthorhombic 1.570 1.600 6.460 6,906 4.376 7.465 e Ho Orthorhombic 1.566 1.598 6.404 6.875 4.379 7.644 e Er Orthorhombic 1.566 1.598 6.354 6.848 4.380 7.814 e Tm Orthorhombic 1.564 1.598 6.283 6.811 4,408 7.971 e ¥b Orthorhombic 1.558 1.568 6.216 6.786 4,434 8.168 Lu Orthorhombic 1.554 1.558 6.181 6.731 4.446 8.44 “E. Staritzky and L. B. Asprey, PASTM X-Ray Diffraction Card No. 8-45. “ASTM X-Ray Diffraction Card No. 6-0325. Anal. Chem. 29, 857 (1957). 9a, Zalkin and D. H. Templeton, J. Am. Chem. Soc. 75, 2453 (1953). ®ASTM X-Ray Diffraction Card No. 12-788. with an even greater increase in the free space fraction for the equimolar cubic phases than for the cubic SNaF :9LnF , phases as Z increases. Sub- stitutional solid solution of Ln®" into the fluorite unit cell gives rise to cation vacancies but is partly compensated by filling of interstitial posi- tions with fluoride ions, as described by Roy and Roy.!? The fact that the free space fraction occupied by the ions in the cubic structure at the S5NaF - 9LnF3 saturation limit is even greater than composition lends additional evidence to the validity of the previous interstitial fluorine model for solution mechanism. We conclude that the combined effect of the reduction in polarizability of the lanthanide ions and the increase in free space within the crystal lattice is to reduce the specificity of Na® and Ln®"' ions with regard to the cation sites they occupy in fluorite-like, orthorhombic, and hexagonal at the equimolar '2b. M. Roy and R. Roy, J. Electrochem. Soc. 421 (1964). 111, crystals. The consequence of this effect is the increase observed in compositional variability of the crystal phases in the NaF-LnF , systems with increasing atomic number of the lanthanide. PHASE EQUILIBRIUM STUDIES IN THE U0,-ZrO, SYSTEM K. A. Romberger H. H. Stone C. F. Baes, ]Jr. In a previous report, a new phase diagram was suggested for the UO,-ZrO, system in which essentially no solid solution formation was indi- cated below about 1100°C.'° which is in variance with previously published This conclusion, 13 A. Romberger et al., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORN1.-3789, p. 243. ORNL.-DWG 5&-~963 MOLE PERCENT U0, N0 99.5 990 98.5 100 90 80 70 €0 50 40 30 20 10 O 05 1.0 0.5 0 I T T 7T T Ty T [T 1go - CUBI —gsoc L1QuID ZBO()/ ~ TETRAGONAL s.s. 1180 o \ i 140~ © CUBICs.s. — 2400 SLIQUID - 1040 o + \ + o TETRAG. 5.5, \ CURIC s.5. \ ‘. HOO |- 2000 - - \ 100 4 CLBIC s. 8. —_— 4 - o 4 TETRAG.| |/ CUBIC s.s. 2 o — ' S5.5. - o DE0 — 1800 |- ~11600 |- ~ 1080 1 : CUBIC s.5. \ | 7T s s I CUBIC < 5. o iz 1 -+ . + et 2 MONCCLINIC .. " li VETRAGGNAL 5.5 il | MONOCLINIC s.5. = 414020 - 1200 ) Y200 w1020 ¢ = o * $ % x4 O = = o J | I A & Tl l C s s 230 | - 800 | CUBIL 5.5, l »‘;L 500 |- | 980 i ! \ / | MONDCLINIC s.¢. | /, \ / | | /, \ MONGCLINIC 5.8, 240 it — 400+ | |/ ] 1400 - - 340 . // I i! ’ \ L e e L \ 500 [ | ol v e ey N A Jsoo 0 0.5 1.0 £5 0 10 20 30D 40 S0 60 TO BC 90 100 385 920 995 100 MCLE PERCENT Zr0, LEGEND : LIQUID - LAMBERTSOMN AND MUELLER (23) COHEN {161} MURMBPTON AND ROY (22) (ABCVE 1600°C] ——— (BELOW $600°C) -~ ) O "WET" CHEMICAL ANALYSIS X ACTIVATION ANALYSIS FOR U AND SCHANER (BELOW 1800°C ) ORNL PROPOSED Fig. 1.3. U0,-Zr0, Phase Diagram. phase diagrams for this s;ystem,””w was based primarily on x-ray and petrographic examination ot UO,-Zr0, solids equilibrated with fluoride melts al temperatures from 500 to 700°C'? and frem 900 to 1020°C."? 141 (1963). 154 160 Cohen and B. E. Schaner, J. Nucl., Mater. 9, 18 M. Wolten, J. Am. Chem. Scc. 80, 4772 {1958). E. Evans, f. Am. Ceram. Soc. 43, 443 (1960). R. Wright, D. E. Kizer, and D. L. Keller, Studies U()Z-Z,r()2 System, BMI-1689 (Aug. 27, 1964). 17T. in the 185, M. Voronov, E. A. Voitekhova, and A. S. Damlin, Proc, U.N. Intem. Conf. Peaceful Uses At. Energy, 2nd, Geneva, 1958 6, 221 (1958). 19¢. F. Baes, Jr., J. H. Shaffer, and H. F. McDuffie, Trans. Am. Nucl. Soc. 6, 393 (1963). Equilibrations of the oxides with fluoride melts have since been completed for temperatures up to 1150°C. The oxide solids from these latter equil- ibrations were not only examined by x-ray and petrographic techniques as before, but also these and all of the previously equilibrated solids have been chemically analyzed io determine the U-Zr composition of the individual phases. (The separa- tion of the phases was performed using hot nitric acid, a liquid in which UO, is readily soluble and Zr0O, is virtually insoluble.) Incorporation of this new informalion gave the more detailed diagram for the U0 ,-ZtO, system shown in Fig. 1.3, The monoclinic-tetragonal transformation of the Zr0 -rich solid solutions occurred at 1110 1 5°C. At this temperature Z:0, is soluble in cubic UO, to about 0.40 mole %, while UO, is soluble in monoclinic ZiO, to about 0.15 mole %, and in tetragonal ZrQ, to about 1.0 mole %. temperature for pure ZrO, has been given by Mumpton and Roy?® as 1170°C. No attempt was made to determine this inversion temperature in the present work. However, microscopic investiga- tion of a sample of pure ZrO, equilibrated at 1190°C has shown that it definitely had been tetragonal. Hence, the Mumpton and Roy value of 1170°C falls within our limits of 1110 to 1190°C. Below the transition temperature of 1110°C, the saturating solids are the cubic UQ,-rich and the monoclinic ZrO _-rich solid solutions. However, even at 1110°C, these solid solutions are very dilute. ities will decrease exponentially as the tempera- The monoclinic-tetragonal transition Moreover, it is expected that the solubil- ture is reduced. Hence, for all practical purposes, the equilibrium solids at lower temperatures are the pure phase materials, that is, pure U0, and pure Zr()z. The correlation between the data given in the present work and results presented previously by other authors'*~ 182921 {5 al50 shown in Fig. 1.3. For temperatures below 1700°C, solid lines have been used to represent the solubility limits as deterinined from log XMO2 vs 1/7T plots of our data and all other pertinent data presently avail- able in the literature.!*~ 1829 The dashed lines those from the recently published diagram of Cohen and Schaner.'* These lines begin to deviate from each other at about 1600°C. As the temperature is further reduced, the deviations are increase rapidly. We believe the differences in the results reported here and those reported previously by others reflect the inability of the other investigators to obtain equilibrinm products at temperatures below 1600°C solely via solid-state reactions. These temperatures are more than a thousand degrees below the melting points of the pure oxides. At such temperatures, solid-state reactions of the oxides should be very slow. In the present investigation the use of a fused fluo- ride phase provided a path by which the slowness of the solid-state reaction could be overcome and equilibrium could be achieved. relatively low 20p. A Mumpton and R. Roy, J. Am. Ceram. Soc. 43, 234 (1960). 21W. A. Lambertson and M. H. Mueller, J. Am. Ceram. Soc. 36, 365 (1953). 10 THE CRYSTAL STRUCTURE OF LiUF, G. ID. Brunton The compound LiUF_ was originally described as Li,{,U6F31 in the molten-salt system Lik- UF4.22"25 The crystal structure was determined in order to resolve the stoichiometry and because LiUF _ is the uranium salt which would be deposited if fue! should freeze in the Molten-Salt Reactor Experiment. Tetragonal LilJE crystallizes in space group I41/a with a = 14.884 + 0.002 A and ¢ = 6.5467 + 0.0003 A. The calculated density is 6.23 g/cc with 16 formula weights per unit cell, parameters are listed in Table 1.3. Atomic Figure 1.4 is an illustration of two asymmetric units of LiUF | related by a center of symmetiy — approximately one-eighth of a uunit cell. 1.5 is a stereoscopic pair showing the contents of one unit cell. Figure Each uranium is surrounded by nine ftluorine ions which form the corners of a 14-faced polyhedron having the form of a prism with triangular bases and with a four-faced pyramid on each of the three prism faces. Each lithium ion is surrounded by six fluorine ions at the corners of a distorted octahedron. The structure of LiUF . consists of 16 uranium- containing polyhedra and 16 lithium-containing octahedra linked together so that every uranium polyhedron shares an end with its centrosym- metrical neighbor, shares corners with five other uranium polyhedra, and shares a face, an edge, and two corners, respectively, with four lithium octahedra. Each lithium octahedron shares edges with three other lithium octahedra, and shares an edge, a face, and two corners, respectively, with four uranium polyhedra. This linkage gives spiraling cross-connected chains parallel to the C axis. 221.. A. Harris, The Crystal Structures of 7:6 Type Compounds of Alkali ¥Fluorides with Uranium Tetra- fluoride, CF-58-3-15 (March 6, 1958). 23¢, J. Barton et al., J. Am. Ceram. Soc. 41, 63 {1958), 241, A. Harris, G. D. White, Phys. Chem. 63, 1974 (1959). 25C. F. Weaver ef al., J. Am. Ceram. Soc. 43, 213 (1960). and R. E. Thoma, J. 11 F5 ORNL - DWG 65 ~{2543 N o - o Fig. 1.4. Two Centrosymmetrically Related Asymmetric Units of LiUdFg. Table 1.3, Atomic Parameters far LiUFS Atom x + % 10° U 0.06176(0.4) 0.05640(0.4) 0.24623(1.7) 0.00026(0.3) F, 0.0361(9) 0.0523(9) 0.8734(25) 0.0017(2) F, 0.2008(9) 0.1754(9) 0.7602(28) 0.0018(2) F, 0.10859) 0.1831(9) 0.7504(22) 0.0013(2) F, 0.2048(10) 0.0918(10) 0.3576(26) 0.0020(3) F, 0.0484(%) 0.1677(10) 0.4787(26) 0.0018(2) Li 0.068(30) 0.163(31) 0.773(84) 0.0015(%) *Isotropic temperature factors: 822 'Bll ' c*2 B .~ R . 337 L7 12 =77 13 23 ORMNé_—DVWG 65 12542 Fig. 1.5. One Unit Cell of LiUF; (Stereographic Pair). THE CRYSTAL STRUCTURE OF Li AlF, 26 J. H. Burns A. C. Tennissen A recent phase diagram study on the ternary systems involving AlF ; and LiF, NaF, and KF 27 26Research Participant from Lamar State College of Technology, Beaumont, Tex. 2’R., E. Thoma, B. ]J. Sturm, and E. H, Guinn, Molfen- Salt Solvents for Fluoride Volatility Processing of Aluminum-Maérix Nuclear Fuel Elements, ORNIL.-3594 (August 1964). indicated that LisAlFfi, NaSAlFé, and K3A1F‘6 are each components of these systems; and while the latter two were known to have the cryolite structure, LiSAIF() had not been studied. The existence of sixfold and twelvefold coordination for the alkali-metal ions in cryolite was not ex- pected to hold for Li*, and indeed a new structure type was found. Crystals of L13A1F6 were prepared by melting together under vacuum a 3:1 mixture of LiF and AlF .. Some of the compound distilled to a cold finger in the vessel, but the larger crystals were obtained from the residue in the bottom. Powder x-ray diffraction examination showed both to be virtually the same phases, with a little excess AlF , on the cold finger, X-ray precession photographs were used to establish that the crystals were otthorhombic with unit-cell dimensions: a = 9.54, b = 8.23, ¢ = 4.88 (£0.02 A). Systematically absent reflections: hOf, L = 2n, and 0k, k + | = 2n, indicated the probable space groups Pama and Pna? .» which differ by the presence of a center of symmetry. The density, calculated with four formula weights in the unit cell, is 2.80 g/cm?®. The density was measured to be between 2.6 and 3.0 by flotation in heavy liquids. | Intensity data for determination of atomic posi- tions was obtained by photographing Akl zones, 13 with I =0, 1, 2, 3, and 4, employing the multiple- {ilm Weissenberg technique. The intensities of the reflections were estimated by comparison with calibrated film strips. Some 238 independent reflections were measured and their intensities reduced to structure factors, F . The Patterson vector method was applied to locate the Al and F atoms, and calculation of a difference Fourier map then was successful in determining the Li positions. All atoms of the structure occupy fourfold general sites of space group PnaZl. Refinement of atomic positions and individual isotropic temperature factors was car- ried out by least squares. the discrepancy factor, The present value of ZlF, |- £ VE R, z is 0.13, ORNL—DW3 66 — 259 Fig. 1.6. Crystal Structure of LizAlF,. It is expected that somewhat better agreement will be achieved with further refinement, but the atomic sites given in Table 1.4 are not expected to change appreciably. A drawing of the Li AII'_ structure is shown in Fig. 1.6. The presence of A1F63“‘ ions is apparent; these are joined together in such a manner as to provide octahedral coordination for each of the Lit ions. Bond distances are within the normal range, but further refinement of the structure will be awaited before reporting them. Table 1.4. Atomic Porameters for Li3A1F6 Atom X y z Li{1) 0.357 0.372 0.525 Li(2) 0.131 0.407 0.499 Li{3) 0.349 0.527 0.017 Al 0.128 0.246 0° F(1) D.222 0.086 0.145 F({2) 0.026 0,249 0.299 F(3) 0.238 0.212 0.694 F{4) 0.022 0.390 0.856 F(5) 0.244 0.391 0.185 F(6) Q0.020 0.091 0.837 Chosen arbitrarily to fix the origin on 2 . REFINEMENT OF THE CRYSTAL STRUCTURE OF (NH ) ,MnF, D. R. Sears The manganic ion is unstable with respect to disproportionation, and as a consequence, simple salts of trivalent manganese are rare. Neverthe- less, it is desirable to collect definitive informa- tion on the structures of manganic salts when possible, for complex anions containing trivalent manganese can be expected to exhibit significant crystal field effects. In particular, octahedrally 14 coordinated Mn>" should be tetragonally distorted as a result of the Jahn-Teller effect.?8 The compound (NH ) MnF _ crystallizes in the centrosymmetric orthorhombic space group D;:- Pnma, with a = 6.20 = 0.03, b = 7.94 + (.01, and c = 10.72 £ 0.01 A. spectrometric zonal =x-ray intensity data were collected in 1957 from specimens of (NH,),MnF _ prepared by T. S. Piper. The data had resulted in a structure determination,?® but it had not been possible to perform a satisfactory refinement with the computing facilities available then. The opportunity arose to perform a modern full matrix least-squares with the existing AOI and Ok! intensity data, using a modifi- cation of the Busing-Martin-Levy program ORFLS.?? Anomalous dispersion effects were included in the calculations. Final least-squares-adjusted atomic positions and thermal parameters resulted in an agreement factor, Excellent photographic and anisotropic refinement between observed and calculated structure factors, lFO | and ’FC , as low as 0.068 for all observable data. The contents of one unit cell and adjacent atoms are shown in Fig. 1.7. Each octahedron has the composition MnF , with manganese occupying the center and fluorines the vertices. Ammonium groups appear as stippled spheres and form infinite hexagonal nets in mirror planes at y - 1/;, ?‘,;_ In- finite kinked chains of MnF octahedra, sharing opposite vertices, pass through these ammonium nets. The manganese atoms lie on symmetry centers, and consequently there are but three crystallographically distinct fluorines. Bond distances and angles are presented in Table 1.5. In this table, £ is the fluorine atom shared by adjacent octahedra. Nonbonding contacts between adjacent fluorines in the octahedra were in the range 2.59 to 2.84 A. Although nitrogen- fluorine distances as low as 2.81 were discovered, 25"See*_, for example, R. Dingle, Inorg. Chem. 4, 1287 (1965). 29D. R. Sears, Ph.D, thesis, Cornell University, 1G658. 30, R. Busing and K. O. Martin, ORFLS, a Fortran Crystallographic Least-Squares Program, ORNL~TM-305 (1962). 15 ORNL-DWG 65-5847 Fig. 1.7. Schematic View of Unit Cell of (NH,);MnF s and Selected Atoms of Adjocent Ceils. Manguanese occurs at the center of sach octabedron, ond fluorine at earch vertex. Nitrogen atoms are stippled spheres. Table 1.5. Bond Distances™ and Angles Distances Bond Type Length {A) Standard Error {(A) (2) Mn-F 2.091 (2.101) 0,005 (0.006) (&) ¥n-F | 1838 (1.853) 0,009 (0.009) (2) MooF 1.842 (1.853) 0.004 (0.004) Angles Bond Angle Angle Standard Error FoMn-F 89.4° 90.5° 0.5° F Mo-F o 37.7° 92.3° G.3° F MaeF 89.7° 90.3% 0,3° 143.4° 0.8~ I’vIn-FI-Mn Mistances in parentheses are corrected for themmal metion, with fluorine assuamed to **ride’” op the manganese. a high coordination number of nitrogen (7 to 9) and the infrared spectrum®® of {NK~§4)2MnF5 suggest strongly that N-H-I' hydrogen boonding is abszent ot very weak. The bond distasces and angles involving man- ganese reveal that there iz indeed a pronounced tetragonal disiortion of the octahedral manganese coordination shell. Doubtless, sharing of the F atoms between adiacent octahedra contributes to the substaniiel elongation of the Mn-F_bonds, but a Jahn-Teller digtortion must contribute signifi- cantly. However, quontitative separation of effects is not practicable. the Dingle®® has discussed both the crystal spectrum of (NH 4),}M11F . and the solution spectrum of trivalent in concentrated HF with excess K7 manganesea , presumed to contain MnF63’". The nearly perfect correspondence of the spectra in the 10,000 to 28,000 cm™! region suggests that tetragonally distorted MnFé3m octahedra appear in solution; such a disftortion in aqueous MnF63” solutions surely cannot result from chain formation. We infer that the distortion arises largely from ligand field effects in solution and, by extension, does so also in the crystal. HIGH-TEMPERATURE X-RAY STUDIES G. D. Brunton J- H D. R. Sears . Burns A Materials Research Corporation high-tempera-~ ture x-ray diffractometer furnace has been employed in several studies of phase change at elevated temperatures, with the following resulis. The orthorhombic lanthanide trifluorides SmF through LuF , and YF have been shown to undergo a phase transition below their melting points.® Of these compounds, SmF . through HoF , invert to the hexagonal LaF , structure while YF , and ErF, through Lul , invert to a crystalline form which is probably hexagonal but whose unit-cell dimen- sions suggest that it is different from the LaF structure. The unit-cell dimensions of the hexagonal phases and the approximate inversion temperatures are summarized in Table 1.6. Table 1.6, UWnit-Cell Parameters and Approximdte Inversion Temperatures Transition “0 (4 “o (A) Temperature (OC) SmkF 3 7.07 7.24 355 EuF‘3 7.04 7.26 700 GdF 7.06 7.20 900 TbF 7.03 7.10 950 DyF3 7.01 7.05 1030 HoF 7.01 7.08 1070 ErF3 6.97 8.27 1075 TmF3 7.03 8.35 1030 YbF 6.99 8.32 985 LuF3 6.96 8.30 945 YF 7.13 8.45 1052 16 Phase equilibrium studies in the UO -ZiO, system have been continued, the principal objects being: (1) to investigate the composition-tempera- ture region in which Thoma®! has postulated the existence of a compound of approximate composi- tion UO 3Zr02, (2) to furnish x-ray diffraction data for the continuing fluoride-flux c‘qulhbratmn experiments of Romberger, Stone, and Baes;*? and (3) to develop constant—composition relationships for the sys- tem UO, (cubic)-ZrO (cubic). Tentative results to date suggest that the sys- tem cubic UOzwcubic Zi0, at room temperature displays a linear relationship between a and mole fraction Z10,, at least to 50% Z10,. The data can be extrapolated to a, = 5.08 A at 100% ZtO,, in reasonable agreement with the results obtained by extrapolation of the CeO,-Zr0O, data (5.11 A) due to Duwez and Odell.?? Attempts to establish the existence of a compound of composition near UO,k -3Z:0, above 1600°C have not been successful. We have not obtained pure tetragonal starting material near this mole ratio. X-ray diffraction experiments in the temper- ature range 1400 to 1650°C at a ZrO,:UO, ratio of 1.36, using a pure tetragonal solid-solution starting material, have resulted in exsolution, to form nearly pure cubic UO, and tetragonal ZrO - UQ, solid solution richer in Z10 . Our high-temperature diffractonieter attachment is being modified extensively to permit operation at higher temperatures in better vacuum and to diminish problems in window opacity caused by filament and sample volatilization. room-temperature lattice- Some high-temperature x-ray diffraction meas- urements were made in order to relate our single- crystal study of Li3A1F6 to the reported* poly- morphism: Qo sy ——> O i 225° 475° 575° 705° € . The pattern for the Li,Al¥ phase whose single crystals were analyzed (see Burns and Tennissen, BIR. E. Thowa, Reactor Chem. Div. Ann. Progr. Repi. Jan. 31, 1965, ORNL.-3789, p. 245, 32K, A. Romberger et al., Reactor Chem. Div. Ann. Praogr. Rept. Jan. 31, 1965, ORNL.-3739, p. 244, 33p, Duwez and F. Odell, J. Am. Ceram. Soc. 247 (1950). 34P. Gross, Fulmer Research Institute, Stoke Poges, FEngland, unpublished results. 33, 17 Table 1.7. Crystallographic Data Compound .'P~TaSc1:*‘4 NagScF 6 BI-KLaF4 I_,i4UF§3 RbPaFG System Hexagonal Monoclinic Hexagonal Orthorhombic Orthorhombic Unit-cell a=12.97 0,03 a=560 £0.02 2=6.53010.002 a-=9.960 £0.001 a-8.06 £0.02 dimensions, A c=9.27 T 0.02 bh=581 12002 c=23.8001+0.002 5:=0.882 L0.001 b=12.00 =0.03 c=8.12 *0.02 c=5.986 1 0.002 c:= 585 3 0.02 [3=90%5" 57 P3 12) P3 21\ Space group P3212) or P3221/ PZl/n Formula weights 2 par unit cell Calculated 2.87 density, g/cm3 Notes a a b P53 or P63/m Prnma or PnaZl Cnuna or Abm?2 3/2 4 4 4,51 4.71 5.05 c d 3ee “Sodium Flusride—Scandinm Fluoride Phase Equilibria,’’ this report. bIsostmctural with cryolite, NasAIFS, “Previously studied by W. 1. Zachariasen, Acfa Cryst. 1, 265 (1948). dPrepared by O, L. Keller, ORNI, Chemistry Division. this report) was calculated. With it and the knowl- edge of the sluggishness of the a —~ § transition, it was concluded that the preparation of the com- pound resulted in a mixture of f§ and y. This mixture was heated to 720°C in the diffractometer futnace, where both of the original components disappeared; by slow cooling the component cor- responding to the single crystals reappeared at about 500°C. From these data we tentatively concluded that the single-crystal study is of v-Li AIF . Further analysis of this compound will be carried cut when the diffractometer is modified, CRYSTALLOGRAPHIC DATA ON NEW COMPOUNDS J. H. Burns D. R. Sears G. D. Brunton Unit-cell dimensions and symmetry information for five compounds have been obtained by single- crystal x-ray diffraction and are summarized in Table 1.7. These data are necessary in order to determine whether each compound has a known structural type; or, if it does not, they are a prerequisite for proceeding with the structure determination. Of these compounds, Na ScF was found to be of the cryolite type (N33A1F6), while the others represent new structure types. Three of these, 8 -KLaF , Li, UF _, and RbPaFfi, are being subjected to complete structure analyses. THE CRYSTAL STRUCTURE OF Na,]ZiréF:” J. H. Burns R. D. Ellison®? H. A. Levy?®® This compound is representative of a structural type which has occurred in numerous moften-salt systems. Among those which have been observed?® PORNL Chemistry Division. 36p . K. Thoma, ed., Phase Diagrams of Nuclear Reactor Matsrials, ORNL-~-2548 (Nov. 2, 1959), 18 Table 1.8. Struciural Parameters for No7Zr6F3]a Atom X y z f))“ [‘322 633 BIQ _['313 623 Na(l) 0.0793 0.3038 0.4827 0.0033 0.0041 0.0050 0.0020 --0.0006 —0.0001 Na(2) 0 0 1/2 0.0046 0.0046 0.0054 0.0023 0 0 Zr 0.189¢6 0.0515 0.1791 0.0017 0.0013 0.0020 0.0007 0.G002 --(,0001 F 0.3556 0.1114 0.0016 0.0020 0.0022 0.0033 0.0010 0.0005 —0.0006 F{2) 0.1837 0.0554 0.3944 0.0032 0.0023 0.0026 0.0012 ---0.0000 —0.0002 ¥ (3 0.2734 0.3706 0.4243 0.0028 0.0017 0.0036 0.0009 —0.0008 --3.0005 F (4) 0.2087 C.1586 0.0020 0.0038 0.0035 0.0028 0.0023 --0.0002 0.0006 F(5) 0.2433 0.5417 0.4413 0.0030 (.0045 (.0072 0.0028 0.0016 —-0.0020 F(6) 0 0 0 0.0176 0.01706 0.0947 0.00D88 0 0 A exagonal coordinates for all atoms in general positions, 18(f), of space group R3, except F(6) and Na(2), which are in 3{a) and 3(b) respectively. and for which the x-ray powder diagrams have been ically by a PDP-5 computer.*’ The structure was reported®’ are the 7:6 compounds of NaF, KF, and solved by an analysis of the Patterson function RbE with UK, and ThF , as well as 7NH I - 6UF‘4.38 computed with these data and was refined by The existence of this curious stoichiometry has least squares. The positional and anisotropic been suggested as being due to the possibility of thermal parameters are given in Table 1.8. an NaZrF _ structure with the presence of sites A portion of the structure is shown in Fig. 1.8. where one extra Nal per six NaZrF_ units could The atoms are represented by ellipsoids showing be accommodated.®?® In order to check this hypoth- their thermal motion. The very large ellipsoid esis and/or ascertain the structural causes of corresponds to the ‘““extra’’ fluorine, which is this formula, a complete crystal-structure deter- indeed contained in a polyhedron of six Zr atoms mination was cairied out. bridged by F atoms. The large motion of this About 900 x-ray intensity data from a small F atom means that either it is rattling about quite spherical single crystal were collected with a freely or else there is a statistical disorder in Picker full-circle goniometer controlled automat- ~ which this atom is bonded randomly to one of the six Zr atoms at a time. The “‘extra’’ Na atom is Na(2); it is closely coordinated by six F neighbors and by six additional F neighbors at a little greater distance. Each Zr atom has eight F atoms about it and each Na(l) has seven F atoms. 37G. D. Brunton et al., Crystallographic Data for Some Metal Fluorides, Chlorides, and Oxides, ORNL- 3761 (February 1965). o 38 R. A. Penneman et al., Inorg. Chem. 3, 309 (1964). Wy R Busing, R. D. Ellison, and H. A. Levy, %0, A. Agron and R. D. Ellison, J. Phys. Chem. 63, Chemistry Div., Ann. Progr. Rept. May 1965, ORNL-~ 2076 (1959). 3832, p. 128. 19 ORNL — DWG 66 ~ 260 Fig. 1.8. A Portion of the NazZr,Fy Structure Showing Probability Ellipsoids of Thermal Motion of the Atoms. 2. Chemical Studies of Molten Salts OXIDE CHEMISTRY OF LiF-BeF -ZrF , MELTS C. F. Baes, ]r. B. F. Hitch In previously reported studies of the oxide chemistry of LiF-BeF 2=~ZrF4 melts, equilibrium quotients were determined by the transpiration method?! for the following reactions: H,0(8) + 2F7(d) == 07 (d) + 2HF (@) , 00 = P/ P07 5 HIO(g) + F7(d) == OH™(d) + HF(g) , Qp= (PHF/PI-I O)[0H~] (D) 2 2 2 H, C(g) + — MF _(d)== - MO_, (s) + 2HF (&), z z 2 - PHF (M?+ = Be?t, Zt**, U Oy /Py o M7127%) - 2 (3) Reactions (1) and (2) were of interest because of their role in the contamination of molten fluorides with oxide, their role in the removal of oxides by HF during salt purification, and, more recently, because of their use in oxide analysis.® Reaction (3) has been of interest because it relates the thermodynamic properties of the dissolved {luoride MF _ to the available thermodynamic data for MO , (), H,0(¢), and HF(g). Estimates of the soluéility product of the oxide MOZ/2 have been obtained from the equilibrium quotients QO and 1A, L. Mathews, C. F. Baes, ]Jr., and B. F. Hitch, Rcactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNI1-3789, pp. 56—65; alsc ref, 7, this section. MsR Program Semiann. Progr. Repé. Sept. 30, 1965, ORNI.-3897 (in press). 20 O uo_,, - (047077 = IM?T[03]2/2 | (4) However, these estimates have been of limited accuracy because of the difficulties in determin- ing QO by the transpiration method, During the past year, a more direct method for determining oxide solubilities has been adopted. A measured volume of salt, presaturated by equilibration with excess solid oxide, is filtered through a sintered nickel filter into a heated re- ceiving vessel, It is then sparged with an H -HF mixture and the total amount of water evolved determined by Karl-Fischer titration, Thus far, essentially complete removal of oxide from the filtered salt samples has been accomplished in 3 hr or less by use of an influent HF pressure of 0.05 atm. The blank has been equivalent to ~2 x 1072 mole/kg of oxide (32 ppm), the sensi- tivity has been ~.5 x 107 * mole/kg (8 ppm), while the experimental values ranged from 1.5 x 1072 to 1.5 x 107 * mole/kg. This method presently is being used to redeter=- miie the solubilities of ZrO_ in (2LiF-BeF2) + ZrE mixtures, which simulate MSRE fuel salt and flush salt mixtures. The results obtained thus far are compared in Fig. 2.1 with previously reported’ estimates based on measured values of the ratio (_)O/Qzr in such salt mixtures. When completed, these measurements should permit the establishe ment of more reliable tolerance limits for oxide levels in the MSRE and in future molten-salt reactors. In addition, the more accurate solubility values - when combined with er values [reaction (3)] — should yield cortespondingly more reliable values of ¢ . These will be useful in optimizing oxide removal procedures during salt purification and salt analysis, 21 0.05 = L > 2 3 002 £ zZ ] 5 om o '.‘ =z LiJ Q & 0.005 (W] wl o3 = t 1 3e = Sm(s) 2,01 0,795 Bel® 4 2e = Be(s) ~1.721 0,715 Th*t 1 4e = Th(s) ~1.73 Zr*t i de = Zr(s) ~1.316 0.755 Uttt de = U(s) 1,319 0.674 U3ty 3e += U(s) 1,410 0.630 vitte = Ut —1.044 0.807 Cr2t s 2e = Cr(s) ~0.390 0.505 Felt= 2e &= Fe(s) —0.011 0.516 NiZt+ 2¢ == Nils) +0,473 0.830 HF(g)+ e = F~ + ,H (&) 0 0 BL@+e =1 —0.343 0.002 1/ng(g) + 2 = 8§77 <0,10 70, @ + 20 = 0% 0.558 0.124 1/Qc,z(,g) + 1/2112(@ + e v OH 1.734 0.472 1/2F2(g) +e = F 2,871 0.044 €Calculated from Aéf values in Table 2.1. bStandard states for ions are defined in footnote a of Table 2.1. VAPOR PRESSURES OF FLUORIDE MELTS species, and to obtain data of importance to the Molten-Salt Reactor FProgram. S. Cantor D.S. Heu'® Rodebush-Dixon'? and boiling point methods were W. T. Ward used to measure total pressure over 16 melts covering the entire composition range. Each melt was measured over at least a 185° temperature Toto!l Pressure Measurements range; the pressure range usually covered 1 to 100 mm Hg. For all cases except one, the linear The vapor pressures of the system LiF.BeF expression are being investigated to derive thermodynamic activities, to determine the significant vapor log p(mm) = A — B/T(°K) (9 18 gummer employee, 1963, from the University of 9. 1. Rodebush and A. L. Dixon, Phys. Rev. California, Berkeley, 26, 851 (1925). Table 2.3. Vapor Pressure in the LiF‘-Ber System Melt Composition Equation: log p(mm) = A — B/T("K) {(mole %) Temperature Range Measured ST o e z 5 190 7791147 10.491 10,953 84.99 15.01 826-1116.5 19.648 11,230 75.01 24.99 890--1112.5 10.255 10,756 70.00 30,00 8431150 10,296 10,879 65.00 35.00 857--1112 9.787 10,266 57.54 42.46 866---1121 8.817 10,449 530,00 50,00 886-1071 S.134 9,738 45.04 54.96 9321155 9.096 9,044 39,95 60.06 930--1224 8.993 16,058 36.00 64,00 9501214 9.279 10,688 30.00 70.00 9661281 B.660 10,138 25.00 75.00 10201272 8.836 10,526 11.00 89,00 891.-1239 6.711 8,175 7.00 93.00 0781234 8.702 11,042 3.00 97.00 1020~1270 10.062 13,178 100 10261272 See below” ?The equation for pure LiF is log p = 3,619 ~ 15,4530/T - 6.039 log T. provided adequate fit to the data (4 and B are constants). The data are summarized in Table 2.3, The constants for each equation were obe tained by the method of least squares, Isotherms at 1000 and 1100°C are shown in Fig, 2.3. To obtain some unotion of the vapor species, vapor was collected in the tubes of the vessel at the conclusion of vapor pressure measurements for several compositions. Chemical analysis showed only traces of lithium ion in condensates where the composition of the melt was 75 or greater mole % Ber. For melts with 70 or less mole % BeF , considerable quantities of lithium were found in the condensates. All that may be concluded from these analyses is (1) at greatet than 75 mole % BeF,, the vapor is virtually pure BEFZ, and (2) at less than 70 mole % Bel’*‘z, the vapor becomes more complicated; the vapor prob- ably contains the compounds LiBeF, and LiQBeF,‘,zo The only activity coefficients that may be calculated from these data alone are those of Be¥ at melt concentrations of 75 mole % or greater in Ber. For these concentrations the activity coefficients of BeF at temperatures of 1000 and 1100°C were found to be greater than unity {varying between 1.02 and 1.10), These values of the activity coefficient are in reasone able accord with results of the H, 0-HF equilibrae tion studies?? of the LlF—BeFZ system, Manometric vapor pressures were also measured for the MSRE fuel solvent (composition: 64.7~ 30.1-5.2 mole % LiFmBeFZ-Zerr) in the tempera- ture range 960 to 1167°C. The data fit the equa- tion log p(mm) = 8,803 - 9936/T(°K). 20 A, Bichler and J. 1. Stauffer, TAEA Symposium on Thermodynamics, July 1855, paper SM-66/26. 1A, L. Mathews and €. ¥. Baes, Jr., ORNL-TM- 1129 (May 7, 1665), Linear ORNL-DWG 65-10170 400 o : | 5 'E“‘_*Efi-hau:_‘::;_ ,,]7)_ — ii bt | 200 - —f— et — — — N ‘ | | \ | 10 - ‘\ —IT T T o— N T T r ° ol TN N ] - o K**———-"-IX\HOO"C - ] - L #Ai_.... e R . 20 F - : — \. - £ | * : \ . . 10 | T 10000(;\3; i ——— ] ! t‘[ qs - JL 0.5 |- 0.2 L ‘ o 20 40 c0 80 100 BeFy mole % LLiF Fig. 2.3. Vapor Pressures in the LiF-BeF, System. 26 extrapolation of the data to 663°C, the highest temperature of MSRE fuel salt at normal power operation, yields a vapor pressure of 0,015 mm, Transpiration Studies Apparatus constructed pressures by the carrier-gas method in order to determine (1) vapor composition in the LiF-Bel system (these measurements will complement the manometric pressure data already obtained for this system), and (2) rare-earth vapor conceatra- tions in equilibrium with liquid mixtures of im~- portance to the molten-salt reactor distillation process (from these concentrations, more accurate decontamination factors for the rare-earth fission products will be obtained). The apparatus, shown schematically in Fig. 2.4, closely resembles that of Sense et al.?? To test the reliability of the apparatus at clevated tempera~ tures and the procedures for removing condensed was to obtain wvapor vapors, several runs were carried out with LiF, Satisfactory agreement with the reliable transpita= tion data obtained by Sense?? attained if argon flow rates were kept below 50 c¢m’/min and was if the condensers were washed for about 12 hr with the solvent (a dilute solution of disodium Ver- senate) at about 70°C., K A, Sense, M. J. Snyder, and J. W. Clegg, J. Phys. Chem. 58, 22324 (1954, 23 A. Sense and R, W, Stone, J. Phys, Chem, 62, 1411 (1958). ORNL—-DWG 65—13115 — MOLECULAR SIEVE DRYING COLUMN — HEATED COPPER TURNINGS \ >} qj VENT FLOW METER /rPQESSURE GAGE (O-— ARGON e @ [—— NICKEL J {5-in. LONG THERMOCOUPLE NO1 {LIQUID SALT TEMPERATURE }— HIGH TEMPERATURE ALUMINA TUBE {36-in LONG) 101In Hzo) THERMOCOUPLE NO. 3- STAINLESS STEEL SLEEVE ) (5Vzin LONi; THERMOCOUPLE / ND. 2. \ y M‘__f ] A : | - 4 | MOLTEN / SALT — S INSULATOR ™ / / f/ MARSHAL FURNACE (16-in LONG)-: ( ARGON TO DRYING / 2 COLUMN; THEN LAVITE ./ S ON TO WET TEST METER . CONDENSER TUBE (CONDENSER HAS 0.021-in. OPENING IN END ABCVE CENTER OF SAMPLE BOAT) Fig. 2.4. Transpiration Apparatus. YISCOSITIES OF MOLTEN FLUORIDES 3. Cantor W, T. Ward Li!':--BeF’2 Viscosity measurements of the LiF-BeF sys- tem were extended to include three more mixtures (36, 40, and 45 mole % BeF ). The viscosity of pure BeF was measured again, this time over a 410° temperature range. As previously reported,?? all measurements were carried out with the Brook- field LLVT viscometer, For the three mixtures, viscosity decreased with decreasing Ber concentration. The degree of decrease with concentration, however, was not very large; below 50 mole % BeF concentration, melts are more typically ionic with relatively low vizcosilies; the clusters of beryllium-fluorine linkages that accounted for the very high viscos- ities at higher BeF, concentrations are greatly diminished when the concentration of BeF is less than 50 mole %. The viscosity of pure BeF, was remeasured to determine if pronounced deviation from Arrhenijus behavior occurs. The plot of log n vs 1/T showed only slight curvature; hence, BeF, is Arrhenius. Usually the physical behavior of BeF is analogous to that of Si0_. Macedo and Litovitz?® postulated that the Arrhenius behavior of 5i0, is due to the constancy of density with temperature which, in turn, means that the free volume of 510 , does not change with temperature. To establish that den- sity constancy with temperature is also why Bel is Arrthenius will be difficult, because density measurements of Bel”_ are a difficult experimental task; nevertheless, we are currently considering methods for measuring the density of pure BeF . The data on the LiF-BeFZ system are summarized in Table 2.4, The constants for the viscosity- temperature equations were obtained by the method of least squares. NGBFA To provide some information for heat transfer calculations on the MSBR secondary coolant, pre- 243, Cantor and W. T. Ward, Reactor Chem. Div. Ann. Progr. Rept. Jan., 31, 1965, ORNL.-3739, p. 81, 5p, B, Macedo and T. A, Litovitz, J. Chem. Phys. 42, 245 (1965), 27 liminary measurements were begun to determine the viscosities of fluoroborates. The first measure= ments were made on NaBF ; the results were: 7 + 2 centipoises at 466°C and 14 1 3 centipoises at 436°C. The precision is poor because the Brookfield viscometer is primarily designed to measure higher viscosities, No mixtures of alkali fluorides with NaBF | were measured because such melts would most likely have been even less viscous than pure NaBF ; the poor precision in this lower viscosity range discouraged us from attempting further measure- ments with the instrumentation at hand. ESTIMATING DENSITIES OF MOLTEN FLUORIDE MIXTURES S, Cantor Several vears ago,2® the author, after examining the published data on density of molten fluorides, proposed that the simple rule of additivity of molar volumes was very useful for estimating densities of fluoride meits., To facilitate calcula- tions, a table of empirical molar volumes was derived from the published data. Since the earlier report, there have been several experimental studies of densities of fluoride melts. In this repott, we reexamine the previous method of esti- mation taking into account the newer measurements. First, it appears that the rule of additivity of mo- lar volumes held for all binary fluoride systems except one; the one exception was the NaF-UF system,?’ where positive deviations from addi- tivity were as great as 6%; the rule held in two studies of the L.iI*"»’I‘hF4 system?®:*? and in the LiF-UF,,”® NaF-ThF,,*® and LiF-KF’? systems. It was also found that the measured densities?® of the eutectic compositions for the KF-NaF, 265. Cantor, Reactor Chem. Div. Aan. Progr. Rept, Jan. 31, 1962, ORNL-3262, pp. 38—41. 27E, A, Brown and B. Porter, U/.S5. Bur. Mines Rept. Invest, 6500 (1964). 28D. G. Hill and S, Cantor, Reactor Chem. Div. Ann. Progr. Rept, Jan. 31, 1963, ORNI.-3417, pp. 4748, e W, Mellors and S. Senderoff, **The Density and Surface Tension of Molten Fluorides,* p. 578 in Pro- ceedings of the First Australian Conference on Elace trochemistry, Pergamcn, New York, 1964; also in Union Carbide Research Summary 1JRS«70, Union Care bide Corp., Parms, Ohio, 28 Table 2.4, Summary of Data and Constants for Viscosity-Temperature Equation log 717 (centipoises) = A/TCK) - B for the System LiF-Ber Composition Temperature Range Viscosity at (mole % BeF ) Measured (°C) 4 o 600°C (centipoises) 36,00 462-600 2,000 1.226 11.6 40.00 441637 2,203 1.367 14.3 45.00 419-638 2,589 1.685 19.1 50.00 376577 3,066 2.073 27% 55,01 389.-584 3,378 2,203 16% 60.00 437584 3,785 2.376 91° 65.00 451724 4,198 2.507 200 70.00 480704 4,695 2.695 480 75.00 490~705 5,362 3.036 1,275 79.99 558745 5,983 3.223 4,260 85.00 539.-747 6,551 3.340 14,600 90.02 564882 7,528 3.766 71,800 91.02 545.-832 7,640 3.702 112,000 93.01 572842 8,198 3,977 258,000 94,91 557837 8,604 4,099 569,000 96.01 601—844 8,928 4,218 1,016,000 97.00 601897 9,587 4.636 2,208,000° 98.01 632--917 10,359 5.179 4,840,000% 99.01 692—-967 11,358 5.816 12,560,000 100 7021112 See below” 63,800,000? aExtrapoIated. bThe equation for pure BeF2 is log # (centipoises) = 14,148/T — 18.345 — 3.382 log T. NaF-LiF, NaF-LiF-CaF , and KF-NaF-LiF sys- tems were all within 1% of densities estimated by the additivity rule (using the molar volumes given in Table 2.5). Two ternary systems, NaF-LiF-ZrF43° and KF-LiF-ZrtF ,?° both studied by the same investigators and by the same method, were each consistently additive in molar volumes. However, the molar volume of ZrF4 that fits one system did not fit the other system (e.g., at 800°C, the molar volume of ZrF‘4 in LiF-KF eutectic 3. LiF-NaF was 55 cm in 3053, W. Mellors and S. Senderoff, Soc. 111, 1355 (1964). J. Electrochem, cutectic, the molar volume was 45 cm’). For density measurements carried out by Sturm®! on four fluoride mixtures, the additivity estimate differed from the experimental results by 5 to 6% for three mixtures; the discrepancy between ex- perimental and estimated density of fluoride mix- tures seldom exceeds 3%. Although errors in the molar volumes of 'E-L"rl:‘4 and KF may be responsible, the reasons for these larger discrepancies are as yet unresolved. 318. J. Sturm and R. E. Thoma, *f*Measurement of Densities of Molten Salts,*®’ this report, Thus, it appears that, although there may be exceptions, the rule of additive molar volumes describes the experimental data on molten fluorides quite well and remains the simplest, most accurate method for estimating densities of fluoride melts. Table 2.5 gives a revised and enlarged set of empirical molar volumes. The molar volumes of AlF, were derived from the studies of Edwards et al.®? on cryolite. Molar volumes of the alkaline- earth fluorides (other than Ber), YFS, and the rare-earth fluorides are based on measurements. of the pure components carried out by Kirshenbaum and co-workers.>%+3* The values for CsF were ob- tained from Yaffe’s measurements.®® The molar volumes of UF in Table 2.5 are based on measure- ment of the pure liquid;®® these volumes, when used additively with those listed for LiF and NaF, provided good agreement between calculated and measured?’ densities in the LiF-UF, and NaF- UF, systems. The previous dual values for molar volumes of UF, (see ref. 26), based on measure- ments of mixtures,?”**® were probably in error because much of the UF, was removed from solu- tion by precipitation of UQ,. {To estimate a density expression of the form d=a-—bt, (10) first solve for two densities by using the equation Y ) i=1 d, LWy ol =1 (11) where Ni and M, are the mole fraction and molecu- lar weight of component v, and Vl,(t) is the molar volume of component i at temperature t. Substitute molar volumes from Table 2.5 at the two different 32]. D. Edwards et al., J. Electrochem. Soc. 100, 508 {1953). 33A. D. Kirshenbaum, J. A. Cahill, and C. S. Stokes, J. Inorg. Nucl, Chem. 15, 297 (1960). A, D. Kirshenbaum and J. A. Cahill, J. Chem. Eng. Data 7, 98 (1962). 351. S. Yaffe, Chem. Div. Semiann., June 20, 1956, ORNL-2159, p. 79. 3% A, D. Kirshenbaum and J. AL Cahnill, J. Inorg. Nucl. Chem. 19, 65 (1961). 37TR, C. Blanke ef al., MLM-1076 (April 1956). 3813 . Blanke of al., MLM-1086 (December 1956). Progr. Rept. Toble 2.5, Empirical Molar Volumes of Fluorides Molar Volume (cm3/mole) At 600°C At 800°C LiF 13.46 14.19 NaF 19.08 20,20 KF 28.1 30.0 RbF 33.9 36.1 CsF 40,2 43.1 BeF 23.6 24,4 Mgk 22.4 23.3 CaF 27.5 28.3 StF 30.4 31.6 BaF 35.8 37.3 ALF 26.9 30.7 YF, 34.6 35.5 Lak 37.7 38.7 CeF 36.3 37.6 PrF 36.6 37.6 SmF 39.0 39.8 ZrF, 47 50 Th¥, 46.6 47.7 UF, 45.5 46,7 temperatures to obtain the two values of d,; since density is linear with temperature, substitution of the two values of dt in Eq. (10) provides the solu- tion for the constants a and b.} ESTIMATING SPECIFIC HEATS AND THERMAL CONDUCTIVITIES OF FUSED FLUORIDES 8. Cantor Specific Heats The simplest rule for estimating high-tempera- ture heat capacities in the condensed states is that of Dulong and Petit, in which the heat capacity per grameatom is approximately equal to 30 6 cal/°K. Accordingly, the experimental heat capacities of molten fluorides were examined in order to modify the Dulonge«Petit value for molten fluorides. The data for pure compounds (see Table 2.6) indicate that the heat capacity per gram-atom is approximately 8 cal/°K. The data for fluoride mixtures would seem to indicate the average value to be somewhat higher than 8, However, the heat capacities of the mixtures are probably uncertain by 10 to 20%, whereas the experimental uncertainty for most of the pure compounds is less than 5%. Hence, 8 cal (°K)™! (gram-atom)™ ! is the value chosen for estimating specific heats of fluoride melts, The temperature variation of C_ in liquids, when determined accurately, is very small; whether this variation is negative or positive has not been adequately established. It is, therefore, prudent to assume that C is temperature independent. [The specific heat is estimated from the follow~ ing expression: 8Y (Wp) cm T (12) 2: (NiMi) where ¢ is the specific heat, p. is the number of atoms in a molecule of component 7, and Ni and Mi are the mole fraction and the molecular weight, respectively, of component 1, Sample calculation: For MSRE coolant, 66-34 mole % LiF-Ber, T p,) = 0.66(2) + 0.34(3) = 2.34 (VM )~ 0.66(26) + 0.34(47) = 33.1, 8(2.34) - —— = 0.57 cal (°K)" ! g7 1.] © 7 33 CK)” e Thermol Conductivities At present, accurate measurement of thermal conductivities of fused fluorides is very difficult; hence, reliable methods for estimating thermal conductivities are extremely useful, especially if one wishes to predict heat transfer rates in circulating systems. Gambill?® has published an empirical method for estimating thermal conductivi- ties which agrees with the available data. This report will describe a new semitheorefical method based on Bridgman’s theoty of energy transport in liquids.*® Bridgman proposed that energy is transferred by collision from one molecular layer to the next at a rate equal to the local sonic velocity and that the distance traveled between collisions is equal to some characteristic molecular distance, DBridg- man obtained the equation for A, the thermal con- ductivity, A, (13) where k is Boltzmann’s constant, g is the velocity of sound, and A is characteristic molecular dis tance between collisions, To evaluate A, Bridg- man substituted the average distance between molecular centers, (V/N)!/3 (V is the molar volume and ¥ is Avogadro’s number), thus deriving the expression A= 3k N) v = V "J_ . Equation (14) is in good agreement with the ex= perimental data on covalent compounds. (14) However, this equation predicts low thermal conductivities for molten fluorides because the characteristic collision distance is too large. The collision distance for molten salts should be less because the ions (the ‘“‘molecular’’ entities in the liquid) are much less compressible than covalent mole~ cules, Rather than attempt to estimate collision distances in fused salts a priori,*! Eq. (14) was empirically adjusted to fit the available data, the equation obtained being: A 4431< (0.693/1‘1/2)WT/Q ) where WT is the total weight of the fuel. Since the half-time for decay of !3° to !*°Xe is 6.7 hr, iodide-removal half-times of the order of an hour might be desired. With Q = 40 kg/mole (2LiF-BeF, at 500°C), half the iodide present in a reactor fuel could be removed in 1 hr by the passage of a minimum of 0.0173 mole of HF (388 std cm?) per hour per kilogram of fuel. REMOVAL OF RARE EARTHS FROM MOLTEN FLUQGRIDES BY EXTRACTION INTD MOLTEN METALS J. H. Shaffer W. K. R. Finnell F. A. Doss W. P. Teichert W. R. Grimes In a two-region molten-salt breeder reactor, the fuel mixture will require routine reprocessing to reduce concentrations of those fission products which have high neutron-capture cross sections. Of the various fission products which form sta- ble chemical compounds in the fluoride fuel rare earths will comprise the major poison fraction, The extraction of selected rare earths from solution in a molten fluoride mixture into immiscible molten metals is being studied as a possible chemical reprocessing Experiments conducted thus far have cerium, mixture, me thod. examined the extraction of lanthanum, neodymium, and europium from the fluoride sol- LiF-BeF, (66-34 mole %), into bismuth metal. This fluoride mixture simulates the fuel solvent proposed for a molten-salt breeder reac- vent, tor; UK, is here presumed to have been removed from the fuel by fluorination. Fluoride mixtures having rare-earth concentra- tions of about 10™% m.f. together with their ap- propriate radioisotopes (for analytical purposes) were prepared in nickel vessels. The mixtures were further treated at 600°C with anhydrous HF and H,, according to established fluoride purification techniques, until the gamma activity in two or more consecutive filtrate samples of Table 3.3. Distribution of Rare Earths Between LiF-BeF, (66-34 Mcle %) and Molten Bismuth when Reduced with Beryllium Metal at 600°C Rare Earth Rare Earth Remaining Dissolved Rare Earth in Salt Phase in Metal Phase (%) (%) Lanthanum 0 41 Cerium 0.1 90 Neodymium 1 49 Europium 3 90 the salt mixture became constant and approached anticipated values. The liquid-metal extractants were prepared in stainless steel extraction ves- sels with low-carbon steel liners; pretreatment with hydrogen at 600°C reduced oxide impurities in the liquid metals. An extraction experiment was started by transferring a portion of the pre- into the Each extraction experiment involved about 2.35 kg of molten bismuth and 1 to 2 kg of salt. The two phases were agitated by sparging helium through a tube that extended into the metal phase. Two types of extraction experiments have been conducted. In the initial experiments, beryllium metal was added as machined turnings to the extraction vessel during preparation of the molten metal. Following introduction of the salt mix- ture, filtered samples of each phase were taken pared salt extraction vessel. at periodic intervals and analyzed radiochemically for their respective rare-earth content. The dis- tributions of rare earths at the conclusion of each summarized in Table 3.3, These results demonstrate the effective removal of rare earths from the salt phase. experiment are The incom- plete dissolution of rare earths in the metal phase may indicate that a third, seolid, phase was formed; it is not unlikely that this insoluble phase rare-earth beryllide such as have been observed by others at this Laboratory,?!3 Spectrographic analyses of samples taken from the metal phase showed the presence of dissolved is a 13 - M. E. Whatley, private communication. lithium in the molten bismuth, suggesting that the beryllium metal had caused reduction of part of the lithium. In two additional experiments, extractions of cerium and neodymium were ac- complished by adding lithium metal directly to the molten bismuth. Samples of the salt and metal phases were withdrawn under assumed equi- librium conditions after each addition of lithium metal. The distribution of cerium and neodymium between the two phases at the conclusion of the experiments essentially duplicated that found in the beryllium-reduction experiments. centration of lithium in the metal phase of each experiment increased linearly with the quantity of lithium added; however, only 1/4 to 1/3 of the added lithium appeared in the metal phase, and it may be that a reduction of beryllium ion to its elemental form accounted for the missing lithium. Furthermore, as shown in Fig. 3.5, the concentration of rare earths in the metal phase in each experiment became independent of the lithium concentration found in the metal phase. Subsequent extraction experiments will be ex- tended to include other rare earths and to c;tudy these inferred reaction equilibria. The coa- ORNL-DWGE 66-988 ° } | t T ‘OO% (,e H&MD\:’AL FR’OM SAL ——— Pt w ; 4 ol ?EI 4 feeeeeeeeees R T — o =2 o CERIUM 1 oo w F = . e b zs ° . | Qb - Z oo 100% Nd REMOVAL FRUM S[\LI g5 ey — e e D T e . L. g 2o s — j e e r £ | ?_._‘ — i < 1l o L) ] OETIRNTTIQ] O e c AT / | . ; NECGCTYMIUM VoM 0 | | _ G 0.2 1.0 1.5 2.0 2.5 LITHIUM FOURND IN METAL PHASE { mole fraction x 107) Fig. 3.5. Extraction of Cerium and Neodymium from LiF'Ber {66-34 Mole %) into Molten Bismuth by Addi- tion of Lithium Metal at 8009 C. 41 REMOVAL OF PROTACTINIUM FROM MOLTEN FLUORIDES BY OXIDE PRECIPITATION J. H. Shaffer . A. Doss W. K. E. Finnell W. P, Teichert W. R. Grimes Since a single-region molten-salt breeder reactor would incorporate the fertile material in the reac- tor fuel mixture, chemical reprocessing schemes for recovering 23U could be made more effective if “%Pa, its precursor, could be removed with- alteration of the relatively large uranium concentration in the fuel mixiure. The precipi- tation of an oxide of protactinium by the delib- erate addition of oxide ion may provide the basis for such a reprocessing method if the simultane- ous precipitation of UO, can be avoided. Previ- ous studies have demonbtratfld the chemical feasibility of oxide precipitation for removing protactinium from a flucride mixture, LiF-BeF - ThF4 (67-18-15 mole %), proposed as the blanket of a two-region molten-salt breeder reactor.!? In the fluoride fue! mixture of the MSRE, sui- ficient ZrF, has been added to accommodate gross oxide contamination without loss of uranium from solution as UO,_. Earlier studies had dem- onstrated that UC, would not precipitate at 700°C from the solvent, LlF Bel', (66-34 mole %) with added UF, and ZcF , unhl the concentration ratio of ZtF, to UF, dropped below about 1.5.'° Therefore, a preliminary study was made ot the precipitation of PaO, from a fluoride mixture known to have Zr0, as the stable oxide phase.!® The f{luoride mixture consisted of LiF-BeF (66-34 mole %) with added Z:F, equivalent to 0.5 mele of zirconium per kilogram of salt mix- ture. About 1 mc of **°Pa was included in the salt prepatation as irradiated ThO_. The mix- ture was pretreated with anhydrous HF and H, to convert oxides to fluorides. ) The deliberate introduction of sclid-phase oxide to the melt was made by adding Zr0O, in small increments. Filtered samples of the salt mix- ture were taken at assumed equilibrium condi- tions after each oxide addition and were analyzed out 14+ H. Shaffer ef al., Nucl. Sci. Eng. 18, 177 (1964). Ppescior Chem. Div. Ann. Progr. Rept. Jan. 31, 1961, ORNL-3127, p. 8. Yo Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, p. 56. 42 The results showed that approximately for 23%Pa by gamma spectrometry. of these analyses 80% of the 2°°Pa activity was removed after the addition of about 67.5 g of Zr0, (equivalent to 0.125 mole per kilogram of salt) to the mixture. labile solid solution with ZrO, or was removed from solu- tion in the salt mixture by surface adsorption on ZrQO,, then its distribution coefficient should have remained constant. The fraction of Pa remaining in the liquid phase could then be ex- If protactinium either formed a pressed as a linear function of added ZtO, by the equation - LW y ZYOZ ? FPa W’salt (10) where D = [Pa]oxide/[Pa]salt’ Fpa = fraction of Pa in salt, and W is the weight of the designated interpretation of the experimental function, phase. An data according to this linear shown ORNL-DWG €66-969 W - _ 1 B ZFOQ Fpg Lo ( ) a WsaLT WHERE £, = FRACTION OF #33pa IN LiQUID a0 2 W“*-‘-‘WEIGHT RATIO OF PHASES SALT & | _ CONC. OF Pa IN SOLID PHASE | ! CONC. OF Pa IN LIQUID PHASE P b . * 3 : - s | ~ i / 4 +— ‘1 o HREp Waarr = 3.5 kg ./ ) 0= 237 RECIPROCAL FRACTION OF Pe REMAINING IN LIQUID PHASE O 10 20 30 40 50 60 70 WEIGHT OF Zr0Q, ADDED AS SOLID PHASE (g) Fig. 3.6. Removal of 233Pa from Solution in LiF-BeF, (66-34 Mole %) with Added ZrF, (0.5 Mole per Kilo- gram of Salt) by Addition of Z:0, ot 600°C. in Fig. 3.6, illustrates the constancy of the distribution coefficient at a calculated value of about 237. The data further suggest that ahout 7 g of the initial ZrO, addition partially dis- solved in the salt phase or was otherwise lost from the reaction mixture. Further experiments will include studies of the effect of ZiO, sur- face arca on protactinium removal from salt mix- tures proposed for a single-region molten-salt breeder reactor. HEMOVAL OF PROTACTINIUM FROM MOLTEN FLUORIDES BY REDUCTION PROCESSES W. K. R. Finnell W. P. Teichert W. K. Grimes J. H. Shaffer F. A. Doss The effective recovery of ?23Pa from a molten- salt breeder reactor will provide more economic production of fissionable #33U by substantially reducing blanket inventory and equipment costs and by improving neutron utilization. Accord- ingly, chemical development efforts supporting the reference-design MSBR are concerned with the removal of protactinium from the blanket mix- ture, LiF—BeFZ-ThF’4 (73-2-25 mole %), by methods which can be feasibly adapted as chemical proc- esses. An experimental program has been ini- tiated to study the reduction of protactinium fluorides from this salt mixture by molten lead or bismuth saturated with thortum metal at about 400°C. It was hoped that protactinium, as Pak in the salt phase, would be reduced to its metal- lic state by thorium metal and could he recovered in the molten lead or bismuth. The primary objective of initial experiments with this program has been the study of protactin- ium removal from the salt phase of the extraction For these experiments sufficient 233Pa was obtained for radiochemical analysis by neu- tron irradiation of a small quantity of ThO,. The simulated blanket mixture was prepared from together with the irradiated ThO,, in nickel equipment. This mixture was treated at 600°C with an HF-H, mixture (1:10 volume ratio) to remove oxide ion and at 700°C system. its components, with H, alone to reduce structural-metal impuri- ties. The metal-phase extractant, lead or bis- muth with added thorium metal, was prepared in the extraction vessel {304l stainless steel with a low-carbon steel liner) by treatment with H, at 600°C. 'The extraction experiments were started by transferring a known gquantity of the prepared salt mixture into the extraction vessgel containing the metal-phase extractant. Filtered samples of the salt phase were taken periodically for radiochemical analysis of dissolved protac- tinium. In each experiment, ?3°Pa was rapidly removed from the salt phase and remained absent from the solution during the approximately 100 ht at 600°C while tests were made. Subsequent hydrofluorination of the extraction system with an HF-H, mixture (1:20 volume ratic) showed that “%3Pa could be rapidly and almost quanti- tatively returned to solution in the salt phase. Typical results of these experiments are shown in Fig. 3.7. In this experiment thorium metal was added after the salt mixture was introduced into the extraction vessel in order to demonstrate the necessity of the reduction teaction. The objective of experiments now in progress iz to examine methods for recovering 2®°Pa from the extraction system, The proposed use .of macro quantities of 2?!Pa may be required to ORNL-UWG 66-970 100 w—-T——-—-—-—-—-—---,‘.______q : ! 80 }—~ ------------------- — - W WEIGHT OF SALT: 6004 w WEIGHT OF LEAD: 900 ¢ 2 | | L T b‘o e b e ) .._.........._.,A:z e e = 2 emeenm < < o o ] & \ : | 5 | < ! o < a0 |- _ [ e ¢ | oy WITHOUT ADDED WITH Gdwt %o N THORIUM & Th IN Pb ADDED 20l ol ol ] s a : o 1= J - 0 Ai,__ VR”;\‘A. . o) 0 20 40 O 20 40 EXTRACTION TIME (hr) Fig. 3.7. Effect of Thorium Metal on the Extraction of 233Pa from LiF-BeF,-ThF, (73-2-25 Mole %) in Salt- Lead System at 603°C. circumvent the anticipated adsorption of the micro guantities of ?°3Pa, currently used, on the walls of the container, or on other inscluble species in the system. Additional studies of the delib- erate precipitation and adsorption of *33Pa on solid, stationary beds such as steel wool will be made for comparative evaluation. SOLUBILITY OF THORIUM IN MOLTEN LE.AB J. H. Shaffer F. A. Doss W. K. K. Finnell W. P. Teichert Current studies of the removal of protactinium from a molten fluoride mixture, which simulates the blanket of the reference design MSBR, have been directed toward the development of a liquid- liquid extraction process. The method proposes that protactinium, as PaF, in a salt phase, can be reduced to its metallic state and extracted into a molten-metal phase. The possible use of a molten mixture of thorium in lead would provide a convenient method for combining: the reducing agent with the metal-phase extractant and for replenishing thorium to the fluoride blanket mixture. The objective of this study has been to establish the solubility of thorium in lead over the temperature range of interest tc this program and to provide a lead-thorium solution of known composition for subsequent protactin- ium-exfraction experiments. : The experimental mixture, contained in low- carbon steel, consisted of approximately 3 kg of lead and 100 g of thorium-metal chips. Values for the solubility of thorinm were obtained by analyses of filtered samples withdrawn from the melt at selected temperatures over the interval 400 to 600°C under assumed equilibrium condi- tions. Samples wete withdrawn during two heat- ing and coeling cycles and submitted for acti- vation and spectrographic analyses. These re- sults, plotted as the logarithm of the solubility vs the reciprocal of the absolute temperature in Fig. 3.8, indicate that the heat of solution of thorium in lead is approximately 19 kcal/mole and that its solubility at 606°C is about 1.85 x 104 m.f. CRNL— OWG 66 - 971 TEMPERATURE (°C} 600 550 500 480 400 30 — - —— T—Ffi* T & | | o . B | | % 2.0 T o BY ACTIVATION ANALYSIS g 0 8Y SPECTROGRAFPHIC ANALYSIS o l QO £ | = 10— N — e — S e o _J_j. . <1___ - = ; T - | ‘ | F - e _ _..._‘r— 1. ; ( O ] o _ i - = \ 9 | — . _ - : Both conductance cells were used for measure- ments on the most dilute (0.001 to 0.02 m) solu- tions. ‘The measurements from the two cells were in agreement to within experimental error (1 to 29). The sesults for 0.01 m Na(Cl solutions are shown in Fig. 6.1, where specific conductances are 1E. U. Franck et al., Reactor Chem. Div. Ann. Progr. Repi. Jan., 31, 1961, ORNL-3127, pp. 5052, _ A, 8. Quist ef al., Reactor Chem. Div. Ann. Progr. Jan. 31, 1962, ORNL-3262, pp. 7375, ¥, Franck et al., Rev. Sci. Instr. 33, 115 (1962). Rupf. 3 63 plotted as a function of temperature at pressures to 4000 bars. The type of behavior shown in Fig. 6.1 is typical for a strong electrolyte, and is similar to the results obtained previously with potassium sul- fate.®'S A 300°C, iofiic mobilities increase rapidly due to the rapid decrease in the viscosity of water. Therefore the conductance of the sodium chloride solutions increases also, However, the dielectric constant of water is also decreasing with increas- ing temperature. This lowering of dielectric con- stant allows ion pair formation fo begin to occur. Apparently, at approximately 300 to 450°C (de- pending on the pressure) the two effects (increas- As the temperature increases from 0 to ing mobility of the ions; increasing association between the ions) begin to offset one another, and the conductance begins to diminish with increasing temperature. It should also be noted that when the temperature of an aqueous solution is in- creased, at constant pressure, the density de- creases so that thete are fewer ions per cubic centimeter, and consequently a smaller conduct- ance. In another study, measurements have been made on 0.01 demal (¥ 0.01 m) potassium chloride solu- tions (this is used as a standard solution for conductance cell constant deferminations at 25°C) to 800°C and 4000 bars. It is thought that this would be the logical solution to use for a reference at elevated temperatures and pressures; thus, as more researchers make conductance measurements at high temperatures and pressures, the results from the different laboratories can be more easily compared. Figure 6.2 shows a comparison of previously ' determined conductances of K,80,, *A. S, Quist et al., Reactor Chem. Div. Ann. Progr. Rept. Jan., 31, 1963, ORNL-3417, pp. 77~82. SA. 8. Ouist ef al., J. Phys. Chem. 67, 2453 (1063 ORNL-DWG 66-974 700 600 - - 4000 bars 500 400 300 200 100 i e e SPECIFIC CONDUCTANCE {ohm™ ! em™1) x 10° 400 600 TEMPERATURE (°C) Fig. 6.1. The Specific Conductance of 0.01 m NoCli from 100 to 800°C ot Pressures to 4000 Bars. Solution ORNL-DWG 66-973 DENSITY (g/cm?) Q.30 1.04 Q.76 0.01 m NoCL . \ ; 400 300 200 100 0.0022 m K550, S ?'”' SPECIFIC CONDUCTANCE (chm™*! o i . 200 400 &00 TEMPERATURE (°C) Fig. 6.2. Comparison of the Specific Conductances of K,50,, KHSO,, H2504, KCl1, and NaCl Solutions as Functions of Temperature. H,SO,, and KHSO, solutions with the new values for NaCl and KCl as a function of temperature at a constant pressure of 4000 bars. The different behavior of the KHSO, and H,SO, solutions as compared to the other solutions?'*~7 is due to changes the first and second dissociation constants of sulfuric acid with temperature and pressure.® ’ Both dissociation constants decrease in ®A. s. Quist ef al., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1964, ORNL-3591, pp. 84—-88. 7A. S. Quist et al., J. Phys. Chem. 69, 2726 (1965). 64 with increasing temperature, and increase with increasing pressure. The difference between the conductances of NaCl and KCl is due to the difference in mobility of the potassium and sodium ions. AQUEQUS SOLUBILITY OF MAGNETITE AT ELEVATED TEMPERATURES F. H. Sweeton R. W. Ray C. F. Baes, Jr. The solubility of magnetite (Fe,0,) in agueous solutions at elevated temperatures is of special interest in pressurized-water reactor systems, in which Fe, 0, is a main component of the corrosion film. It also is of interest to geologists in under- standing the origin of Fe O, deposits.® The flowing system for measuring the solubility that was reported earlier® has been modified in several ways to improve the precision of the data, One important change has been the substitution of 66 g of nonradioactive Fe O, for 2.7 g of radioactive material, thus making possible a longer contact time with the solution. This Fe,0, was prepared by oxidizing a carbony! iron powder with steam at 400 to 500°C. Its final specific surface area was 0.12 m?/g, about twice that of the radio- active material. Water for the tests was purified and equilibrated with H, as before. All batches had conductivities of less than 0.08 micromho/cm at room tempera- tures, and concentrations of dissolved O, below 10 ppb (and usually less than 5 ppb). The purified water, sometimes containing added HCl, was pumped first through a recombiner in which it made contact with platinoum black at 260°C (to catalyze the reaction of the remaining O, with H,), and on through the bed of Fe 0, in a column held at a controlled temperature. The solution was then cooled and passed through a Millipore filter with 0.1-u pores. The recombiner, column, and filter holder were made of gold-plated stainless steel. Platinum tubing was used to connect the units. The equilibrated solution was 8. T. Holser and C. J. Schneer, Geol. Soc. Am. Bull. 72, 36986 (1961). 9F. H. Sweeton, C. F. Baes, Jr., and R. W. Ray, Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNI1.-3789, pp. 120-21, ORMNI. ~DWG 66 --975 20 \ * r \ » o \ \\ . o ° \\ 10 0 n THEDRETICAL.\___ SLOPE _ B3 5 (NO H‘1’D'ROLYSIS)\_.__..,..,....__.-__;._ 5 ° | \ 2 : \ BB [T e e e \ o \ \ 2 o 5o ] \ = 4 | e 150°C \C 5 © 200°C \ g ] 8 & [ ® 2 ® ® % 1 _ ® 5.4 5.6 5.8 6.0 6.2 pH AT TEMPERATURE (CALCULAYED FROM 25°C pH DATA) Fig. 6.3. The Observed Solubility of Fe30, in Aqueous Selutions Presaturoted at 25°C with Hy at 1 atm. passed through flowing conductivity and pH cells and then through beds of cation exchanger to collect the dissolved iron, which was later removed and assayed spectrophotometrically by the phenanthroline method. The results are shown in Fig. 6.3. The pH rmeasured at 25°C has been converted to pH at the aquilibration temperature, using dissociation con- stants of water at 25, 150, and 200°C of 0.01, 2.24, and 5.01-107 1'% (molal units), respectively, at any tem- S o- and assuming no hydrolysis of Fe?® perature. Sixfold changes in the flow rate produced no significant change in the composition of the product solution, indicating that the Fe O, had reached equilibrium with the flowing solution. 65 We have used the upper two groups of data in Fig. 6.3 to calculate the solubility product (in molal and atmospheric units) for the reaction Y% Fe,0,(s) + 2H (soln) + %11,(g) = Fe?*(soln) + 4@1‘120(301"1) : The resulting values of log K at 150 and 200°C are 7.00 and 6.08 with standard deviations of 0.11 and 0.06 respectively. We think the other 150°C points are low due to a slight leakage of 0, into the pH cell to give the reaction a2t ] Fe~" + /402+H20 - I/zFezfis + 2HT, and we are now modifying the pH cell to prevent this. The solubility product estimated for 200°C predicts that the solubility of Fe,0, in pure H,0 should be 1.4 pm; this is in good agreement with our previcusly reported® figure of 1.5 um. We will continue measurements to cover the range of 150 to 260°C with HC! solutions from 0 to 30 pm with the intention of examining the data to see if the dissolved iron is hydrolyzed. SOLUBILITIES OF CALCIUM HYDROXIDE AND SATURATION BEHAVIOR OF CALCIUM HYDROXIDE~CALCIUM CARBONATE MIXTURES IN AQUEOUS 50DIUM NITRATE SOLUTIONS FROM 0.5 TO 350°C L. B. Yeatts, Jr. W. L. Marshall The solubilities of calcium hydroxide and the saturation behavior of calcium hydroxide—calcium carbonate mixtures were determined at tempera- tures from 0.5 to 350°C in aqueous solutions of sodium nitrate from 0 to above 3 m in concen- tration. Commercially available reagent grade calcium hydroxide wand calcium carbonate were digested in boiling deionized water to remove soluble im- purities. The hot solutions were filtered and the recovered product dried at 100 to 110°C. The dry calcium hydroxide was ignited at 1150°C for about i6 hr to convert calcium carbonate impurities to calcium oxide; the carbon dioxide content of the freshly ignited oxide was about 250 ppm. Excess solid calcium hydroxide and mixtures of calcium hydroxide and calcium carbonate were equilibrated with aqueous sodium nitrate, prepared with de- ionized water, using rocking equipment. The containing vessels for the 0.5 and 25°C experi- ments were Pyrex stoppered bottles; for the 50 to 350°C experiments, they were titanium alloy bombs. The solutions were filtered during sample withdrawal at the temperature of the experimeit in order to remove suspended solids. A portion of each sample was treated with excess concen- trated nitric acid, evaporated to dryness at 95 to 100°C, and weighed as calcium and sodium ni- trates. Another portion of each sample solution was used for the determination of calcium ion concentration by a potentiometric titration with standard EDTA solution. tration in the solution was determined by difference from the results of these two analyses. The dissolution of calcium hydroxide was as- sumed to reach the following equilibrium;: The electrolyte concen- Ca(OH),(s) = Ca?*(ag) + 20H(aq) . The thermodynamic solubility product expression for this equilibrium is given by: 3..3 v 748}’i, Kg - m 2+m2 B P oH 2 ca? ) OH where m is molal concentration, y is the activity coefficient, and s is the molal solubility of calcium hydroxide. Converting to logarithms, substituting an extended Debye-Hiickel expression for log y, and rearranging the equation leads to the form: Sit7? log K, - log K2 + 6+ 3BI+ 3CI?, (1 AI'/?% where S = theoretical Debye-Hiickel limiting slope for a given temperature, + 3mCa(OH)2’ 0 1; oy _ KSp = the solubility product constant at I = 0, ! = ionic strength = nzNaN03 A, B, C = constants. Since Ksp - 4s? and Ksop: Hs3, then 511/2 4+ BI+ CI*. log s = log sOyr2_ (1 + AIY/2) 66 The dependence of the calcium hydroxide solu- bility upon the ionic strength function, I'/2/(1 + AI'/?), at various temperatures is shown in Fig. 6.4. These curves represent the best least-squares fit of the experimental data, using the A values listed in the legend. The intercepts for the curves are the log s° The inverse relationship between solubility and temperature and the direct relationship between solubility and ionic strength are clearly seen. The poorest fit of the curves with the data occurs at the higher temperatures, values. where the solubility of calcium hydroxide is suf- ficiently low to make precise analytical measure- ments difficult. By using the values of Kgp determined from the data of Fig. 6.4, assuming that ACp - D + ET(°K), where D and E are con- stants, and then by determining the change in log Kgp as a function of 1/7(°K) with the van’t Hoff expression, the standard quantities were calculated and are presented in Table 6.1. thermodynamic CRNL-DWG 6512010 Ca(OH}, {(molality) 03 05 72144 1%2) 06 Fig. 6.4. Solubility of Ca(OH), in Aqueous NaNO; from 0.5 to 350°C. 67 The saturation behavior of calcium hydroxide— alone over the same range of temperature and calcium carbonate mixtures was identical, for ionic strength. all practical purposes, to that of calcium hydroxide Table 6.1. Standard Thermodynamic Properties of the Equilibrium: Ca(OH),(s) =2 Cq‘2+(aq) + 20H " (ag) Temperature AF® Af"” As® (OC‘.) Kgp (kcal/mole) (kcal /fmole) (cal mole ! deg™ 1) 0 1.32 % 1073 6.10 ~1.26 ~26.9 25 9.61 x 1077 6.84 ~2.97 ~32.9 50 5.81 % 107° 7.74 4,79 ~38.8 100 1.46 x 10~° 9,96 —~8.74 ~50.1 150 2,60 % 1077 12.7 ~13.1 ) 200 3.75 x 1078 16.1 ~17.9 ~71.9 250 4,70 % 1072 10.9 —023.2 ~82.4 ano 5.34 x 10710 24.3 ~28.9 -92.8 350 5,69 x 10711 20,2 -35,0 ~103 7. SURFACE CHEMISTRY OF THOR!A C. H. Secoy Heats of Ilmmersion and Adsorption H. F. Holmes E. L. Fuller, Jr. J. E. Stuckey! The caleorimetric investigation of the interaction of water with the surface of ThO, is continuing, The apparatus and procedure have been de- scribed.?™ 3 Net differential heats of adsorption (AH_) of water on ThO, have been derived from calori- metrically determined heats of immersion (h) of ThO, samples containing known amounts of pre- sorbed water (adsorbed after outgassing at 500°C but prior to the immersion experiments). These studies have been completed for three samples of ThO : A (650°C, 14.7 m?/g); B (800°C, 11.5 m?/g); and E (1600°C, 1.24 m?/g). Data in pa- rentheses refer to the calcining temperature and specific surface area respectively. Initial values of AH_ ranged from -20 to —30 kecal/mole. no case did z.i\Ha decrease to zero with completion of the f{irst adsorbed monolayer. These large exothermic values clearly indicate that the water is in a chemisorbed state, most probably as sut- In !professor of Chemistry, Hendrix College, Conway, Ark. ‘c. H. Secoy, H. F. Holmes, and E. I.. Fuller, Jr., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, pp. 164—69. 3¢. H. Secoy and H. ¥, Holmes, Reactor Chem. Div, Ann. Progr. Rept. Jan. 31, 1963, ORNL-3417, pp. 12429, ‘H. F. Holines and C. H. Secoy, J. Phys, Chem. 69, 151 (1G665). SH. F. Holmes, E. IL.. Fuller, Jr., and C. H. Secoy, to be published in J. Phys. Chem. (February 19686). Interaction of Water 68 with Particulate Solids resulting from hydrolysis At larger coverages addi- face hydroxyl of surface oxide ions. tional contributions to AH_ arise from hydrogen bonding between the surface hydroxyl groups and the subsequently adsorbed layers of water. Sam- ples A and B gave AHa vs coverage curves which are typical of adsorption on a heterogeneous sur- face.® The corresponding curve for sample E exhibhited three distinct regions in which water is adsorbed with a constant AH_. The most plausible explanation for this unusual behavior is the formation of successive immobile homo- geneous monolayers. Such an idealized adsorp- tion is more likely to occur with sample E because of its relatively large crystallite size (>2500 A) as compared to samples A and B (200 A}, The dependence of the heat of immersion on physical properties such as specific surface area, particle size, and crystallite size is a complex 4.5.7 The heats of immersion of a series of low-surface-area ThO, samples are shown in Fig. 7.1. In addition to sample E, these include samples D (1200°C, 2.20 m?/¢); ] (1200°C, 2.96 m?/g); K (1400°C, 1.55 m?*/g); and 1. (1600°C, 0.95 m?/g). Maxi- mum deviation of the experimental points from the curves in Fig. 7.1 is about 2%. sideration of the combined groups and unresolved question. From a con- uncertainty in the surface area and heat measurements, one must conclude that samples E, J, K, and L have iden- This is the first known case of such behavior. It would thus appear that in order to have reproducible idealized surface tical heats of immersion. ®a. C. Zettlemoyer and J. J. Chessick, Advan, Chem. Ser., No. 43, p. 88, American Chemical Society, Wash- ington, D.C., 1964. ’W. H. Wade and N. Hackerman, Advan: Chem. Ser., No. 43, p. 222, American Chemical Society, Washington, D.C., 1964, ORNL-0OWG G6--9786 TOO 600 <2 & 2 - " 1 h 50G 400 OUTGASSIMG TEMPERATURE {°C) Fig. 7.1. Heats of Immersion of Low-Surface-Area Thoria Samples. reactions, one must work with samples having low specific surfaces and a relatively large crys- tallite size (the crystalliie size of these samples ranged from 1500 to >2500 A). This does not, however, the results obtained with sample D. The only physical difference in sam- Samples E, J, K, and L. have a mean particle diameter of about 1.5 p, while the corresponding value for sample D is 3.0 y. On this basis it appears that the number of crystallites per particle is an important account for ple D is the particle size. factor. Conceivably, this could influence the number and type of crystal faces exposed for surface reactions. Further experiments are re- quired to resolve this important point, Water Vapor Adsorption and Desorption €. L. Fuller, Jr. H. F'. Holmes The complex nature of water adsorption on tho- rium oxide has been studied by use of a sensitive microbalance. High-temperature sintering (1200°C) appears to produce a material which predominantly presents the 100 cubic face in the surface, upon which there are three distinct modes of adsorption. There is an initial rapid chemisorption, formiung surface hydroxyl groups which are slowly hydrated. In addition and as a precursor for surface hydra- tion, physical adsorption occurs, constructed at 25.00°C reveal the type II physical adsorption isotherms character- istic of polar adsorbates. In addition, the slow hydrating process occurs as a perturbing effect: each successive isotherm fails to close upon the preceding one by a decreasing amount until, after six months of repetitive adsorption and de- reproducible Isotherms sorption, a isotherm is achieved. The rate of irreversible binding is a complex function of both the water vapor pressure and the amount of water previously bound. The rate is increased markedly with the first increments of pressure, but higher pressure less accelerating effects. The rate diminishes appreciably as more water is bhound under the physically adsorbed water. The final vacuum weight of adsorbed water corresponds exactly to the aforementioned stoichiometry. increments have The 1000°C calcination decreases the number of pores but does not alter the distribution (there are siill pores of radii near 10 A); whereas the small pores anneal out at the 1200°C calcination, leaving a minimum radius of 30 A. The fact that thorium oxide must be heated to 1000°C in vacuo to remove all water indicates that there are two different types of sintering involved. The amount of bound water is stoichiometrically equivalent to that required to form the surface analog of a hydrated bulk hydroxide. In addition, this is the amount of water required to build up one completed face-centered cubic lattice unit on the surface above the 100 Th()é plane, with the hydroxide and water oxygen occupying the image pesitions of the substrate oxide ions. Thus this bound water may be similar to the cubic form of ice, where the O—Q spacing is nearly equal to that present in bulk Th()z. The shape of the nitrogen isotherms at --195°C changes, showing that the partial pressure at which monolayer (BET) coverage occurs increases from below 0.05 to 0.13 as more water is: pre- adsorbed. This is undoubtedly due to the fact that the water-covered surface is less energetic and has less affinity for the nitrogen. The extremely low vapor pressure of the initially adsorbed waler and the complex kinetics of hydra- tion must be considered before evaluating thermo- dynamic properties or specific surface areas from water isotherms. One must be equally cauntious when attempting to predict the amount of adsorp- tion from these isotherms. TRANSMISSION —a= AMBENT S b 2800 200C 1800 — —_l 4000 3600 3200 70 ORNL-DWG G5 -10810 7 1800 FREQUENCY (o ) Fig. 7.2. Infrared Spectra of ThO, (Sample A) in Vacuo After 24 hr Outgassing at Indicated Temperatures. Ambient spectrum obtained in laboratory envirenment at a relative humidity of 30%. Infrared Spectra of Adsorbed Species on Thoria C. S. Shoup, ]Jr. Exploratory infrared spectra of a thin, self-sup- porting pressed disk® of thorium oxide (sample A)? were recorded in vacuo (107 torr) as a func- tion of pretreatment conditions. The disk was held in a nickel sample holder within an infrared cell and, by means of an external magnet, could be raised to a section of the cell away from the silver chloride windows and heated as high as 500°C. After pretreatment at a known temper- ature for at least 24 hr, the cell was sealed off under a vacuum, the sample cooled and lowered to the optical section, and the cell placed in the spectrophotometer.® All spectra were obtained at about 35°C, and the sample was not exposed to the atmosphere between measurements. Several features of interest are shown in the infrared spectra of Fig. 7.2, The most obvious 8¢c. 4. Secoy and C. 8. Shoup, Jr., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, pp. 172-73. IPerkin-Elmer model 521 infrared spectrophotometer provided through the courtesy of the Chemistry Division of ORNL. feature is the general decrease in intensity of the various absorption bands with increasing out- gassing temperature, particularly between 200 and 500°C. In addition, several of the absorption bands are displaced in frequency, and in some new bands appear with increasing out- gassing temperature. This indicative of the formation of structurally different adsorbed species as some of the previously adsorbed species are desorbed. The broad bands around 3500 cw~! and less, due to perturbed O—H stretching modes of vibra- tion, show that a considerable amount of hydrogen- bonded adsorbed water remains on the surface after evacuating at 200°C. Even after outgassing at 500°C, the thoria is not free of surface O—H groups, as shown by the rather sharp bands above 3600 cm™!'. Spectra obtained with an improved signal-to-noise ratio display three sharp bands at 3715, 3650, and 3520 cm ' when the quantity of adsorbed water corresponds to approximately The relative intensities cases 13 one monolayer or less. of these bands and their changes under a variety of conditions are such as to indicate the presence of three different types of surface O—H groups at low coverage. Some contamination of the surface is evident from the bands due to C—H stretching vibrations in the 2850 to 2950 cm™! region. This hydro- carbon-like material is easily removed by suffi- cient heat or by treating briefly with oxygen at high temperatures and has only minor effects on the rest of the spectrum, A prominent absorption band at 1630 cm™ in the spectrum of the thoria—water vapor interface in the presence of waler vapor at a partial pressure of about 14 torr. Pumping at toom temperature was sufficient to remove this band, however, confirming it to be due to the H—O—H bending motion of physically adsorbed hydrogen-bonded water molecules. A series of in vacuo infrared spectra were oh- tained after the 500°C-out-gassed thoria had been exposed to water vapor for periods of 24 hr to as long as 11 days each before subsequent evacu- ation at room temperature. Each successive spectrum revealed the presence of a greater quan- tily of adsorbed water than the previous spectrum, “irreversible’ adsorp- tion of water that had been originally detected b was revealed thus confirming the slow gravimetrically. 10 Electrokinetic Phenomena at the Thorium Oxide—Aquecus Solution Interface!! . S. Shoup, Jr. H. F. Holmes Electrokinetic effects of agqueous solutions in porous plugs of thorium oxide have been investi- pated from the standpoint of irreversible thermo- dynamics.'® A kinetic electroosmotic technique was used to determine the electrokinetic poten- tial. ' This method gave results in agreement with streaming potential data except in the pres- ence of acidic solutions, In general, the electro- kinetic zeta potential was observed to decrease with incteasing pH, varying from about +25 mv to —55 mv with an isoelectric point near pH = 9.4. 10-. H. Secoy, E. L. Fulier, Reactor Chem. Div. Ann. ORNL-3789, pp. 16972, Y4, F. Holmes, C. S. Shoup, J. Phys. Chem. 69, 3148 (1965). 125 R, deGroot, Themodynamics of [rreversible Prr)(— esses, Inferscience, New York, 1951. 3¢, H. Secoy and H. F. Holmes, Chem. Tech. Div. Ann. Progr. Rept. Aug. 31, 1959, ORNL-Z?’SB, pPp. 85—86. Jr., and H. Progr., Rept. Jan. F. Holmes, 31, 1965, er., and C. H. Secoy, 71 ORM DWG p5-2979R HaO AND D NH@GH IN POROUS £1LUGS - v M0 AND KCI IN GAPILLARIES No.?ro3 IN POROUS PLUGS Fig. 7.3. Specific Surface Conductivity as a Function of Electrolyte Specific Conductivity. An investigation of ThO, plugs from a variety however, indicates that the electro- influenced by the of sources, kinetic potential is strongly method of preparation of the thoria. one batch of thoria calcined at 1600°C gave no indication of a positive zeta potential over the pH range of 5.9 to 12. The complex behavior of the electrical double layer has been observed in other oxide~aqueous sclution systems!*=1% and appears to be highly dependent on the chemical and thermal history of the sample. Until this dependence can be put on a concrete basis, zeta- For example, potential mesasurements for such systems must be assumed to be valid only for the specific sam- ples investigated. | Measutements of the electrical conductance of aqueous solutions flowing through porous plugs of ThO, were correlated with measurements using ceramic caplllarles* with known geometries.'” This made it possible to measure the cell constant associated with the surface conductance®! and thus to determine the specific surface conduc- tivity for several electrolytes. The variation of the specific surface conductivity, A, with the specific conductivity of the bulk solution, A,, is illustrated in Fig. 7.3 for a variety of porous plugs and capillaries of thorium oxide. Hp, Chem. 1 SD. Trans. J. O’Connor and A. S. Buchanan, Australian J. 6, 278 (1953). J. O°Connor, P. G. Jobansen, and A.S. Buchanan, Faraday Soc, 52, 229 (1956). 100, Street, Australian J. Chem. 17, 8§28 (1964). 17¢. H. Secoy and C. 8. Bhoup, Jr., Reacfor Chem, Div., Ann. Progr. Rept. Jan. 31, 1964, ORNL-3591, “pp. 1089, An important feature of the results shown in Fig. 7.3 is the fact that they are some two to three orders of magnitude larger than would be predicted from the classical theories of surface conductance.!® In addition, there seems to be no obvious relationship between the surface con- ductivity and the electrokinetic potential as pre- dicted by these theories. Classically, the expo- nential dependence of surface conductivity on the electrokinetic potential is based on the pres- ence of excess ions in the double layer due to the potential difference. In the present system, however, it appears that the contribution of the excess ions in the double layer is masked by a second mechanism which is much larger in magni- tude. It is believed that this second mechanism is the ionization and subsequent conductance of surface hydroxyl groups. This mechanism is further indi- cated by the relatively slight dependence of the surface conductivity on the bulk ionic concentra- tion, implying that the concentration of the sur- face conducting species is also only slightly dependent on the bulk ionic concentration. If the major mechanism for surface conductance is ionization of surface hydroxyl groups, it should be dependent on pH, as shown by the consistently high values obtained with NH OH. In the case of basic solutions of Na,CO,, however, the slight but consistent reduction in surface conductivity may be due to strong specific adsorption of carbonate ions. GAS EVOLUTION FROM SOL-GEL URANIUM-THORIUM OXIDE FUELS D. N. Hess B. A. Soldano Thoria-3% UO, sol-gel material prepared for use as a reactor fuel has been found to evolve gases when heated or when in reactor service. Such gases could develop excessive pressure within fuel elements or could react with the fuel element cladding to weaken it or to affect its Previous study of gas of this heat transfer capability. evolution from representative samples 181 Th. G. Overbeek in Colloid Science (ed. by H. R. Kruyt), wvol. 1, chap. V, Elsevier, New York, 1952, material !® had shown that particle size, or sur- face area, was an important parameter, with respect to both the amount and composition of the gas evolved, and that CO, was reversibly absorbed at temperatures below 500°C and easily desorbed at higher temperatures. Studies during the past year have been directed toward the development of treatments which would reduce the surface area of the prepared sol-ge!l and thus reduce the amount of gas which it could release in service. Batch exposures at temper- atures of 1000°C to CO, or H,0 alone did not give any substantial change in the surface area of the sol-gel or in its capacity for reversible adsorption of COQ. On the other hand, mixtures of the two gases, CO, and H,O, at 1000°C were found to give approximately 50% decreases in the surface area and absorption capacity of the sol-ge!l for each treatment. The duration of the treatment did not appear to be important. Thus, two treatments would reduce the surface area to 25% of the original — three to 12.5% of the orig- inal. Tests in which the sol-gel pellets were ground to successively smaller particle sizes between successive batch treatments with mix- tures of H,O and CO, at 1000°C suggested that the effect of the treatment permeated the entire material and was not localized at the superficial surface exposed by grinding. Table 7.1 illus- trates these results; a threefold reduction BET surface area was accompanied by a fourfold reduction the particle in in CO, adsorption capacity, although reduction would have, in an untreated sol-gel, caused an eightlold increase size in both surface area and gas evolution with an accompanying increase in CO2 adsorption capacity. Since batch treatments are not convenient for larger-scale concepts and since single batch treatments, however extended did not give more than 50% reduction in area, the effect of treatment in a flowing stream of H,0 + CO, at 1000°C was studied. As shown in Fig. 7.4, the flowing-stream technique was much more effective; the capacity of the sol-gel for gas adsorption appeared to be a predictable function of treatment time, regardless of whether the treatments were continuous or interrupted. A 16-hr treatment ap- peared appropriate for a 90% in time, reduction in the 9H. N. Hess, W. T. Rainey, and B. A. Soldano, Reactor Chem. Div., Ann. Progr. Rept. Jan. 31, 1965, ORNI.-3789, p. 177. original gas-adsorption capacity of the material. Corresponding reductions in the BET surface area of the material were obtained. The linearity of the semilogarithmic plot in Fig. 7.4 sugpgests that there may be a simple first-order process that is responsible for the surface area reduction, but the detailed mechanisms of the chemical reac- tions involved have not been elucidated. A new series of laboratory samples of sol-gel material, prepared with various furnace atmos- phetes during heating, calcination, and cooling and designated the PL series, was received and Table 7.1. Effect of Treatment with HzO«CO;Z on Sol-Gel Properties BET Reversible CD2 Mesh Size Sequence of Surface Absorptive of Tests and Avea Capacity Particles Treatment (mz/'g) (std co/e) 40=8G Original material: 2.57 0.20 depassed at 1000°C 4080 After first surface 0.10 area reduction 140 Ground and sized 1.58 0.10 to 80--140 mesh (otherwise untreated) 80-~-140 After second sur- D.08 face area reduction 270 round and sized 1.67 Q.06 to 200-270 mesh (otherwise untreated) 200270 After third surface 0.78 0.05 area reduction 73 evaluated with respect to gas evolution and revers- ible gas adsorption. All PL samples gave much less gas evolution upon being ground io the as- received condition and heated to 1000°C in vacuo; about 0.30 std cc/g of gas was evolved, compared with 1.30 for the H-II samples previously studied. An even more striking reduction was noted in the capacity of the FL series for reversible CO, adsorption; values of 0.02 std cc/g were obtained, compared with 0.27 for the H-II material. Thus the new material, if its preparation proves to be reproducible, should be much superior from the point of view of gas evolution and might not re- quire any treatment fo reduce its surface area. ORNL-DWG B6-977 0.5 e ; e g } o e MULTIPLE, CUMULATIVE EXPOSURES = r | & SINGLE EXPOSURES - ~ £ ] > t: [ Q b e % M NG e A N Q\\ “ —_ _ - L SN e - g 0.05 L _i\ N\ oeim e o SO S ——— \\. SN ol | ' E{-_’ 0.02 —t e - Lé \\ u T 00 L- | -4 — A& RN S L T s i R J - 0.005 ——-—Jr— e 0 4 8 12 16 20 24 28 TREATMENT TiME (hr) Fig. 7.4. Effect of Flowing COy and H, 0 Mixture at 1000°C on the Reversible CO, Capacity of ThO;-3% U0, Sol-Gel. Part il Gas-Cooled Reactors 8. Diffusion Processes TRANSPORT PROPERTIES OF GASES Rotational Relaxation Numbers for Nitrogen and Catbon Dioxide Thermal Transpiration. A. P. Malinauskas It is well known that serious errors in pressure measurement can result if the manometer is maintained at a temperature other than that of the region whose pressure is of interest. The phenomenon responsible for the discrepancy is known as thermal transpiration (or the “thermo- molecular pressure effect®’) and is defined as the transport of mass as the result of a temperature gradient. Unlike thermal diffusion, however, which is similarly defined, the thermal transpiration phe- nomenon is not restricted to distinguishable par- ticles. In fact, ever since its discovery in 1879 by Reynolds,! investigations of the effect have been conducted exclusively with pure gases. Unfortunately, the utility of the phenomenon was thought to be solely that of a pressure correction; as a result, the scientific community appears to have been content with the empirical and semi- empirical formulations which were proposed and shown experimentally to be qualitatively correct at best. Recently, however, Mason and co- workers,?*? in an attempt to describe gas trans- port in the region where free-molecule and hy- 3 1o, Reynolds, Phil. Trans. Roy. Soc. London, Ser, B 170, 727 (1880); paper No. 33, Scientific Papers, pp- 257.~390, The University Press, Cambridge. R, B. Evans III, G. M. Watson, and E. A. Mason, J. Chem. Phys. 35, 2076 (1961) and 36, 1894 (1962); E. A. Mason and A. P. Malinauskas, f. Chem. Phys. 39, 5272 (1963). g, A. Mason, G. M. Watson, and R. B. Evans III, J. Chem. Phys. 38, 1808 (1963); E. A. Mason, J. Chem. Phys. 39, 522 (1963). 77 drodynamic mechanisms compete, were led to reinvestigate the thermal transpiration effect and uncovered a relation by which it appears possible to obtain information regarding inelastic mo- lecular collisions from thermal transpiration data.? The first real test of the theories involved have been completed in this laboratory during the past year.” Thermal transpiration studies were con- ducted with the gases Ar, Xe, N, and C'Oz‘ The phenomenon was generated by maintaining temperatures of about 290°K and 545%K on the opposite ends of fourteen O.l-mm-ID Pyrex glass capillaries arranged in parailel. Although several apparently minor discrepancies between theory and experiment were noted, the utility of the method for studies of inelastic col- lisions has been demonstrated to be extremely promiging. For example, the rotational collision numbers for N, and CO, which were obtained from the experimental data were found to be 4.4 and 2.4, respectively, and are in excellent agreement with reported values which were determined by the more conventional, elaborate, methods. * and more Gaseous Diffusion in Noble Gas Systems A. P. Malinauskas Unlike related transport properties, diffusion, the transport of mass or identity as the result of a concentration gradient, is totally govemed by unlike-molecule interactions, that is, like-molecnle collisions affect the mecha~ nism only as small perturbations. However, the relative insensitivity of diffusion to the choice gaseous almost ‘A, P Malinauskas, to be published in The [fournal of Chemical Physicsa. of the intermolecular potential characteristic of the interactions necessitates measurements either over a wide range of temperature or of an accuracy far better than that currently possible. Of the two alternatives, the former appears to be the more feasible. Indeed, some diffusion measure- ments have been reported at temperatures as high as 1100°K without a corresponding loss in accuracy,5 but the experimental requirements are so stringent and the procedure ig so elaborate that a more simple approach is desirable. Such an approach appears to have been provided by theory, in that it is possible to relate the com- position dependence of viscosity to the diffusion coefficient characteristic of a given gas pair.® Thus, because viscosity coefficients at high (or low) temperatures can be readily determined experimentally, the evaluation of diffusion data via the theoretical relationship is worthy of in- vestigation. Although the objectives of the dif=- fusion program at the Laboratory are primarily oriented toward an elucidation of intermolecular interactions, the initial studies have been con- with an affirmation of diffusion interrelationship. cerned the viscosity- The first series of investigations in this regard were made with the systems He=Ar, He-Xe, and Ar-Xe and have resolved a previous discrepancy between theory The second phase of the program involves the systems He-Kr, Ar-Kr, and Xe-Kr. Analysis of the latter data has not yet been completed; however, the preliminary results tend to further validate the theoretical relation- ship between the viscosity and diffusion phe- nomena. and experiment.’ Gaseous Diffusion in Porous Media A. P. Malinauskas E. A. Mason® R. B. Evans III One of the major developments regarding the migration of gases within a porous medium has been the formulation of the ‘‘dusty-gas’’ model. SR. E. Walker and A. A. Westenberg, J. Chem. Phys. 31, 519 (1959), 3. Weissman and E. A. Mason, J. Chem. Phys. 37, 1289 (1962). 7A. P, Malinauskas, J. Chem. Phys. 42, 156 (1965). 8Consultant, University of Maryland, Institute for Molecular Physics. 78 This model, which is particularly useful for a description of gas transport in the theoretically difficult and hydrodynamic mechanisms are of equal im- portance, utilizes the somewhat unorthodox view that the solid surfaces of a porous medium or of the walls of capillaries may be mathematically re- garded as agglomerates of giant gas molecules (dust). As a result, a gas-surface interaction is treated as a special type of molecular encounter as would occur between two gas molecules when one of the interacting molecules is immobile and preponderantly larger and heavier than the other. Use of the mode!l has yielded the first con- sistent transition region where f{ree-molecule treatments of gas transport in porous septa for the following cases:?:? (1) diffusion due to a composition gradient; (2) gas flow as the result of gradients in composition and pres- sure; (3) thermal transpiration, in which pressure and temperature gradients are involved; and (4) migration (and separation) of gases under the combined influence of composition and tempera- ture gradients. However, our inability, at the time, to cope with the combination of diffusive and viscous modes of transport in a manner which was satisfactory from a theoretical viewpoint necessitated the introduction to a certain degree of empirical methods in those cases in which pressure gradients were at least partly respon- sible for the particular transport phenomenon. Somewhat erroneously, perhaps, we sought a clarification of this so-called ‘‘forced-flow”’ problem in higher kinetic theory approximations. The approach was fruitful in an indirect manner in that it led to the discovery of the proper procedure of combining diffusive and viscous flows, and prebably to a complete solution of the gas transpoit problem. Stated briefly, one can separate (mentally) motion in a gas mixture into a diffusive flux and a viscous flux. What we have learned, however, is that the total flux is simply the zum of these two fluxes; that is, there is no direct interaction between viscous and diffusive flows; the only ““cross term’’ of consequence can be absorbed into a change in the pressure-diffusion coeffi- cient.? Although this simple additivity relation had been used previously by others, its theoretical °S. Chapman and T. G. Cowling, Proc. Roy. Soc. London, 8er. A 179, 159 (1941); V. Zhdanov, Yu. Kagan, and A. Sazykin, Soviet Phys. JETP (English Transl,) 15, 596 (1962). ® appears to have been unnoticed. justification To date, the theoretical reinvestigation of dif- fusion in the presence of a pressure gradient has been the most fruitful aspect of this work; a semiempirical equation which had been derived earlier has now been developed completely from theory, with an explict expression for what had been an empirical (and adjustable) coefficient. Moreover, the new results lead directly to the prediction of the Kramers-Kistemaker'? or Kirken~ dall'! effect, in which a pressure drop is gen- erated when the net flux in a diffusing mixture is zero. ' SOLID-STATE TRANSPORT PROCESSES IN GRAPHITIC SYSTEMS Recoil Phenomena K. B. Evans IlI J. L. Rutherford R. B. Perez? Utilizatien of oxide fuel kernels and multilayer . pyrocarbon coatings enhances the efficiency of coated fuel particles. Tendencies for fuel mi-~ gration, which are greatest during coaling op- erations, are reduced by the use of oxide!® fuels. Spearhead propagation is minimized'® by applying a porous inner coating followed by an outer coating with good retention properties. It is not unlikely that the success of the multi- layer be partially attributed to sacrificial absorption by the inner coating of recoiled fission fragments and knocked-on fuel concept can atoms. For maximum absorption, ratios of inner-coating thickness to recoil range should be greater than unify. Our objective is the specification of this 1043, A. Kramers and J. Kistemaker, Physica 10, £99 (1943). By, p. McCarty and E. A, Mason, Phys. Fluids 3, 908 (1960). IzConsultant, University of Florida. R B. HEvans I, J. O. Stiegler, and J. Truitt, Actinide Diffusion in Pyrocarbons and Graphite, QOENL-3711 (December 1964). 4R L. Hamner, GCR Program Semiann. Progr. Rept. Sept, 30, 1965 (in press). 130, Sisman et al., GCR Program Semiann. Progr. Rept. Sept. 30, 1965 (in press). 79 ratio for various carbonaceous materials. Thus we are engaged in the determination of the dis- tribution of fission fragments that have recoiled to average positions within pyrocarbon specimens. From the curves related to the distribution we can extract values of the range R and a distribution parameter associated with the recoil range, called the straggling factor a. Although special low-density pyrocarbon specimens have been prepared for experiments to be conducted in the near future, present experiments have been re- stricted to General Electric pyrocarbon speci- mens. To initiate a series of experiments, pyrocarbon specimens are bombarded with 2?3U% at 40 kev to form fission-fragment sources. specimens, accompanied by an uncontaminated pyrocarbon target specimen, thermal-neutron fluxes. incrementally The source are exposed to All specimens are then and assayed by gamma counting so that integral curves related to the distribution of fragments can be con- structed. Applicable transport equations predict that plots of F.R., the total activity remaining in a specimen after grinding to a penetration z, vs z can be approximated by a straight line over most of the interval between zero and E. The only contribufion of « ig reflected in the small tail beyond /R = 0.95. Extrapolations to the z value corresponding to F.R. = 0 will indicate the true range value. Some resulis appear in Fig. 8.1. Several blank experiments were performed using 2327, to establish the as-deposited condition of the actinide on source specimens and to provide a z = 0 correction for the range dala. DBased on the tlank data, we conclude that the source layer is ground various very thin (an important requirement for data correlation) and that the average layer position is 0.8 u beneath the source-specimen surface. As demonstrated by the recoil data in Fig. 8.1, the range value for the light fragments (13.0 u) is greater than the range value for the heavy fragments (11.0 p); the ratio of the range values for the two groups is 1.18. conducted 1o order to Similar experiments in air’'” reveal a smatic of 1.32. 1In egtimate a value of the straggling- factor/range-value ratio, results for three of our best range experiments have been replotted in 16, D. Evans, The Afomic Nucleus, p. 668, McGraw- Hill, New York, 1955, CRNL-DWG 65-12434 1O e RANGE VALUES (u) EXP A-1 EXP. A-2 09 N (CLOSED PTS.) (CPEN PTS.) ' “O. 104 1.5 }11 o v s %oe 2 10.7 H % v 27 12.6 13.5 } 0813 ' ’ 13. X oy 125 3.8 3.0p = | — x i g 07| o . g . g 06 |y - Ny ! : : v \ & > X ™ RANGE DATA - x 4 o N '3 : ALL CURVES < j . 1 REFERRED TO g 04— NN (meos G % ‘ 1 \‘ i — X . WO L & © o3 7 BN “ X A Q £ \ § X * £ 0%y SLANK DATA // % OfF- % X ;J( A.\ D DV % \sl o A, x “m a v O 0 ’f.:‘_’f_fl_‘_y.’});&":effixm!xwx_..__ xm........f,\l 73:&\1};?’? C 0.2 04 06 08 10 12 t.4 Ir Fig. 8.1. Comparison of Range Measurements for Average Light and Heavy Fission Fragments from 235U in General Electric Pyrographite. Data for five blank experiments are also shown. terms of concentration; that is, (A activity/Az) Correlation of these plots with applicable equations suggests an /R value of 0.126. We note that the « and R values cited are automati- cally averaged with vs Z. respect to the and directions by the conditions of our experi- ments, in which fragments enter the target at all solid angles from 0 to 27 steradians. This point has been verified experimentally using and < ¢ > direction specimens. Actinide Diffusion R. B. Evans III J. L. Rutherford F. L. Carlsen, Jr.'7 Investigations of uranium and thorium diffusion in graphite matrices have been carried out on 17F0rmerly with ORNL Metals and Ceramics Di- vision; present address Stellite Division, Union Car- bide Corporation, Kokomao, Ind. 80 High Temperature Materials pyrocarbon (IITM- PyC) and General Electric pyrocarbon (GE- PyC). Although superficial examinations suggest similar structures and properties for these ma- terials,!® we find a distinguishing feature of the GE-PyC to be continuity of structure across basal planes. The integrity of this structure is retained at all temperatures up to 2400°C without delamination from crystalline rearrangement and growth.!? Under reasonable conditions of tem- perature and low actinide concentrations, ac- quisition of reliable direction results {(never obtained with HTM-PyC) is assured. This and the fact that GE-PyC has been studied intensely elsewhere 20 constitute prime reasons for our in- terest in this material. Present results evolved from two distinct ex- periments in which actinide invaded pyrocarbon from either a thin layer at low concentrations or a constant-potential source at the maximum con- centration C .- the apparent solubility. In each case, determinations of D, and Di (coefficients for diffusion in the and directions) were attempted. Estimates of C_ ~ could be ex- tracted only from constant-potential results since surface concentrations in thin-layer experiments diminish with time. Some average results obh- tained at comparable diffusion times are pre- sented in Table 8.1. Clearly, uranium diffuses faster than thorium — particularly in the direction. Additional thorium data reveal a DL<+C>/ D, <--c> ratio of 1.5 at 2065°C. This may be a valuable clue for an interpretation of the well- 20,21 direction diffusion From both thoriuim and uranium diffusion data (not shown), we found the coefficients to be invariant with time over the x/2+/Dt ranges investigated. known nonuniformn patterns, temperature and In other words, 18B5th G.E. stratic-columnar and HTM pyrocarbons possess turbo- {sometimes called granular) strucs tures which lead to high matrix densities (™2.2 g/cn12) and abnormally high anisotropic ratios for many proper- ties., Vg, 1. Carlsen, Jr., ‘*Effects of Pyrolytic-Graphite Structure on Diffusion of Thorium,"” thesis submitted at the University of Tennessee, issued as ORNL-TM- 1080 (June 1965). 20_]. R. Wolfe, D. R. McKenzie, and R. J. Borg, The Diffusion of Non DII D_L D” Temperature, e 2065 2.1 % 10770 1.5 x 1078 3.0x 10710 2.2 x 1077 1865 2.0 x 10711 1.2x1077? 4.3 1071 1.4%x 1078 1697 1.0x 10712 6.8 x 10741 4.6 x 10772 1.6 x 107° Related Results [cl?, g/cm? 3.1 < 1074 2.8 x 10”4 2.1x 107% 5.0 x 10™7 AE®, cal/mole 1.2 = 10° 1.6 % 10° 1.0 x 10° 1.2 % 10° AES, cal/mole 1.1 % 10° 1.4 % 10° 1.3 % 10° 1.2 x 10° A verage actinide concentration, Q, (rsz‘)"l/z bActivation energy for present results. exXp '-—(n’)"l; Gy = 2 pglem”. 3 “Activation energy reported by J. R. Wolfe eof al.,, UCRL~7324 (1964). ORNL-DWG 635-12433 TEMPERATURE (°C) 2400 2000 8GO GO0 1400 1300 T """“‘T‘""""""“”‘[]““"‘“"]’"'T""—"‘“‘ s ORNL om0 [JORI, . N | 3 e e N 2 oy e ~ N o X £ \ o o O o 2 5.0 10,000/, o) Fig. 8.2. Comparison of Uranium Diffusion Coeffi- cients Reported for General Electric Pyrogrophite (Columnar). the coefficients are true constants, and structural changes induced by high-temperature exposure (sometimes catalyzed by high actinide concentra- tions) do not occur at the conditions of these ex- periments. Dramatic verification (see Fig. 8.2) of this is evidenced by the fact that our uranium coefficients, obtained at concentrations danger- ously near the anticipated C value, show ex- cellent agreement?? with uranium coefficients *° obtained at practically zero actinide concentra- tions (carrier-free 2°2U). In opposition to thin-layer results, preliminary constant-potential data suggest comparatively high coefficients and low activation energies. How- ever, these results may agree with the high- concentration results?? for HTM-PyC. Con- siderations of combined structural and concentration effects might enable reconciliation of the anomalous diffusion behavior as observed at high and low concentration levels. It is clear, however, that thin-layer data are not indicative of fuel migration in carbide fuel-particle coatings since the actinide concentrations carbide interfaces are quite high. at coating~ Self-Ditfusion K. B. Evans III Several dramatic changes occur in pyrolytic carbons when subjected to temperatures above 22 . . .. Similar comparisons indicate a less-than=excellent agreement of thorium coefficients. ‘The magnitudes of ceefficients reported by Wolle: ef al., are greater than ours by a factor of 5; however, .D”/D ratios and AE vajues are comparable, L those at which they are deposited. At high tem- peratures the structures ‘‘improve’ in that they tend to assume a graphitic structure; values of the thermal conductivity and impurity-atom dif- fusivity decrease. The tate and mode of this an- nealing process are diffusion controlled and are best interpreted through self-diffusion measure- ments. In systems other than graphite, self-diffusion measurements may be circumvented, since the self-diffusion coefficient has been correlated with other properties more amenable to experimenta- tion. Analogous correlations for graphite are undetermined; the necessity of conducting the formidable self-diffusion experiments still re- mains, if for no other reason than to obtain cor- relations of the type mentioned. We have initiated a systematic study of the role of self-diffusion in the annealing phenomena associated with graphites and carbons. From this information we hope to estimate diffusion 82 parameters for perfect graphite crystals via extrapolation of data for improved structures. A series of special thorium-pyrocarbon diffusion experiments has been completed pursuant to the selection of materials and annealing conditions for use in the present investigation. Observa- tions of actinide diffusion patterns are quite valuable in surveys of this sort, since actinides migrate preferentially along defective paths in carbons. As a result of these experiments, sev- eral possible carbon candidates were chosen for our initial self-diffusion studies. Equipment and techniques used to study dif- fusion of actinides and fission product species in graphites and carbons are being modified as re- quired in order to pemmit study of '*C diffusion. The development of methods for accurate place- ment of tracer on surfaces and the demonstration of certain means of assay for the diffused isotope pose the most immediate problems. 9. Reactions of Reactor Components with Oxidizing Gases I.. G. Overholser REACTIVITY OF ATJ GRAPHITE WITH LOW CONCENTRATIONS OF OXIDIZING AND REDUCING GASES J. P. Blakely Rates of reaction of a 1-in.-diam sphere of AT] graphite with low concentrations of water vapor and carbon dioxide in flowing helium (1 atm), ob- tained from continuously recorded weight changes and compositions of the effluent gases as estab- lished by a sensitive gas chromatograph, have been reported.’’? These studies have been ex- tended {o gas mixtures containing both water vapor Addition of catbon dioxide to the water vapor increased the reaction rates above those observed for water vapor alone, but the rates found for the mixed oxidants were less than the sum of the rates for the two oxidants. Addition of hydrogen inhibited the reactions of both the oxidants with graphite. In view of the retarding effect of hydrogen on the water vapor—graphite reaction noted earlier,’'? similar studies were performed? to establish what . . . - 3 and carbon dioxide in helium, 'L, G. Overholser and J. P. Blakely, GCR Program Semiann. Progr, Rept. Sept. 30, 1964, ORNI.-3731, pp. 161-67. 2J. P. Blakely and L. G. Overholser, “Oxidation of AT] Graphite by Low Concentrations of Water Vapor and Carbon Diexide in Helium,’' Carbon (in press). *L.. G. Overholser and J. P. Blakely, GCR Program Semiann. Progr, Rept, Mar, 31, 1963, ORNL.-3807, pp. 150-35, . .. G. Overholser and J. P. Blakely, GCR Program Semiann., Progr, Rept, Sept, 30, 1965, ORNL-3885 (in press ). 83 effect, if any, the addition of methane might have on this reaction. Data obtained from these studies are given in Table 9.1. Carbon removal calcu- lated from CO, + CO found in the effluent gases shows that the water vapor—graphite reaction was retarded by the addition of methane, although to a lesser extent than by an equal concentration of hydrogen.? If only weight changes are detennined, one might conclude that the oxidation of graphite by water vapor was completely suppressed. Data obtained from the effluent gases, however, show that both oxidation and deposition occurred. The calculated and observed weight changes agreed satisfactorly in those cases where weight losses or small weight gains were found. The calculated weight changes were significantly larger than the measured weight changes in those instances where large weight gains prevailed. 'This behavior could be due to spalling of carbon from the graphite sur- face. Both the oxidation and deposition rates de- creased with decreasing temperature and were not measurable at temperatures much below 700°C. Carbon deposition rates from methane-helium mix- tures were determined in the absence of water vapor to establish what effect water vapor might have had on the cracking of methane. A compari- son of the calculated deposition rates listed in Table 9.1 with those given in Fig. 9.1 shows that water vapor has very little, if any, effect on the cracking of methane. The overall rates are dif- ferent, however, in the two cases due te loss of carbon by oxidation in the presence of water vapor. The slopes of the plots given in Fig. 9.1 cor- respond to activation energies of ™~ 50 kcal/mole for all concentrations of methane examined. The Table 9.1. Carbon Deposition from Methane —Water Vapor—He!lium Mixtures HZO Concentration of 110 ppm CH4 Concentration Loss and Gain of Carbon {mg em” % hr™h Observed Temperature Flow Influent Effluent Effluent Concentration (ppm) Loss Gain Net Weight o lem? (STP)/min] Gases Gases CO, CO TotalH, Corrected H, Calculated from Calculated from Gain or Change {ppm) {ppm) C02 + CO Correcied H2 Loss (mg cm 2 hrkl) % 1073 % 107 x 1073 x 1077 800 280 18 2 37 8.1 8.1 --8.0 BOO 275 142 132 15 <1 46 16 6.0 3.2 2.8 -3.9 800 300 270 235 i4 <1 46 18 6.1 3.0 -2.2 -1.8 800 275 335 320 i4 1 63 34 6.9 6.7 +3.7 +0.8 800 300 560 520 11 <1 62 490 4.8 8.6 +3.8 +2.0 830 300 635 630 9 <1 63 45 4.0 9.7 +5.7 +2.8 800 275 680 650 10 <1 78 58 4.0 11.1 +7.1 +4.7 750 285 290 280 © <1 20 S 2.5 1.7 —0.8 0.6 750 285 560 550 6 <1 29 17 2.5 3.5 +1.0 1.2 700 360 260 260 2 <1 13 6 0.9 1.3 +0.4 2.7 rg ORNL-OWG 65-11488 TEMPERATURE (°C) _1850 800 750 700 | - . O T T T T }i%‘ """""""""""""""""""" I ] o | T T T T ............. b Lo . f O 560 ppm ZH ) .\ ,,,,,,,,,,,,,, e _ ppmCHy L I \\ [ ® 300 ppm CHy T i A 120 ppm CHy, \‘.f\ 1O~2 Ad.‘:.’xi, ”””””””””” ki T . : o™ 5 E tud *.. = . 2 5 i W g 103 L & 5 l ol S R R } | . | 89 9.4 Q3 95 9.7 9.9 104 t0.3 10,000/, e Fig. 9.1. Effect of Temperature on Carbon Deposition Rates at Yarious Methane Concentrations. apparent order of the deposition reaction with respect to methane falls between 0.5 and 0.8 in the temperature range of 750 to 850°C, Studies® also were made of the disproportiona- tion of carbon monoxide on AT] graphite, using 1501500 ppm of carbon monoxide in helium at tem- peratures of 550 to 850°C. The complete lack of agreement between deposition rates calculated from the carbon dioxide found in'the effluent gases and those obtained from measured weight changes is not understood. Using values for the deposition rates calculated from the carbon dioxide formed, it was found that deposition rates increased with increasing carbon monoxide concentrations. The deposition rates at constant carbon monoxide con- centration increased upon raising the temperature from 550 to 750°C and then remained constant or decreased slightly when the temperature was raised to 850°C. The low rates found suggest that 85 AT] graphite is not an efficient medium for pro- moting the disproportionation of carbon monoxide under the experimental conditions examined. COMPATIBILITY OF PYROLYTIC-CARBON. COATED FUEL PARTICLES WITH WATER YAPOR C. M. Blood Current developments indicate that fuel particles coated with pyrolytic carbon will be present in all- ceramic cores of high-temperature gas-cooled re- actors. Inleakage of water vapor during operation of such a reactor poses hazards because of the high temperatures of the core components, Oxida- tion of the fuel particle coatings could cause coat- ing failures and a resultant release of volatile fis- sion products into the helium coolant. The compatibility of wvarious batches of fuel particles with water vapor has been examined® in an attempt to evaluate the hazards. In these studies, fuel particles were exposed at 1000°C to flowing helium--water vapor mixtures having partial pressures of water vapor of 4.5, 45, and 570 torrs. The fuel particles had various types of pyrolytic carbon coatings and cores of UC, or (U, TH)C,. The reactivity of the fuel coatings was determined from weight changes and effluent gas composi- tfions, The incidence of failure of the coatings was obtained from the quantities of wranium and thorium removed by acid leach following exposure fo water vapor. Microscopic and metallographic examinations were made following exposure to Surface area measurements were made on both oxidized water vapor and also after acid leach. and unoxidized fuel particles. Reaction rates for different batches of fuel par- ticles obtained at 1000°C using various partial pressures of water vapor are given in Fig. 9.2, These data show that the apparent order of the water vapor—coating reaction increased with in- creasing partial pressure of water vapor in virtu- ally all cases. This could be due, in part, to dif- ferences in burnoff at the various partial pressures. C. M. Blood and L. G. Overholser, GCK Program Semiann., Frogr. Rept, Mar. 31, 1965, ORNL.3807, pp. 140—42; GCR Program Semiann, Progr., Rept, Sepi. 30, 1965, ORNL-3885 (in press). QRNL-DWG 55-11362 300 e ——— s o 200 - ' = 160 - 50 _ 20 = > 10 o E 1 5 z - G =z o 2 - : E LAMINAR 1 1 o LAMINAR I e 1 m CRANULAR IV -~ 00 . 7 ISOTROPIC V as - e ISOTROPIC VI T 4 [SOTROPIC VII . : o NCC-217 : 01 ! [ . cea . oo - - . - - 1 2 5 10 20 50 100 200 500 1000 H,0 PARTIAL PREZSSURE (torr) Fig. 9.2. Effect of Partial Pressure of Water Vapor on Reactivity of Fuel Particles at 1000°C. The incidence of failure of the coatings also may be responsible for this behavior. a few particles failed at a partial pressure of 4.5 torrs, whereas at 570 torrs appreciahle fractions failed in all cases. If the core residues catalyzed the reaction, this could account for the change in apparent order observed. This behavior makes an extrapolation to other partial pressures hazardous. Surface area data obtained for oxidized particles from the various bhatches of fuel particles were In general, only extremely variable. In general, no correlation of Also, no consistent relationship of incidence of failure reactivity with surface area was possible. of coatings to reaction rate, surface area, or burn- off Photomicrographs of oxidized particles from various batches showed that the mode of attack varied from batch to batch; this might be anticipated from the variable reaction rate and surface area data. Studies of the effect of support media for the fuel particles on the reaction rate showed that rates obtained with alumina or platinum were essentially the same. Use of a graphite sleeve to support the fuel particles reduced the attack of water vapor on the coatings; the degree of protection increased with increasing sleeve length, This suggests that a was evident. 86 graphite body can afford protection for the particle coatings but that the configuration of the fuel as- sembly will determine, to a large degree, the extent of protection obtained. The studies are continuing, with emphasis being placed on the effect of higher temperature (1200 to 1400°C) and of prior irradiation on the reactivity with water vapor. COMPATIBILITY OF METALS WITH LOW CONCENTRATIONS OF CARBON MONOXIDE J. E. Baker In the BeO-graphite compatibility tests,® metals are used as heat shields in an inert-gas atmosphere containing ~ 200 ppm of carbon monoxide at tem- peratures ranging from 400 to 800°C. Possible deposition of carbon on the metal surfaces and the resulting decrease in reflectance prompted a lab- oratory study of the metals in an atmosphere ap- proximating that prevailing in the engineering tests. Specimens of molybdenuni, gold-plated stainless steel, and mild steel were exposed to flowing helium (1 atm) containing 250 to 300 ppm of carbon monoxide at temperature intervals of ~100°C in the temperature range of ~450 to 850°C. Expo- sure times averaged ™30 hr at each temperature. Weight changes were continuously recorded by a sensitive analytical balance, and the influent and effluent gases were analyzed by a sensitive gas chromatograph. X-ray studies were made of the exposed specimens in an attempt to identify any deposits formed. The molybdenum specimen gained ~10 pg/cm? during an exposure time of ~150 hr. The weight changes were so small that the effect of tempera- ture could not be measured. The exposed speci- men had a very thin dark-gray film and a few spots suggesting localized attack. The film is probably an oxide, which could have been produced by very low concentrations of oxygen and/or water vapor present as contaminants in the helium stream. There was no evidence of carbon deposition. The gold-plated stainless steel specimen showed an overall weight gain of ~ 100 ug/cm? for an ex- posure time of ~150 hr. The rate of weight gain ®C. A. Brandon and J. A. Conlin, GCR Program Semiann. Progr. Rept, Sept. 30, 1964, ORNI.-3731, pp. 202 --6. varied by about a factor of 2 over the temperature range examined, with no consistent effect of tem- perature evident. The exposed specimen had a dull green color, suggesting failure of the gold plate. X-ray analysis showed the film to be Cr,0, with no carbon present. A bright specimen of mild steel lost ~ 170 pug/cm? during ~ 150 hr. The rate of weight loss increased with temperature. The surface was bright at the end of the test and showed no evidence of carbon deposition. A rusty specimen of mild steel lost ~ 1500 pg/cm? during ~150 hr, with no consistent effect of temperature on the rate of weight loss 87 evident. The surface brightened during the ex- posure, and x-ray analysis found only FeO present in the surface film. Since the original specimen had Fe O, and Fe O -H,O present, carbon monox- ide reduction of the oxides seems responsible for the large weight losses. The results indicate that carbon is not deposited on these metal surfaces in this temperature range from low concentrations of carbon monoxide in helium containing essentially no hydrogen or water vapor. The situation could be different in the presence of the latter. 10. Irradiation Behavior of High-Temperature Fuel Material QOscar Sisman FISSION.GAS RELEASE FROM PYROLYTIC- CARBON-COATED FUEL PARTICLES P. E. Reagan J. G. Morgan J. W. Gooch T. W. Fulton C. D. Baumann 2 we showed that ra- In previous experiments' diation damage to the pyrolytic carbon coating was due primarily to fission fragments originating at the core-coating interface. This damage can be alleviated by providing a gap at the core-~ coating interface, by a porous carbon layer, or by a sacrificial pyrolytic carbon layer. Particles of dense UO2 have been coated (in the ORNL Metals Division) by techniques which provide one of these protective measures. We have measured fission-gas release rates from these three kinds of coated particles during irradiations in the A9, B9, and Cl facilities in the ORR.? The irradiation temperature, burnup, and fission-gas release data are given in Table 10.1. Particles in batch OR-298 had heen heat treated to introduce a gap between the uranium oxide core and the inner layer of pyrolytic carbon. The coat- ing on these particles consisted of an anisotropic and Ceramics Iy, E. Reagan, F., L. Carlsen, and R. M. Carroll, “Fission~-Gas Release from Pyrolytic Carbon Coated Fuel Particles During Irradiation,’”’ Nucl. Sci. Eng. 18(3), 30118 (1964). ’p. E. Reagan, J. G. Morgan, and O. Sisman, “Fis- sion-Gas Release from Pyrolytic Carbon Cgated Fuel Particles During Irradiation at 2000 to 2500°F,* Nucl. Sci. Eng. 23(2), 210—23 (1965). 3p. E. Reagan ef al., **Fission Gas Release from Coated Particles,’”® GCR Program Semiann. Progr. Rept. Sept. 30, 1965 (in press). 88 J. G. Morgan inner layer and a granular outer layer. After a little more than 26% bumup at 1400°C, a few bursts of fission gas and the fractional release (see Table 10.1) increased by a factor of about 7. Three experiments were conducted using par- ticles with a porous carbon buffer layer between the uranium oxide core and the pyrolytic carbon coating. Particles from batch OR-354 had a relatively thick (82 u) isotropic coating over the The fractional fission-gas release, as shown in Table 10.1, remained very nearly con- stant throughout the test. Particles from batch OR-348 were similar to those of OR-354 (porous, isotropic on an oxide core), but the isotropic coating was more dense. Initial values for frac-~ release from the OR-348 particles were near 10~ %; the values increased to near 107 by the end of the test. No bursts of fission gas were observed. The third batch of particles (batch OR-HB23) with a porous inner layer were made with sol-gel uo, cores;* these were coated with an isotropic layer next to the porous casbon buffer layer and a granular outer layer. These particles operated for 25 at. % heavy metal burnup at 1600°C. The fission-gas telease rates were quite low (see Table 10.1); they increased by less than a factor of 2 during the entire test. Particles from batch OR-343 were made with an inner coating which would shrink away from the uranium oxide core. These particles had a low- density (~1.5) pyrolytic carbon coating applied directly to the core, followed by a higher-density were released, porous buffer, tional 4J. P. McBride, *‘Preparation of UO Microspheres by a Sol-Gel Technique,’’ ORNL~3874 (in"press). Table 10.1. Fission-Gas Release from Pyrolytic-CarbonsCoated Uronium Oxide Fuel Particles During Irradiation Average Fractional Fission+Gas Release, R/B® Capsale Batch Coating Burnup Temperature 5 ; _ )} (ate % U) {°C) MK B8 87y 13ixe Piixe AY-2 OR-298 Gap + 27.9 1400 9.3% 107" 7ox1077" 57x10777 11,7 x1077F 6sx 10778 anisotropic + 8.5%X107°%° 49x107% 3.4x107% d o granular BY-26 OR~354 Porous carbon + 12.1 1350 4.7 % 1079 4,2 % 1078 3.9 x107° 2.7x107% 17 x 10”8 isctropic ‘BO=27 OR=348 Porous. cathon + 5.4 1580 16X 1670 1.3x 107" 1.2 x 1078 1.0 x 10~° 0.6 x 10~ F dense isotropic Clel6 OR-343 Isotropic I + 11.9 1400 t1x1077 3.8x1077 06x 1077 0.9x1077 0.4 x 1077 isotropic It 2.8 1500 1.9 %1077 1.4 x 1077 1.2% 1077 1,7 x 1077 g7 x1077 13,7 1400° 1.1 x 1977 0.8 % 1077 9.7 x 1077 d d C1-17 OR-HB23 Porous carbon + 25.2 1600 6.2x107% a3x107% 3.0x1078 7.6 x107% s.4x1078 isotropic + granular fRelease rate /birth rate; average for last week of test unless otherwise noted. PBefore bursts of fission gas, CAfter bursts of fission gas. dData not available. €After 2.8% buenup at 1500°C. 6% CRNL-DWG 65-118435 ST OIS GA-3G of ¢ T Qo > T aggee - CUTTTTT B9-26 OR-354 g 11 LT 43500¢ ""1“‘7%“ P L . \/« ,,%,77‘,7?,,‘ . L B00°C e 2 L . e e 1078 U e - 1o’ 2 5 1978 2 5 107> CORTAMINATION, BASZC OM ALPAA COUNT (grams URANIJM] Fig. 10.1, Relationship Between 98Kr Relcase and Cooting Contamination. (~1.8) pyrolytic carbon outer layer. As shown in Table 10.1, the fractional fission-gas release nearly doubled when the temperature was increased from 1400 to 1500°C, but it returned to about the original value when the temperature wasg again decreased to 1400°C. We observed that a correlation could be made between the degree of contamination in the coat- ing and the fission-gas release from particles with unbroken coatings (see Fig. 10.1), This leads to the conclusion that (below 1400°C) the fission-gas release from unbroken coatings is predominantly from contamination in the coating rather than from the fuel core.? We have found that coated oxide particles have less contami- nation in the coating than coated carbide par- ticles. POSTIRRADIATION EXAMINATION OF COATED FUEL PARTICLES P. E. Reagan E. L. Long, Jr.5 The irradiated uranium oxide particles described in the previous section were examined with a 5l‘.h:,-tal's and Ceramics Division. 90 microscope (at 30x), and about 100 particles from each experiment were selected for metallographic examination.? The particles made with an in- tentional gap at the core-coating interface (batch OR-298) revealed no failed coatings, although their behavior during irradiation suggested that a few particles did fail. Metallographic exami- nation (see Fig. 10.2) showed that the inner pyrolytic carbon coating had undergone severe delamination but that the outer coating had re- mained intact. No reaction product was obsetrved at the core-coating interface. cores showed small metallic inclusions in the grain boundaries and a collection of fission-gas bubbles, Postimadiation examination of the uranium oxide patticles from batch OR-354 revealed that only minor microstructural changes had occuired as a result of the irradiation. No failures or evidence of potential failures was noted. The only changes apparent densification of the porous carbon inner coating and the presence of fission-gas bubbles and small metallic inclusions No coating failures and only minor microstructural changes were noted when irradiated particles from batch OR-348 were examined. The only change in the coating was a continuous gap that formed at the core-coating interface. The third batch of particles with a porous inner layer (OR-HB23) had been irradiated to 25 at. % uranium burnup at 1600°C. An apparent densification of the porous catbon layer, as shown in Fig. 10.3, was the only The urasium oxide observed were in the grain boundaries of the uranium oxide. microstructural change noted in these coatings after irradiation. The particles of batch OR-343 were coated in ~an experiment to determine whether a low-density coating applied directly to the UO, core would shrink during high-temperature irradiation and provide the gap necessary to relieve the stress on the coating. This did indeed occur, as is shown in Fig. 10.4. Although about 20% of the coatings showed short fractures spiraling into the inner coating, no failed coatings were observed. The advantages of the UO2 cores over UC cores became more obvious when it was observed that the U0, wonld not convert to UC, as long as the coating was intact. From our experience with a variety of coated UO, particles, we conclude that they are superior to the UC, particles for the following reasons: (1) the oxide does not flow at high burnup and expand into voids or 91 PHGTO 282346 E Y.58967 L ot a et e X e S i ""&és’w:' T "%?';'-&, "'5.9 o 1 - T IS T i L ! Fig. 10.2. Pyrolytic-Carbon-Coated Uranium Oxide Particles from Batch OR-298. Magnification 200x: (a) unir- cadiated; (b} irradiated to 28% burnup at 1400°C in capsule A9-2. 92 PHOTO 82347 Y-62437 15 T Fig. 10.3. Pyrolytic-Carbon-Coated Uranium Oxide Particles from Batch OR-HMB23. Magnification 200x: (a) unirradiated; (b)irradiated to 25% burnup at 1600°C in capsule C1-17. PHOTO 82348 93 e 2 ...fiu’tufl”uu f”%r ! S e e S wm unw I ey /.Nmm.hv % ication 200x: {a) unir- nif Mag icles from Batch OR-343. ium Oxide Part Coated Uran burnup at 1400 Carbon- Pyrolytic- 4. ) . 10 Fig le C1-16. in capsu C C and 3% at 15007 iated to 12% adiated; {&) irrad - 1 cracks as the carbide does, (2) uranium from the oxide will not diffuse into the pyrolytic carbon coatings even at high temperatures, (3) uranium contamination in the coating can be kept to a much lower level during fabrication with oxide cores, and (4) the oxide is not reactive and is much easier to handle during fabrication of the coated particle. POSTIRRADIATION TESTING OF COATED FUEL PARTICLES M. T. Morgan R. L. Towns C. D. Baumann Annealing studies of irradiated coated fuel par- ticles were continued to determine the stability of the coatings at temperatures higher than that 94 of the irradiation and to measure their ability to contain fission products. The particles tested had fuel cores of UC,, (Th,U)C , or UO, with duplex or triplex pyrolytic carbon coatings. They were heated to temperatures up to 2000°C in a flowing helium atmosphere; the fuinace com- ponents were periodically removed for analysis to measure fission product release as a function of time and temperature. ® The release rates obtained in five annealing experiments at 1370, 1700, and 2000°C are com- pated in Table 10.2. The values shown were averaged over 6 hr of anneal following an initial 1-hr heating period at each temperature. A graph of the cumulative releases vs time for barium and strontium from three batches of particles is shown in Fig. 10.5. 5M. T. Morgan, Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, pp. 212~13. Toble 10.2, Fissjon Product Release Rates fream Pyrolytic-Carbon-Coated Fusl Particles During Postirradiation Annealing Annealing T Sample® - Fission Product Release Rates (%/hr) emperature / No. (OC) 140Ba 898r 144Ce 13'.7CS 91Y ______ ; < < - B9-20 1370 0.1 0.3 =0,003 =0.03 B9-21 0.1 0.2 So0.01 0.01 C1-15 0.1 0.04 0.04 0.02 < < B9-20 1700 0.8 0.6 =0,002 =01 Bo-21 0.6 2. 0.04 0.02 Cl-15 0.5 0.4 0.3 0.08 0.3 Cl-l6 6. 2. 0.0006 0.3 0.02 BO-20 2000 2. 2. 0.03 §0.2 BOo-21 6. 7. 5 0.3 Cl-15 3. 5 3. 1 3 C1l-16 10. 7 0.005 7 0.01 ®B9-20 and C1-16 samples are duplex-coated UO2 with uranium burnups of 4.6 and 14.7% respectively. Cl-16 is the sample with the loweredensity outer coating. B9-21 and Cl-15 are from the same batch of triplexscoated (Th,U)C2 particles, but with burnups of 0.29 and 8.9% respectively. ORNL—DWG 65 ~123486 CAPSULE B92-2G, BATCH OR - 201 U0, CORE ~Ba Sr CAPSULE B9-21, BATCH GA-314,(Th,U) G, CORE - rE;!cj ’ 3r CAPSULE C{~15, BATCH GA-314,(Th U] C, CORE--Ba ANNEALING TIME (nr) P OHE O8 Sr FISSION - PROCUCT RELEASE () 0 S0 20 30 :-10 50 SQUARE ROOT OF ANMEALING TIME (min2) ' Fig. 10.5. During Postirradiation Anneal of Coated Fuel Particles at 2000°C, Accumulated Fission Product Release The experiments so far have been performed with fuel materials of current interest. Though no systematic and basic study in fission product release has been made, some trends are evident. The behavier of triplex pyrolytic-carbon-coated (Th,U)Cz particles shows some change with burnup. Particles irradiated to 8.9% burnup suffered 2% coating failures during the initial hour at 1700°C or at 2000°C, while par~ ticles irradiated to 0.29% uranium burnup had no failures during 19 hr at 2000°C. The lower re- lease of barium and strontium at the higher burnup remains without explanation. Particles of UQ_ with an outer coating density of 1.8 g/cm? had higher release rates for barium, strontium, and cesium, but lower release rates for cerium and yttrium than did {Th,U)C2 particles with outer coating densities of about 2 g/cm?. The higher release rates of barium, strontium, and cerium may be due to the lower-density coating, while the lower release rates of cerium and ytirium in the UQO_ particles may indicate better retention by the fuel core itself. uranium POSTIRRADIATION EXAMINATION OF FUELED GRAPHITE SPHERES D. R. Cuneo H. E. Robertson J. G. Morgan C. D. Baumann ' E. L. Long, Jr. We have completed the evaluation of five irra~ diation experiments that contained fueled graphite elements in the form of spheres. These elements were fueled with pyrolytic-carbon-coated particles 400 p in diameter dispersed in a graphite matrix. A summary of the irradiations is shown in Table 10.3. All of the spheres reported in the table were 6 cm in diameter except those in experiment 8B-5, which were 1% in. in diameter. The spheres of interest to the German AVR Pebble Bed Re- actor Program were fabricated with unfueled outer shells of machined graphite or shells molded around the core matrix. Two of the spheres were completely fueled (no unfueled shell). In addition, we examined two spheres of interest to the TARGET program. These were spheres of graphite that contained several drilled holes into which were poured loose coated particles. Ex- periment 8B-5 contained eight 17 -in.-diam spheres which underwent the highest burnups of any spheres tested (™25 at. % of the heavy metal). Metallographic examinations were carried out for four spheres in this experiment; the remaining four spheres were duplications of those examined and were used for compression testing only. We observed a high percentage of failure of the laminar-coated particles in two spheres of this experiment. Laminar-coated particles fabricated by the same manufacturer but irradiated to about 1,45 the burnup were undamaged. Laminar-coated (U,T‘ry)if,2 particles made with normal uranium were found te have ceating fractures at the fuel- coating interface and evidence of loss of crystal- line detail in the fuel core. This is the only case of damage we have observed in particles containing normal uranium. Figure 10.6 is a photomicrograph of a mnormal and an enriched (U,Th)Cz particle from this experiment. We found no broken coatings in the spheres that con- tained either duplex- or triplex-coated fuel par- ticles. Production-run Carbon Products Division spheres, experiment O1A-8, operated successiully. No damaged particles were found, and the non- volatile fission products found in the graphite Fig. 10.6. 96 CHES g N A ~ Pyrolytic-Carbon-Coated (Th,U)C2 Particles in Sphere Mo. 1 from Experiment 8B-5. The enriched particles were from batch 3M-120; the normal particles were from batch 3M-119, The enriched particle (right) reached a burnup of 25.6% heavy metal, but the normal particle (left) reached a burnup of approximately 0.2% heavy metal. powder packed around the spheres showed a relatively small amount of migration through the unfueled shells of the spheres, Relationships compression strength and shrinkage as related to types of sphere manu- facture are given in Table 10.4 for spheres itra- diated to high burnups in experiment 8B-5. All spheres had molded fuel cores. The two spheres that were completely fueled (no unfueled shell) underwent large shrinkages in diameter and an apparent gain in compression strength. The two spheres with molded unfueled shells were found to exhibit the same degree of shrink- age and, for one sphere tested, a decrease in for the The machined unfueled had the least outside diameter However, from the large losses in compression strength (irradiated vs unirradiated spheres), we conclude that the molded cores underwent considerable shrinkage, so that the compression tests reflect only the strength of the unfueled shell. Such loss of contact between the core and shell adversely affects the heat transfer from the core as well as the mechanical strength of the sphere. The same trends can be shown for 6-cm-diam spheres, although not to this extent because of their low burnup. compression strength. shell spheres shrinkage. 97 Table 10.3, Fueled Graphite Sphere ltradiations ORR Fuel Particle Type of Sphere Estimated a Burnup Percent Fxperiment Composition Coating Shell Temperature (% heavy 'metal) Failed b No. (°C) Particles AVR-type spheres 8B-5 (U,Th)Cz, Laminar 1 molded, 700-~1000 25.6-27 67100 2 spheres M) 1 machined 8B-5 (U,Th}Cz, Triplex Machined 800 26.6 D 1 sphere (GA) 8B-5 (U, Th)C , Duplex No shell 800 27. 0 1 sphere (CPD) O1A-8° uc,, Duplex Machined 7501200 9.1-10.8 0 3 spheres (CPD) O1-8 ucC 2 Duplex Machined 700 2.1 Q 1 sphere (ORNL) Q8-8 ucC 2 Puplex Machined 750--1200 3.5-4 0 2 spheres {ORNL) 08-8 uc o Duplex Molded 900 4.2 0 1 sphere (3M) 05-8 (U,Th)CZ, " Laminar Molded 900--1200 7.3-8.6 0 2 spheres (3M) ' TARGE T-type spheres O1-8 (U,Th}CZ in Duplex Solid sphere 7751200 2.1 0 18 holes, with 19 holes UO2 + ThO2 in one hole; 1 sphere o1-8 uc 2 Duplex Solid sphere 850 2.2 0 1 sphere (CPD) with 18 holes AWhere two temperatures are given, the second is central temperature; only one with central thermocouple. bBy metallographic examination. CAVR productionerun spheres. sphere per experiment is equipped 98 Toble 10.4. Shrinkages and Compression Strengths of Graphite Spheres lrradiated to Burnups of 25 to 27 at. % in Experiment 8B-5 Unfueled Shells of Spheres Average Diameter Thickness Shrinkagea Type (in.) ' (%) Machined 0.2 0.85 Machined 0.2 0.77 Machined Q.2 0.61 Machined 0.2 0.44 No shell 0 1.49 No shell 0 2.32 Molded 0.2 1.54 Molded 0.2 1.38 Unirradiated Compression Loading Equivalent Spheres, to Failure (1b) Compression loading toe Failure (1h) 1725 (2)° 900 1500 (2) 225 1725 (2) 275 1500 (2) 2650 1740 (4) 3050 1740 (4) 1250 2200 (3) 2200 (3) FAverage of readings taken at pole, equator, and temperate regions of spheres, Number in parentheses indicates number of unirradiated spheres tested. Value given for lecading to failure is an average. POSTIRRADIATION EXAMINATION OF EGCR FUEL ELEMENT PROTOTYPE CAPSULES M. F. Osborne E. L. Long, ]r.5 H. E. Robertson J. G. Morgan Except for one EGCR prototype capsule which is still being irradiated, all the elements in this series have been examined. Of those elements irradiated in the ETR,’ six were examined vis- ually and three were evaluated in detail, The capsules coantained UQ, pellets fabricated at ORNL; these pellets were either solid, hollow, or hollow with a BeQ bushing. The design power rating during irradiation was 35,000 Btu hr™! ft—!, and the stainless steel cladding tempera- ture varied from 1250 to 1550°F. The fuel re- ceived burnups of from 4500 to 14,300 Mwd per metric ton of uo,. Only one element experienced cladding failure. The other elements had only minor cladding deformation such as circumferential ridges at pellet inteifaces and a collapse of the cladding against the fuel. These effects were observed in previous tests.?® The element with cladding failure contained solid UO, pellets. A longitudinal tear in the cladding was apparently caused by overpowering of the element early in the irradiation. Grain growth in the cladding at the site of failure indi- cated the presence of a hot spot. Unlike earlier instanceg of cladding failure in this series, there was no evidence of nitride formation in the clad- ding or columnar grain growth in the fuel. We have completed the evaluation of an EGCR piototype capsule which had been irradiated in the ORR. This element was of special interest as it contained actual EGCR production-run UO, pellets and 304H stainless steel cladding. The irradiation was carried out at a heat rating of 31,000 Btu hr™! ft~! and a cladding temperature of 1300°C. The dished-end hollow fuel pellets achieved a burnup of 10,000 Mwd per metric ton of UOz. Under these conditions, the cladding had collapsed around the fuel but no circumferen- tial ridges were formed. Bowing of the element was nominal, and there was no shifting of the fuel. The element had not failed, and its per- formance was satisfactory under EGCR design conditions. M. F. Osborne et al., Reactor Chem. Div, Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, pp. 218--19, 8c. D. Baumann, Irradiation Effects in the EGCR Fuel, ORNL~3504 (June 1965). 1. Fission-Gas Release R. M. Carroll Oscar Sisman G. M. EXPERIMENTAL From the resuits of in-pile experiments on single- crystal UO, specimens, we have concluded that fission gas is not released by classical diffusion We conclude that the release is con- 2,3 These observa- PIOCESSEes. trolled by a trapping process, tions are consistent with those by other experi- menters?'® and have led us to a defect-trap theory of fission-gas release.® This theory postulates that a defect in the UO, crystal structure will trap migrating fission gas; the rate of fission-gas es- cape, accordingly, is controlled by the number of traps present. Gas-trapping defects consist of in- herent flaws, such as grain boundaries and internal pores, in the UO, as well as point defects and clusters of point defects created as a consequence of the fission process. Steady-state experiments with single crystals’ and with polycrystalline UO, (ref. 8) have supported this defect-trap theory. l¢onsultant from the University of Florida. ’R. M. Carroll and Oscar Sisman, Nucl, Sci. Eng. 21, 147--58 (1965). 5R. M. Carroll and Oscar Sisman, Fuels and Materials Development Program Quarterly Progress Report, Septem- ber 1965 (in press), TR, M. Carroll, ‘““The Behavior of Fission-Gas. in Fuels,”” AIME: Conference on Radiation Effects, Septem- ber 1965 (1o be published by AIME). ' SR. M. Carroll, Nucl. Safety 7(1) (in press). ‘R, M. Carroll, R. B. Perez, and Oscar Sismun, J. Am. Ceram. Soc. 48(2), 5559 (1965). "R, M. Carroll and P. E. Reagan, Nucl, Sci. Eng, 21, 14146 (1965). 3R. M. Carroll and Oscar Sisman, “‘In-Pile Fission- Gas Release from Fine-Grain UO,,"" J. Nucl. Mater. (in press). - During Fissioning of UO, T. W. Fulton R. B. Perez! Watson 99 We have modified the experimental assembly to permit a slow, controlled oscillation of the speci- men through the flux of neutrons.”? '? These modi- fications permit a controlled sinusoidal variation of neutron flux (fission rate) or specimen tempera- ture. The fission gas released during the oscilla- tions is monitored continuously by a gamma-ray spectrometer. Time dependence of the temperature (or the fission rate) and of the fission-gas release rate is measured and recorded by punch-tape read- out.'1-12 A computer program has been developed to analyze the fission-gas release waves in terms of their Fourier components; this analysis yields amplitude and phase-shift information as a function of frequency of the oscillation. The effect of temperature oscillations was ob- tained by comparing the steady-state release (at zero frequency) with that at different frequencies.!? The gas release was found to increase as the fre- quency of oscillations increased; at very slow oscillations the release rate approached steady- state levels. This result is predicted by the defect- trap model; '? that is, as the oscillation frequency IR, M. Carroll and Oscar Sisman, Trans. 4Am. Nucl. Soc. 8(1), 22 (June 1965), accepted for publication in Nuclear Applications. ]'OR. M. Carroll and Oscar Sisman, Fuels and Materials Development Program Quarterly Progress Report, Septem- ber 1965 (in press). "R, M. Carroll et al., GCR Program Semiann. Progr. Rept, Mar, 31, 1965, pp. 8087, ORNI.-3805. IR, M. Carroll et al., GCR Program Semiann. Progr. Rept, September 1965 (in press). 3R, B. Perez, Trans. Am. Nucl. Soc. 8(1), 22—23 (1965); to be published in Nuclear Applications. EMISSION {atoms fsec) - o 88y, ORNL-DWG 85- 4305 3 4 (x10'3) NEUTRON FLUX (neutrons fem? - sec) Fig. 11.1. Knockout Release During Fission Rate Oscillations. increases, the frequency-dependent factors over- come the trapping effect; the release rate in- creases and approaches that predicted by models hased upon classical diffusion theory. Fission-gas release at temperatures below 600°C occurs primarily by a recoil process wherein fis- sion fragments passing through the surface of the UO, specimen eject or ‘‘knock out’” UO, molecules along with any fission-gas atoms in the knockout zone. Knockout release during fission-rate oscil- lations at a constant temperature of 575°C is shown in Fig. 11.1. The data points of Fig. 11.1 form a sort of hysteresis curve; arrows by the data points indicate whether the fission rate was increasing or decreasing when the data were ob- tained, The hysteresis curve shows the gas re- lease rate to be higher during fission-rate oscilla- tions than when the fission rate was constant. We suggest that the data of Fig. 11.1 reflect the rhythmically changing concentration of fission-gas traps created by the fission process. destruction and creation of traps allow the fission gas to reach the specimen surface in surges. These surges account for both the hysteresis and the in- The cyclic creased amount of gas at the specimen surface available for knockout. MATHEMATICAL MODEL Material balance equations for fission-gas re- lease {parent and daughter nuclides) from a thin slab of fissionable material are based on produc- tion, loss, and diffusion-leakage terms. In general: Rate of change of concentration — diffusion-leakage contribution -- loss terms + production terms . For the diffusion-trapping model the loss terms in- volve radioactive decay and trapping by intrinsic flaws and point defects. The production terms in- volve the fission rate, fission yield, and, for the daughter, radioactive decay of the parent. The concentration of point defects is a balance be- tween their rate of formation as a consequence of fission and their destruction by annealing at high temperature. 101 Three material balance equations (for which the symbols are defined in Table 11.1) result; they are coupled and nonlinear. oM 3?2 Parent: e D, 57T (A + {20) — BN, (B | M(z,0) + B, F(D) . oH a2 Daughter: =5 D, Pl Ay + 8,) — hiV () | H{2,1) + B, F(O + A, M(z 1) . o, Point defects: m?zf—i =aF() —v.N, . ; , At steady state (OM/dt = OH/dt = ON /dt = 0) the equations become linear and can be solved by conventional methods. The steady-state solutions for the release rates yield expressions which are similar in appearance to the expressions obtained from the old diffusion model,’? except that the decay constant, A, is replaced by an effective decay constant, A’, which includes the effect of trapping. An increase in temperature will increase the rate of trap annealing and will decrease the effective decay constant, A", tron flux will increase the value of the A”, will tend to neutralize the increase in production, and will make the release rate fairly insensitive to flux An increase in neu- Table 11.1. Definition of Symbols Symbol Definition M, H Atomic concentration of parent and daughter nuclides, respectively, atoms,/cm3 DW’ 1 Diffusion coefficients, cmz/sec, = DG[EXP (~AE/RT)) Fission vields, fission fragments /fission 24 Rate constant for trap annealing = L/O[exp (-“AE“/R’I')] a Traps formed per fission gfi Trapping probability, due to permanent de- fects, s.ec'"1 er Concentration of points defects, defects/cm3 R Second-order rate conslant, (cm3)/(dei‘ect) (sec) changes. This behavior agrees qualitatively, at least, with the observed phenomena at steady state. An approximate solution has been obtained for the case where the flux is varied sinusoidally while the temperature remains constant. To com- pate the diffusion-trapping model with the simple diffusion model, the analytical expressions for the total release rate transfer functions for each model were coded for numerical computations. The ampli- tude of the release rate transfer function vs fre- quency is shown in Fig. 11.2a. When the diffusion model is used this amplitude decreases with the square root of the frequency; when the diffusion- trapping model is used this amplitude exhibits an extended plateau in the low frequency range. As the frequency increases, the frequency-dependent factors overcome the trapping effects and both ORNL-DWSG 85~1165 -3 b /T,-"“‘a/ e L - 1 7 o / i P o } ! A7 = 30 L b . Al [ | * o -, L e DIFFUSION-TRAP THEDRY ~ & e DIFFUSION THEORY /{/’i’ E)J ~2Q b b R /./ . - < 2 & & ; A | O =00 S | | | |2 T=100 °C ‘ { ! ; j b= w2 5 1073 2 5 107E o [radians x sec™) Fig. 11.2. Release Rate Transfer Function vs Fre- quency. models tend to coincide. For lower frequencies the trapping effect predominates; hence the fre- quency-dependent terms do not significantly affect the magnitude of the transfer function. Equivalent comparisons were made with the phase shift (Fig. 11.2b) of the transfer function. In the diffusion model, the phase shift quickly reaches an asymptotic value of —45° and is independent of 102 temperature. The phase shift for the diffusion- trapping model tends slowly toward the --45° asymptote, and is a func- tion of temperature. Data have been obtained by oscillating the tem- is always smaller, more perature at constant flux; for this case, the equa- tions do not linearize readily and the solution has not been completed. 12. EQUILIBRIUM STUDIES IN THE SYSTEM ThO,-U0,-UO, L. O. Gilpatrick C. H. Secoy Equilibrium studies of the system ThO,-U0,-UO, were continued at partial pressures of O, equal to that found in the atmosphere and at temperatures ranging from 1200 to 1550°C. Previous work has established the general features of the phase dia- gram for the system. '—* A study was made to determine if the scatter in analytical compositions could be due to experi- mental techniques. A temperature of 1550°C and a composition of 90 mole % urania and 10 mole % thoria were chosen for this work. in grinding, storage conditions, and lapsed time had little or no effect on the observed equilibrium composition. Initial oxidation state and degree of initial solid solution formation had a very small effect on the equilibrium composition (less than 2.5%). When gquench conditions were varied by substituting liquid N, submersion in place of rapid air cooling, a reduction in UQO, content from 29 to about 26 mole % was obtained. Com- positions the unit cell size* Variations calculated from derived from x-ray powder patterns also confirmed that the fcc phase had a UO, composition near 25 mole %. These findings would indicate some surface oxidation during the quench period by the older technique, 13, A. Friedman and R. E. Thoma, Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1963, ORNL-3417, pp. 13034, 21.. O. Gilpatrick, H. H. Stone, and C. H. Secoy, Reac- tor Chem. Diyv. Ann. Progr. Rept. Jan. 31, 1963, ORNL- 3417, pp. 13430, 3. o. Gilpatrick, H. H. Stone, and C, H. Secoy, Reac- tor Chem. Div. Ann. Progr., Rept. jan. 31, 1944, ORNL~ 3581, pp. 16064 4. o Gilpatrick and C. H. Secoy, Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNIL.-3789, pp. 239-43, 103 Miscellaneous Studies for Solid-Fueled Reactors Unit cell dimensions were computed from x-ray powder patterns using a least-squares fitting pro- gram written by Williams® in place of the more tedious graphical procedure. Test cases showed no bias due to this change. Comparisons were also made between the Debye-Scherrer camera and the bench diffractometer methods of measur- ing lattice parameters by the powder technique. Good agreement was found in this case also. An effort was made to better define the transition temperature at which orthorhombic U30 4 converts to the face-centered cubic UO,, phase in air. The transition is sluggish, but the first sign of the fcc phase appeared at 1528°C and progressed fairly rapidly (50% in 2 hr) at 1540°C. No fcc phase was cbserved after 2 hr at 1514°C. Tem- perature calibration was felt to be reliable to +5°C as measwred by two independent methods. the year, Cohen and Berman of the Westinghouse Bettis Laboratory reported on this system® and found a lower urania composition (50 mole %) for the two-phase boundary at 1200°C than had been found in this study (~63 mole %).? In view of this difference, a remeasurement at 1200°C was made using more recent technigues. No effect could be ascribed to the use of pellets at 60 mole % urania. This composition showed some orthorhombic phase (V2 to 5%). Composi- tions of 56, 53, and 45% urania displayed no detectable orthothombic phase. This indicates that the phase boundary is between 56 and 60 mole % urania. Composition of 90% urania showed little change from the older work. Apparatus and equipment have been designed and built to extend this study to lower partial pres- sures of oxygen in the range of 107% atm, where During D, R. Williams, LRC~2: A Forlran Tattfice Constant Refinement Program, 1S-1052 (November 1964)., %1, Cohen and R. M. Berman, Am. Ceram. Soc. Bull, 44(4), 391 (1965). shifts in the oxygen-metal ratio should be large enough to be easily measured. starting compositions have also been prepared by more extreme procedures ous material in the solid oxides have been reduced in H, at temperatures in excess of 1650°C, and reduced mixtures have been fused in helium at temperatures above 2200°C. A new series of to assure a homogene- golution form. Mixed BEHAVIOR OF REFRACTORY-METAL CARBIDES UNDER IRRADIATION G. W, Keilholtz R. E. Moore M. F. Osborne Refractory-metal carbides of groups IV to VI have potential applications in high-performance nuclear power plants and in reactors for special applications requiring extremely high power den- sities. A series of experiments in progress is aimed at determining the changes in physical of monocarbides of and mechanical properties Ii, Zr, Nb, Ta, and W during fast-neutron irradi- 104 the form of 1/2-in. X l,é-in. cylinders over the temperature interval 100 to 1400°C, the neutron flux range 0.5 to 3.0 x 10!* neutrons cm™? sec™! (>1 Mey), and the neutron dose range 0.5 to 5 x 102! peutrons/cm? (>1 Mev). Included in the experimental series are specimens of each of the monocarbides made by three different methods: (1) hot pressing, (2) slip casting and sintering, and (3) explosion pressing. Three low-temperature (100 to 400°C) uninstrumented assemblies and one instrumented high-temperature (1100°C) assem- bly containing these carbides are undergoing iria- diation in the ETR. Other experiments are planned to achieve temperatures up to 1400°C. Examination is complete for an assembly of one sample of each of the five hot-pressed monocar- bides irradiated (see Table 12.1) in the ORR at about 100°C.7 Most of the specimens sustained very little gross damage. Metallographic examina- tions revealed no evidence of grain-boundary sepa- ration in any of the samples. The volume increase of the carbide specimens calculated from dimen- sional measurements and the volume increase calcu- lated from the lattice parameter expansions are 7G. W. Keilholtz, R. E Moore, and M. F. Osborne, ation. The specimens are being irradiated in ORNL-TM-1350 (in preparation) (classified). Table 12.1. Gross Volume Expansion and Lattice Parameter Expansion of Refractory-Meta! Carbides Irradiated at 100°C Volume Increase Volume Increase Crystal Fast-Neutron Thermal-Neutron Aa/aoa /\c/cea from Lattice from Dimensional Material Structure Dose, nvt ]?ose, nvt {7) (%) Parameters®? Measurements (>1 Mev) (>1 Mev) (%) ) x 1021 x 10°1 WwWC Hexagonal 1.7 3.4 0.17 0.35 0.7 0.3 TiC Cubic 1.7 3.4 0.58 1.7 2.5 NbC Cubic 1.7 3.4 0.16 0.5 (3.6)° TaC Cubic 1.4 2.7 2.7 ZrC Cubic 1.4 2.7 0.33 1.6 2.6 4, L. Yakel, ORNL Metals and Ceramics Division, personal communications. obtained from x-ray diffraction patterns of unirradiated specimens and ¢ in each case were based on values for a o 0 from the same batches. The values for Aa/ao and AC/CO "The volume increases were calculated from the equation AV/V0 = 2Aa/ao + Ac/c0 for hexagonal crystals and from the equation f.\V,/VO = SAa/aO for cubic crystals. ©This value was extrapolated from dimensional measurements of the diameter of the niobium carbide specimen; the length of the specimen could not be measured because one end was found to be broken off after irradiation. given in Table 12.1. The lattice parameters were measured from x-ray patterns obtained from reflec- tions from polished surfaces, The gross volume ex- pansion of the tungsten carbide specimen, the only one of the five carbides which does not have an isotropic crystal structure, was only about 0.3%. The gross volume expansions of the cubic carbides of titanium, tantalum, and zirconium were much greater (™2.6%); the lattice expansion accounts for about 60 to 75% of the gross ex- pansion of the titanium and zirconium carbides. An x-ray pattern could not be obtained on tan- talum carbide. The niobium carbide sample ap- peared o expand more than the others, but its lattice parameter expansion was very small, These preliminary results indicate that refrac- tory-metal carbides are =ufficiently resistant to fast neutrons to merit consideration for nuclear applications involving high neutron doses. Tung- sten carbide, in particular, is attractive because of its small lattice parameter expansion and small gross volume expansion. The magnitudes of these expansions are generally reliable indi- cations of the degree of neutron damage. EFFECTS OF FAST-NEUTRON IRRADIATION ON OXIDES G. W, Keilholtz R. E. Moore M. ¥, Osbome A systematic investigation of the effect of fast- neutron irradiation on sintered compacts of MgO and Al O, has been completed, and the results have been compared with those previously ob- tained® 16 for BeO. Several hundred cylindrical specimens of each oxide with various grain sizes and densities were irradiated in the Engineering Test Reactor over the temperature range 100 to 1100°C and the fast-neutron flux range 0.5 fo 812, P. Shields, J. E. Lee, Jr., and W. E. Browning, Jr. Effects of Fast Neutron Irradiation and High Tempers- ture on Beryllium Oxide, ORNL-3164 (March 16, 1962). R, P. Shields, J. E. Lee, Jr., and W. E. Browning, Jr., Trans. Am. Nucl. Soc. 4(2), 338 (1961), 1064 Ww. Keilholtz et al., Radiation Damage Solids, Proc. Symp., Venice, 1962, vol. 11 (Vienna, IAEA, 1962). e, w. Keilholtz, J. E. Lee, Jr., and R. E. Moore, The Effect of Fast-Neutron Irradiation on Beryllium Oxide Compacts at High Temperatures, ORNL-TM-741 (Dec. 11, 1963). 3.0 « 10%* neutrens cm™? sec™! (>1 Mev). The results are reported elsewhere, 717720 Among the three oxides, the greatest difference in behavior was between MgQ, which has a cubic structure, and BeQO, with a hexagonal crystal structure. Previous resulis have shown that the primary mode of damage to BeO is grain- boundary separation caused by anisctropic crystal The degree of damage, that is, frac- turing or powdering, iucreases with increasing dose and decreases with increasing temperature. The cubic structure of MgQO precludes anisctropic expansion, and the lattice parameter expansion is much smaller than that of BeO. Accordingly, irradiated MgO exhibits virtually no grain-bound- ary separation, no powdering is observed, and the pross volume expansion is small relative to that of Bel, Transgranular fracture is severe in MzO. For example, about 40% of specimens irradiated at 150°C over the dose range 0.2 to 1.4 x 107! neutrons/cm? were fractured. Unlike BeO, dam- age to MgQ specimens was not related to the neutron dose. Approximately 40% of the specimens irradiated at 800°C and about 0% of the speci- mens irradiated at 1100°C over the dose range 0.5 to 5.1 = 104! neutrons/cm?® (>1 Mev) were frac- The randomness of fracture can expansion. tured randomly. 12(}. W. Keilholte ef al., Behavior of Be( Under Neu- tron Irradiation, ORNL-TM-742 (Dec. 11, 1963). 13G. W. Keilholtz, J. E. Lee, Jr., and R. E. Moore, J- Nucl. Mater. 11(3), 25364 (1964). 14, W, Keilholtz ot al., J. Nucl, Mater. 14, 87—95 (1964). 15(}. W. Keilholtz, J. E. Lee, Jr., and R. E. Moore, Ylrradiation Damage to Sintered Beryllium Oxide as a Function of the Fast-Neutron Dose and Flux at 110, 650, and 11007¢C,*" submitted to Nuclear Science and Fngi- neering for publication. 16()1. W. Keilholtz, J. E. Lee, Jr., and R. E. Mocore, Reactor Chem. Div, Ann. Progr. Rept. Jan. 31, 1965, ORNL-378Y9, pp. 23338, 17G. W. Keilholtz, J. E. Lee, Jr., and R. E. Moore, ORNL~TM-1050 (Mar. 26, 1965) (classified). 13G. W. Keilholtz, J. E. Lee, Jr., and R. E. Moore, ORNL-TM-1140 (July 9, 1965) («olassified). 196, W. Keilholtz, J. E. Les, Jr., and R. E. Moore, ORNL-~TM-1300 {in press) (classified), 204, W. Keilholtz, J. E. Lee, Jr., and R. E. Moore, “Properties of Mapnesium, Aluminum, and Beryllium Oxide Compacts, . Irradiated to Fast-Neutron Doses Greater than 107 Neutrons cm ° at 150, 800, and 11007 (2,** accepted for pablication in Proceedings of Joint Division Meeting of the Materials Science and Technology Division of the American Nuclear Society and the Refractories Division of the American Ceramic Society, May 8—11, 1966, Washington, D.C. 106 Table 12.2. Lottice Parameter Expansion of MgQ, CL-A1203, and B0 lrradicted at Low Temperatures to Comparable Fasi-Meutron Doses Crystal Irradiation Fast-Neutron Material . . , , ‘1, @ Structure Temperature Dose, > 1 Mev /,,\a/ao Ac = AvV/v ) (neutrons/cmz) % 1021 Mg Cubic 150 1.1 0.0005 0.0015 CL-A1203 Hexagonal 150 1.0 0.0023 0.0024 0.0070 BeD Hexagonal 110 1.0 0.0013 0.0312 0.0338 9The fractional volume incrcase, AV/VO, for BeO and Al,O0; was calculated from the equation .«f\V/VO = Q(Aa/ao) + (,t'\c/co). The equation AV/VO o S(g:\a/ao} was used for the case of cubic MgQ. be explained by postulating that a minimal neu- tron dose can weaken the crystals encugh to per- mit the propagation of cracks from randomly dis- tributed sites within MgQO compacts. The minimal dose is probablv near the lower limit of our ex- periments (~2 x 102° neutrons/cm?, >1 Mev), since increases in strength have been reported for 21,22 lower doses by other experimenters. Direct evidence is lacking that higher doses produce weakening of MgO crystals, but electron micro- graphs of MgQO irradiated to doses greater than 102! npeutrons/cm? (Figs. 12.1 and 12.2) show a severe deterioration which, it seems likely, would be accompanied by a loss in strength. In-pile ancealing at high temperatures reduces the volume expansion and lattice parameter ex- pansion of MgO. Thermal stresses within neutron- damaged compacts probably account for the mark- edly greater gross damage observed in irradiations at 1100°C. The crystal stiucture of Al O, is anisotropic, but the behavior of A1203 on exposure to fast neutrons resembles that of MgO rather than BeO. 2lp A, J]. Sambell and R. Bradley, Phil. Mag. 9, 161 (1964). 22_]. Elston, ‘*Behavior of Neutron-Irradiated Beryllium Oxide,’ Saclay Center of Nuclear Studies Report DM-1182 {(1662). A comparison of the lattice parameter expansions of the three oxides irradiated at low temperatures to a dose of ~102! neutrons/cm? is shown in Table 12.2. The a and ¢ parameters of Al,O, expanded by about the same amount under these conditions; this is in sharp contrast to the behav- ior of Be(), in which nearly all the expansion is in the ¢ parameter. Recent irradiations of trans- lucent Al O, specimens (lucalox) to somewhat higher doses (~1.4 x 102! neutrons/cm?) pro- duced a greater expansion of ¢ parameter, how- ever, and resulted in an anisotropic expansion ratio (Ac/co)/(Aa/ao) of about 3.8. 'This ratio, however, is small compared with that for BeO. Therefore, the mechanisms through which Al O, is fractured during fast-neutron irradiation appear to be the same as in MgO. Neutron damage to all three oxides, as judged from lattice parameter expansion and gross vol- ume expansion, decreases with increasing irradi- Because of in-pile thermal gross fracturing of Mg0O and ation temperature. stress, Al,0, increased markedly at high temperatures (1100°C). Therefore, to minimize damage to these oxides in nuclear reactor application, the temper- ature should be maintained as high as practicable, and the system should be designed so as to mini- mize thermal stress. In the case of BeO, thermal cycling, which tends to promote grain-boundary separation, should also be avoided. however, 107 F PHOTO 81629 S Fig. 12.1. Electron Micrograph of Unirradiated MgQ of Density 3.4 g/cm3 ond Grain Size 10 ¢t at 32,000x Magni- fication. Reduced 11%. 108 *: - A, Fig. 12.2. Elzctron Micrograph of Mg0 of Density 3.4 g/cm3 and Grain Size 10 ;¢ hrradioted to a Fast-Neutron Dose of 1.2 x 102! Neutrons/cm? (>1 Mev) at 150°C at 32,000x Magnification Showing Genera! Deterioration and Small Cracks. Reduced 12.5%. ANNEALING OF IRRADIATION-INDUCED THERMAL CONDUCTIVITY CHANGES OF CERAMICS C. D. Bopp The annealing of the neutron-induced thermal conductivity change has been measured in several ceramic materials. The annealing temperatures are listed in Table 12.3. A fluxmeter apparatus?3 was used with the materials of high conductivity, and a mercury-contact apparatus?? was used with the materials of lower conductivity. Previously,?® the mercury-contact apparatus was used for pre- and postirradiation measurements of these same materials, but annealing was not studied. Com- parison with the present results shows good agreement except in the instance of thin speci- mens of beryllia, for which it appears that the mercury-contact apparatus is unsuitable because of the extremely high conductivity of this mate- rial. The annealing temperatures for beryllia shown in Table 12.3 are in good agreement with those found by other workers?® for material given nearly the same irradiation dosage. 23¢. D. Bopp and O. Sisman, Reactor Chem. Div. Ann. Progr. Rept, Jan. 31, 1965, ORNL-3789, p. 232. 24 b, Bopp, J. Am. Ceram. Soc. 47, 154 (1964). 250. Sisman, C. D. Bopp, and R. L. Towns, Solid State Div. Ann. Progr. Rept. Aug. 31, 1957, ORNL-2413, p. 80, 26y Elston and C. Labbe, J. Nucl. Mater. 4, 143 (1961). 109 Taoble 12.3. Annealing of Irradiation-lnduced Therma! Resistance Temperatureb at Which the Irradiation-Induced Irradiation Thermal Resistance Material Dosage” Is Decreased by the Indicated Percentage 20% 40% 60% 80% Sintered beryllia 2x 10'% 600 650 750 900 Sintered alumina 2% 101° 600 1050 c Sapphire 4-8x 107 600 850 1000 1250 Spinel 2x16Y 556 900 Porcelain 2x10'® 5350 950 ¢ Forsterite 2 x 101° 750 1250 c Zircon 2x10'% 1000 1050 1100 1250 Cordierite 2x10'% 755 1000 1100 1250 Steatite 2x10*% 550 800 ¢ “The dosage unit is fast neutrons /em? (>1 Mev) ex- cept in the_instance of =zircon, for which the unit is fissions/cm” (since the presence of a trace of uranium impurity dominated the radiation effect in zircon; see C. D. Bopp et al., Reactor Chem. Div. Aan. Progr. Rept. Jan. 31, 1965, ORNL-3789, p. 231). bAnnealing was conducted isochronally; the sample was heated for periods of 1 hr at successively higher temperatures using 50°C increments, except that the pericd of heating at 1250°C, the highest temperature used, was 4 hr. ®This amount of annealing was not attained after the 1250°C heating. Part IV Other ORNL Programs 13. Chemical Support for Saline Water Program THERMODYNAMICS OF GYPSUM IN AQUEOUS SODIUM CHLORIDE SCLUTIONS W. L. Marshall Ruth Slusher In an extensive investigation of the equilibrium phase relationships of gypsum (CaSO4-2[-120) in NaCl-H_ O solutions continuing from the previous work,! solubilities were determined at 12 dif- ferent temperatures from 0 to 110°C and in solu- tions of sodium chloride from very dilute to those saturated with sodium chloride or a new solid phase. All results and available literature data were evaluated by an equation, log K, = log K2 +8S\I/(1 + AVD) sp —~ Bl - CI* -2 log a, , (1) where 85 is the Debye-Hiickel limiting slope for a 2-2 electrolyte corrected for molal units, a_ is the activity of water, I is the ionic strpngth, A, B, and C are adjustable parameters, K is the sol- ubility product constant (at I = 0), and K __ is the o product [(Ca?™) (80,77 DetermmedpvaTues of K”p are shown in Fi 1g 13.1, and the parameters A, B, and C are included in Table 13.1. With the obtained parameters, the differences between most calculated and experimental values of K op Were between 0.5 and 3% over the entire ranges of con- ceatration and temperature, Standard Thermodynamic Yalues for Gypsum Assumptions were made that the standard heat of solution, AHS, and the standard change in heat lw, L. Masrshall, Ruth Slusher, J. Chem. Eng. Data 9, 187 {1964). and E, V. Jones, 113 ORML--DWG A5-- 11415 7{*C) Lo &0 80 40 20 0 BT T { ‘ """" T ,,,,, . ] I Lo i . L L_v /..—P‘ i a4 R ot . '-\.\\i' Jl A ™ oo N // : - ° % 4'5 e —— L - b _— —_ X / g /s 1 /P’ T Tty T T T T 4-6 S / ........ / X */“ - 1% OF ABSOLUTE VALLE o | —— a7 lf —— — 26 28 o) 32 24 15 Y X107 Fig. 13.1. Logarithm of Kfp vs 1/T(°K) for Ca$0,s 2H,0 (Gypsum), 0~110°C. capacity at constant pressute, ACZJ could be ex- pressed by AH® = E + {ACS dT ) and AC; =F + GT, 3) where &, F, and G are parameters, These ex- pressions were substituted into the van’t Hoff equation, d1n Kgp/d(l/Tj =—-AHYT , {4 which was then integrated over all values of Kgp and T(°K) to obtain a four-parameter equation. With the sepatately determined values of K: (unsmoothed but obtained using smoothed values 114 Table 13.1. Standard Thermodynamic Quantities and Parameteis 4, B, and C Used in Eq. (1) for the Equilibrium CaSO ;2H,0(s) e=2Ca?(aq) + $0,2(aq) + 2H,0 T (°C) AF® Ar® As® A B c (kcal/mole) (kcal/mole) (cal mole ! °C~ 1) 0 5.58 +2.50 -11.3 1.450 —0.0680 0.0264 10 5.71 +1.55 —14.7 1.468 —-0.0274 0.0212 20 5.87 +0.68 -17.7 1.490 —~0.0072 0.0178 25 5.96 +0.27 —-190,1 1.500 1+ 0.0006 0.0164 30 6.06 —0.12 —20.4 1.510 +0.0076 0.0160 40 6.28 -0.85 ~22.8 1.530 +0.0178 0.0150 50 6.52 —1.50 —24.8 1.544 +-0.0200 0.0138 60 6.77 —2.07 ~-26.5 1.558 +0.0200 0.01256 70 7.05 -2.57 —-28.0 1.570 +0.0200 0.0112 80 7.33 -3.00 ~20.2 1.580 +0.0200 0.0094 90 7.63 —-3.35 —30.2 1.588 + 0.0200 0.0072 100 7.94 —~3.62 --31.0 1.594 +0.0200 0.0050 110 8,25 —3.82 —31.5 1.595 +0.0200 0.0038 of Asp) from Fig. 13.1, the four parameters were evaluated by the method of least squares to obtain the equation log Kgp = 390.9619 — 152.6246 log T — 12545.62/7 + 0.081849287T . (5) The average deviation from this equation of the experimentally determined values of Kg shown in Fig. 13.1 was 10.6%. Values of AHI® at each temperature were obtained by differentiating Eq. (5) with respect to 1/T(°K) and substituting the the result into the van’t Hoff equation, (4), while those values of AC® were obtained by differen- tiating with respect to T the resulting expression for AH® While separate values of AC® might be expected to be somewhat inaccurate, the average value of AC® from 0 to 110°C of —57 cal mole™! °C~! is believed to be significant. Values of AS® and AF® were obtained from the standard thermodynamic equation, AF®= AH® — T AS°. (6) Representative calculated thermodynamic values obtained by these procedures are Table 13.1. included in The Additional Sslid Phase At concentrations of NaCl above 3 to 5 m and at temperatures from 70 to 95°C, a saturating solid other than CaSO4-2H20 or NaCl was found. In special experiments at 70°C, this second saturating solid phase was identified by petro~ graphic examination? to be N32804-5CaSO4-3H20, found previously in the system CaClz-NazSO4- H2O.3 By the formation of N82$O4-5Ca804-31{20 from solutions initially only of NaCl and CaSO,, the system becomes a four-component system, tather than three, and must be defined by the components CaSO4, NaCl, CaCIz, and H20. 2Thanks are due G. D. Brunton, Reactor Chemistry Division, for these examinations. 3A. E. Hill and J. H. Wills, J. Am. Chem. Soc. 60, 1647 (1938). THE OSMOTIC BEHAVIOR OF SIMULATED SEA-SALT SOLUTIONS AT 123°C P. B. Bien B. A, Soldano The isopiestic technique previously developed at ORNL* has been used to test the behavior of thtee simulated sea-salt solutions at 123°C. The tests were made to assess the applicability of calculation methods proposed for determining thermodynamic properties, including the wvapor pressures, of solutions of interest to the vacuum- distillation process for water.” desalination of sea- The three sea-salt solutions were prepared to simulate closely the ‘‘standard seawater’” defined by K. S. Spiegler.® In order to avoid the evolution of gases and possible corrosion in the vapor chamber, the troublesome univalent anions HCOs‘ and Br™ were replaced by equivalent quantities of chloride ion. Solution A was further modified by replacing the calcium with magnesium, thereby increasing the magnesium concentration to 0.06451 m; the sulfate concentration in solution A was kept at 0.02856 m. Solution B was modified further by completely removing the calcium to- gether with an equivalent amount of sulfate; the sulfate concentration in solution B was thus re- duced to a concentration of 0.01822 m. Saline C was not modified except by the replacement of chloride for the bicarbonate and bromide. Thus, the compositions of the three solutions were: A (m) B (m) C {m) cat? 0,01084 Mgt 0.06451 0.05417 0.05417 gt 0.01007 0.01607 0.01007 Na 0.47564 0.47564 0.47564 50,7 0.02856 0.01822 0.02856 1™ 0.55761 7.55761 0.55761 I 0.7078 0.6664 0.7078 Four dishes of each solution, four dishes of NaCl standard solution, and four dishes containing standard weights for internal calibration of the balance were loaded into the chamber together with a larger dish containing NaCl solution which 8. A. Soldano and G. S. Patterson, J. Chemn. Soc., 1962, 937, “R. W. Stoughton and M. H. Lietzke, Data 10, 254 (1965). 6K. S. Spiegler, Sea Water Purification, Wiley, New York, 1962, J. Chem. Eng. 115 gerved as a buffer reservoir, After the air was evacuated from the sealed chamber, the apparatus was brought to and held at 123°C. Each day, the weights of every dish were recorded as readings on an electrically operated magnetic balance, after which a small amount of steam was vented from the apparatus, thus increasing the concen- trations of all the test, standard, and buffer solu- tions. The changes in the concentrations of the test solutions were inferred from the changes in the weights of the dishes. The jonic strengths of the solutions were then calculated from the formula 1 2 I :;21111.2‘. , with the concentrations m, deduced from the changes in the solution weights. Table 13.2 presents average values of I for the NaCl standard Toble 13.2. Experimenta! Yalues for lonic Strength, }:mj.ziz/?. Date, 1965 [NaCJ. IA IB IC March 9 0.6030 0.8550 $.7965 0.9130 10 0.6683 0.8228 0.7884 0.8714 11 0.6734 0.8318 0.8019% 0.8580 12 0.6987 0.8662 0.8087 0.5134 15 0.6693 0.8758 0.8358 0.%911 17 0.7744 0.9611 0.9278 1.0011 12 0.8658 1.0229 1.0165 1.1158 23 1.4601 1.6585 1.5788 1.6892 24 1.5750 1.7812 1.7510 1.8034 25 1.6342 1.8484 1.7464 2.0445 26 1.8466 2.1308 2,0421 2.0568 26 1.9674 2.2016 2.1102 2.2416 23 2.0728 2.2008 2.1406 2.3070 31 2.6250 2.6484 2.5905 2.9369 April 1 2.6496 2.7254 2.6530 2.9761 2 2.7894 2.8929 2.8513 2.9808 5 3.1081 3.0117 2.9846 3.3108 5 3.3583 3.3064 3.2420 3.5966 6 32.6166 3.5664 3.4352 3.7656 7 3.6801 3.3466 3.4851 3.8864 5 4.3018 4.0551 4.0087 4.43950 2 4.6016 4,1999 4.0G093 4.4954 9 6.1244 5.5620 5.2566 6.2986 12 5.1064 4.5226 4.4029 3.0060 13 7.9924 6.3590 6.5914 7.6788 14 7.9248 6.5999 6.2487 6.9634 , ISOP{ESTIC RATIOS Yygct Igact A= o w0 0.8 116 ORNL—DWG €6—278 ..—-“""—'—- | /,_,-«-F SALINE A~ SALINE ¢ e e e S 1 2 3 4 5 6 7 8 IONIC STRENGTH QF NaCl Fig. 13.2. lsapiestic Ratios of Salines A, B, and € Compared. solution and the three test solutions at 26 dif=- ferent levels of water activity., Values for the isopiestic ratios R for each of the test solutions were obtained at each point from the ratio INaCl/ Itest soln * The equations fitted to R as a func- ion iven as foll : tion of [Nac1 are giv s follows R, =0.7701 + 0.08797 - 0.00471%; o, =0.0292, A RB = 0.8082 + 0.0889! — 0.0052[%; Op = 0.0270 , B and R_ = 0.7427 + 0.0784/ — 0.00551% o, = 0.0344 C Plots of these equations, as shown in Fig. 13.2, suggested that solution C was clearly different in behavior from solutions A and B. The con-~ siderable overlap between solutions A and B suggested that a more sensitive test could be made of the possible significance of any dif- ference between them. To this end, the hypothesis was made that the solutions did not differ, and a t test was applied to the 26 values of I, — IB obtained from the data in Table 13.2. An average value of the difference found was 0.077 unit of ionic strength (not units of R); this difference was found to be well above the 0.001 level of significance. Marshall al.” have recently shown that calcium sulfate should begin to precipitate from seawater at about 117°C. Therefore, it is likely that all the tests with solution C were performed in the presence of solid CaSO, (anhydrite) and that, consequently, the data for solution C should not be suitable for analysis until and unless accurate values for the solubility of the com- ponents make it possible to infer the true solu- tion composition. It may be noted, however, that the values of R calculated from the true solution composition should be higher than those plotted in Fig. 13.2, probably bringing the data more closely in line with those of solutions A and B. The difference between A and B was observed to be quite consistent over the whole concentra- tion range, and the average value of this dif- terence, 0,077 ionic strength unit, is about twice the value calculated for the solutions as made up at room temperature, 0.0414. The solubilities of et 7W. L. Marshall, Reacfor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, p. 294. the components of solutions A and B have not yet been determined as a function of concentration, and it is still assumed that no precipitation took place in the dishes containing these solutions. With respect to the original objective, that of testing the proposed method of calculating vapor pressures of saline waters, solutions A and B may not be sufficiently different to permit good evaluation of such a method. Further tests should be made with solutions known to be phase stable at elevated temperatures and probably should begin with simpler systems such as those con- taining mixtures of NaCl and Na2SO4 or mixtures of NaCl and MgCl1,. ALUMINUM- AND TITANIUM-ALLOY CORROSION IN SALINE WATERS AT ELEVATED TEMPERATURES E. G. Bohlmann J. C. Griess F. A. Posey® J. F. Winesette The studies of the corrosion of aluminum and titanium alloys in sodium chloride solutions have continued. They have included investipations in the 100-gpm titanium loops,® the small titanium efectrochemical loop,'? and conventional labora~ tory equipment. It was found necessary to modify the electrode assembly in the small titanium loop to reduce troublesome IR drops; this entailed ptoviding coaxial polarizing electrodes and dual reference electrodes as shown in Fig. 13.3. Galvanostatic polarization studies of the cor- rosion of the 5454 (27% Mg, 0.8% Mn, 0.1% Cr, <0.01% Cu) and 6061 (1.0% Mg, 0.6% Si, 0.25% Cu, 0.25% Cr) aluminum alloys in 1 M NaCl at 150°C have been catried out in this equipment.!! The results obtained were generally similar to those obtained by other workers at lower temperatures. Thus, a pronounced minimum exists in the cor- rosion rate of aluminum and its alloys in chloride solutions in the vicinity of neutrality. Changes BORNL Chemistry Division. YSaline Water Conversion Report for 1962, U.S, Dept. Intericr, pp. 12, 16, 10962, g, a. Bohlimann, F. A. Posey, and J. F. Winesette, Reactor Chem. Div. Ann. Progr. Rept. fan. 31, 19635, ORNIL-3789, p. 296. Ny G. Bohlmann and F. A. Posey, *“‘Aluminum and Titanium Cortosion in Saline Waters at Elevated Tem- peratures,’’ Proc. First Intemational Symposium on Water Desalination (in press). 117 with pH in the polarization curves of anodic and cathodic processes occurting at the aluminum- electrolyte interface provide a kinetic basis for understanding this and other aspects of the cor- rosion behavior. At low potentials, the rate of the anodic or corrosion reaction is independent of the electrode potential, but it increases with increasing pH. The rate of the anodic process is controlled by the rate of mass transport of hy- droxide ions to the oxide-solution interface. At higher potentials, in the presence of chloride ions, the anodic-polarization curve exhibits a pitting potential which independent of the anodic current density. The pitting potential does not vary with pH, but decreases with in- creasing chloride concentration. The cathodic reaction in alkaline solution consists of the re- duction of water molecules to {form molecular hydrogen; this process is pH independent. With increasing acidity, reduction of hydrogen ions be- 18 comes increasingly important. The minimum cor-~ rosion rate represents a compromise between the decrease in the rate of the transport-controlled anodic reaction with increasing acidity and the increase in the rate of the cathodic hydrogen- evolution reaction. Oxygen in solution may also increase the corrosion rate by providing an addi- tional cathodic process. Comparison of polarization curves of the 5454 and 6061 alloys shows that the rate of the cathodic hydrogen-evolution reaction of the 6061 alloy is considerably greater than that of the 5454 alloy. The enhanced rate of the cathodic process on the 6061 alloy accounts for its greater corrosion rate at any pH and for its susceptibility to pitting attack. Catalysis of the cathodic process on the 6061 alloy may be attributable to its copper content. It is hypothesized that accumulation of copper at the metal-oxide interface as corrosion progresses results in pit formation. These studies stemmed from results obtained in the 100-gpm titanium loops, which showed gross pitting of 6061 alloy specimens after 248-hr exposure to pH 6.0, 1 ¥ NaCl at 150°C. Under comparable conditions, the 5454 alloy showed uniform attack after 1620-hr exposure; continuing corrosion rates were "~ 6 mils/year at 7 to 25 fps. Similar results are being cbtained at 100°C; 6061 specimens show pitting attack after 500 to 1500 hr, whereas 5454 specimens show uniform attack with continuing corrosion rates of ~1 mil/year atter ~2800-hr exposure, ASBESTOS WICK FROM CALOMEL ELECTRODE GASKET INSUL ATOR ORNL-DWG 65-42264 POLARIZING ELECTRODE CALOMEL REFERENCE ELECTRODE SPECIMEN NQ. 3 — Investigations of the severe corrosion of titanium in saline waters, reported last year, have also Three racks of specimens are being exposed in the water box of the No. 1 effect at the Freeport Desalination Plant. In this location, continued. the specimens are exposed to near-normal con- centration, pH ~7.5 brine (sulfuric acid treated and neutralized for scale prevention) at 129 to 138°C. Racks have been examined at 12-, 32a, and 82-day intervals and have all shown negligible to severe attack in areas of contact with Teflon; no attack was observed in metal-to-metal contact areas., This was consistent with the previously noted promotion of attack by contact with Teflon. It appears likely that this results from the release of small amounts of fluoride ion!? by the Teflon. Anodic polarization studies have shown that as little as 12 ppm fluoride substantially increases the active corrosion rate and the rate of activation of titanium. It is also likely that the absence of more general attack on the rack specimens stemmed from the somewhat low temperature of exposure — the bulk of the exposure was at temperatures less than 138°C. The previously noted dominant importance 12k, G. Bohlmann and J. C. Griess, Reactor Chemn. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3780, p. 297. of temperature was reemphasized by recent ex- perience with 100-gpm titanium-loop piping. For- mation of several large pits was observed during operation at temperature 2150°C. Subsequently, the loop has been operated at 100°C for ~2800 hr with no apparent further attack. A similar pit initiated in another part of the loop penetrated the 0.2-in. pipe wall after approximately 5400 hr at 150°C and after 300 hr at 200°C, 1! A particularly significant point conceming the very sharp temperature dependence has just been discovered. This is that the breakdown or pitting potential for titanium, reported as ~10 v in room- temperature chloride solutions,!® has a very sharp temperature dependence. Studies in the small titanium electrochemical loop showed pitting po- tentials at plus 500 mv (vs S.C.E.) or less for commercially pure titanium at temperatures of 150°C and above. Laboratory and loop electro- chemical studies have shown this sharp tempera- ture dependence over the range 25 to 190°C. The actual potentials measured and temperature de- pendence, however, are greatly influenced by alloy composition and surface condition, so no quantitative data will be presented at this time. 13N, Hackerman and C. D. Hall, Je., J. Electrochem. Soc. 101, 321 (1954). CHEMISTRY OF SCALE CONTROL E. L. Compete J. E. Savolainen The tendency of seawater to deposit scale in distillation limits range and brine~concentration factor of the process and affects the efficiency. The problem economic: scale control costs probably should not exceed a few cents per 1000 gal. This elimi- pnates many technologically feasible processes, unless salable by-products are produced. Raw-materials costs alone appear to be too high for most processes except those using compounds of carbon or sulfur. Sulfuric acid addition is conventionally used to prevent the formation of calcium carbonate or magnesium hydroxide scale and permits operation up to temperatures near 250°F. Calcium sulfate may precipitate above this temperature. The prevention of alkaline scale in seawater heaters by means of carbon dioxide additions has been supgested. Ellis'* observed the effect of carbon dioxide pressure on the solubility of calcite in various sodium chloride solutions at elevated temperatures. Thermodynamic equilibrium constants and mean activity coefficients of cal- cium and bicarbonate ions were reported. We adapted these findings to the composition and alkalinity of standard seawater, assuming that the solutions and activity coefficients are similar at equal ionic strengths. It was thus estimated that the addition of 0.9 1b of CO, per 1000 gal of seawater feed, corresponding to a CO, partial pressure of .09 atm at 25°C, might prevent calciie precipitation up to 125%C (257°F). In a once-through flash-evaporator system, alkaline scale would precipitate in the brine bulk as catbon dioxide flashed in the first stage. Heating surfaces would not be fouled, and C02 might be recycled without compressors. Data were not available permitting consideration of magnesium hydroxide precipitation. The assumption that seawater resembles sodium chloride solutions of equal ifonic strength neg- lected consideration of ion-pair formation. The chemical model of seawater at 25°C of Garrels!® and Thompson showed a significant degree of equipment the temperature is 144, J. Eliis, Am. J. Sci. 261, 239~67 (1963). 15R. M. Garrels and Mary E. Thompson, Am. J. Sci. 260, 57—66 (1962). 119 association of sulfate, bicarbonate, and carbonate jong with magnesium and sodium and also with ions. Our preliminary estimates for 100°C have indicated that magnesium sulfate pair formation will be greater at the increased tempera- ture. Calcium sulfate is similatly affected to a lesser extent. These phenomena may also affect the application to seawater problems of the studies of Marshall, Slusher, and Jones!® on the solu- bility of calcium sulfate in sodium chloride solu- calcium tions. The economic removal of calcium sulfate scaling tendencies may be possible by thermal precipitation techniques in which seawater is heated to a temperature of calcium sulfate super- saturation; precipitation may be facilitated by contacting with calcium sulfate particles. An understanding of ionic and solubility equilibria and precipitation kinetics are needed to provide a rational basis for an economic process using this technique. The rate of precipitation from a supersaturated solution is affected by mass transfer to the crystal surface, reaction with surface sites, and nu- cleation of new particles. Without nucleation, the rate cannot exceed that of mass transfer. Co- efficients for mass transfer to particles were related by Harriott!? to that for a single freely falling sphere in the given liquid. The coefficient decreases as particle size increases up to about 100 ;2 and changes little thereafter. For anhydrite crystallization from standard seawater with about 10°F superheating, we estimated that the mass- transfer limitation would permit growth rates up to several centimeters per hour. Such rates would permit an effective unit for removing calcium sul- fate from seawater to be designed. Neilsen!? has indicated that, as supersaturation is decreased and the surface reaction becomes limiting, crystallization rates decrease more than the concentration by a substantial proportion. On the other hand, if the concentration is in=- creased and high levels of supersaturation are reached, undesirably rapid nucleation is to be anticipated. Thus, the supersaturation must be 16w, L. Marshall, Ruth Slusher, and E. V. Jones, [. Chem. Eng. Data 9, 18791 (1964). 17peter Harriott, A.L.Ch.E. (Am. Inst. Chem. Engrs.) 7. 8(1), 93-102 (1962). 185, E. Neilsen, Kinetics of Precipitation, Mac~ millan, New York, 1264, maintained in the range controlled by mass trans- fer for the most effective operation of “‘contact- stabilization’ equipment. In addition to capital and heat costs of the contact-stabilization process, pumping-energy costs may be appreciable. At higher temperatures, a higher head is required in order to pressurize the 120 feed to prevent boiling. For example, at 330°F the vapor pressure of water is 103 psi, and about 1 kwhr is required to pump 1000 gal of feed against this head. The maximum economic tem- perature is thus limited by the exponential rise in vapor pressure. 14. Effects of Radiation on Organic Materials W. W. Parkinson EFFECT OF RADIATION ON POLYMERS W. W. Parkinson W. K. Kirkland R. M. Keyser Polytetrafluorcethylene (Tetlon) degrades at very moderate radiation doses in air, but it has been found to show much less sensitivity when irradi- ated in an inert atmosphere ot vacuum. 1.2 Because of the usefulness of Teflon in deleterious environ- ments and because the films used in previous work ? would be extremely sensitive to atmosphere, sheet stock of 1[3,) and 1/6 in. thickness has been investigated. Tensile specimens have been ir- radiated at about 25°C in air and vacuum over a range of doses up to 7 x 107 rads. Tensile properties were measured in air at room tempera- ture. The specimens in vacuum were outgassed at 5 > 107° torr and 140°C and sealed in glass capsules. The dose rate was 1.2 x 10° rads /hr. The tfensile properties of the 1/1 ;-in. sheet ir- radiated in air and in vacuum are plotted vs dose in Fig. 14.1. It is significant that in vacuum the tensile strength decreases to 40-45% of its orig- inal value at 2 x 10° tads but remains almost constant from this dose to perhaps 10° rads or more, It is seen that useful mechanical properties are refained in an inert atmosphere to doses in excess of 3 » 107 rads. In air both the "/16- and 1/$,)=-ixl.-t}1i(:1': specimens retain 33% of their original tensile strength at 10° rads, in contrast to the films of the earlier work,? where the tensile el b Bopp and O. Sisman, Nucleonics 13(7), 28 {1955). 20, A. Wall and R. E. Florin, J. Appl. Polymer Sci. 2, 251 (1959). 121 Oscar Sisman strength decreased to very low values in this dose range. The maximum in the elongation at break, ob- served in the 1/32--1'.11. as well as the 1/1 gmin. spec- imens, is interesting in the light of “zero strength time’’ measurements, demonstrating that scission predominates at doses above 10% rads.® Probably, radiation-induced defects in the crystal structure, indicated by a minimum in the density-dose curve at 10? rads,® increase the ductility and account for the maximum in the elongation. RADIATION-INDUCED REACTTIONS OF HYDROCARBONS R. M. Keyser W, W. Parkinson Two processes are under study as possibilities for utilizing the fission energy which appears as kinetic energy of the fission fragments. A *°Co assembly ig used currently as a more convenient source of ionizing radiation than a nuclear reactor. One of the processes is the hydrogenation and alkylation of coal in mixtures with alkanes. The other process is the synthesis of amines in mix- tures of ammonia with alkanes and alkenes. The radiation-induced reactions of naphthalene in hexane are being investigated initially as a model system for the coal process.® A tempera- ture-programmed gas chromatograph is being recon- ditioned, and high-temperature columns suitable *A. Nishioka et al., J. Appl. Polymer Sc¢i. 2, 114 (1959). 4W. W. Parkinson et al.,, Keactor Chem. Div. Ann. Progr, Rept. Jan, 31, 1965, OENL-3789, p. 320. 122 ORNL.-DWG 66-979 7000 -\ | ‘ e . ; | : J | A TENSILE STRENGTH, IRRADIATED IN AIR 6000 - Lo el o | A ELONGATION, IRRADIATED IN AIR O TENSLE STRENGTH,IRRADIATED IN VACUUM ® ELONGATION, IRRADIATED IN VACUUM (Y4—in. SHEET) 5000 | J ~~~~~~ © A | /"—m"h':’\ | | 4000 3000 ENSILE STRENGTH ( psi 2000 1000 P , i L | —NL ; 10° LY RADIATION DCSE ({rads) Fig. 14.1. for analysis of the expected products are being procured, The effect of strong bases on reactions involving ionic intermediates in irradiated hydrocarbons has been demonstrated recently.® These results have led us to search for amines in the radiolysis products from solutions of ammonia in n-hexane and in hexene-1. Solutions for irradiation were prepared by condensing known amounts of an- hydrous ammonia into deaerated samples of n- hexane and hexene-l. most solutions were of the order of 2 to 3 mole %, which is about the limit of ammonia solubility in these hydrocarbons at room temperature and Ammonia concentrations in atmospheric pressure. Irradiations were carried out in a 20,000-curie ®°%°Co source at a dose rate SW. R. Busler, D. H. Martin, and T. F. Williams, Dis- cussions Faraday Soc. 36, 102 (1963). Tensile Properties of liradiated Teflon. of 8.5 x 10'% ev g7 ! min~?!. Total doses were in the range 5.0 x 102% to 8.6 x 102! ev/g. After irradiation the samples were opened and subjected to gas chromatographic analysis using a column with a liquid phase consisting of tetra- hydroxyethylenediamine with amine added as a tail reducer. tetraethylenepent- This column gives good separation with only slight peak tailing of calibration mixtures containing expected products such as the 1-, 2-, and 3-aminchexanes and lower- molecular-weight amines. The results obtained so far have not been prom- ising. No amines have been detected in the ir- radiated solutions, the limit of detection corre- sponding to a G(amine) of 0.2 molecule per 100 ev, Mechanistic considerations indicate that a possible reaction path leading to amine formation in ir- radiated ammonia-hexane solutions may be formu- lated as follows: 123 + —_ CH,,—>CH, S +e, (1 4 + + CH,~-—>CH " +H, 2) + + CH +NH, —>I[CH NH]I, (3) . 34 _ [Cfil:{laNHSj +e” —> CH NH +H-, (4) C H o ellgg * 07 > CGH, g () In view of the telatively low concentrations of ammonia possible at atmospheric pressure, reac- tion (5) may occur before the C6H1_3+ carbonium ion encounters an ammonia molecule as in (3). Experiments with ammonia at pressures above atmospheric are currently in progress in an effort to increase the ammonia concentration in the system. ADDITION REACTIONS OF FURAN DERIVATIVES C. D. Bopp W. D. Burch® W. W. Parkinson Polar organic compounds have been observed to add to olefinic groups in chain reactions initiated by chemically produced free radicals or ionic intermediates. Such chain reactions offer the possibility of the high vyields required for com- utilization of the radiation from radio- isotopes. Materials which may be candidates for upgrading through such reactions are the furan derivatives. The preducts, hicyelic and tricyclic ethers, would be useful for solvents and for further processing into chelating agents, surface-active mercial agents, ete. A survey is being performed of the radiation- induced addition of saturated furan and similar ethers to unsaturated furan denvatives and al- kenes. Compounds which have been investigated are listed in Table 14.1 and grouped as saturated compounds (telogens) and unsaturated compounds. Solutions containing a telogen and an unsaturated compound in concentrations of 10 to 1 by volume irradiated and partially analyzed by gas In many cases, infrared spectra were chromatography. 5Temporary employee from Kansas Stafte University, Manhattan. were recorded of the starting materials and the product mixture. The telogen was present in greater concentration in the mixtures since it was desired to promote the formation of 1:1 telomers in the reactions (1)—(3) below and to minimize the addition of second and third molecules of the unsaturated compound. Hp e, He [P SO L/H ~AM - He Hal X Hz N 0 TCHy o~ “CH (1) HE H2 H ™~ oI L e T 2 ~. 2 0 TcH o E -0 A Lo H Tj» H Ha Ha ~o7 (2) H') HE _ =, —— sl ¢ H:T“ T Hyp ~ L?H 3 Hp ~y Hyp M Mo . H. . H . |__\ j\/\uu e . o T Ha Ha % , 07 CHa H?. ~ F{Z * 0 {3) Solutions containing 2-methyltetrahydrofuran were studied over a range of concentrations and doses since the hydrogen abstraction, step (3), proceeds with greater ease in the case of a tertiary hy- drogen. The chromatographic analyses show that con- siderable guantities of products are formed in the molecular-weight range of dimers and simple telomers. The results for solutions of dihydro- furan in Z-methvltetrahydrofuran indicate that the major product changes trom a dimer or 1:1 adduct to a trimer or 1:2 adduct when the dose exceeds about 8 x 107 rads. Increasing the concentration of dihydrofuran in the solution favors the lower- molecular-weight product at the higher doses. Comparison of solutions of tetrahydrofuran and 124 Table 14.1. Compounds lrradiated in a Survey of Addition Reactions Saturated Compounds (Telogens) Unsaturated Compounds Name Structure HE o H2 Tetrahydrofuran Hy L 1 H2 0 1‘12 s H2 2-Methyltetrahydrofuran L H Hp L 0~ CHj Hp Ha 2,5-Dimethyltetrahydrofuran H H CH3 Q CH3 CHgz CHz Diisopropy! cther H({—‘O"CIZH [ [ CH3 CH3 g Isopropyl alcohol HC —OH I Cl"f3 Structure Name Cyclohexene H H Dihydrofuran \[—-.—.,j/ Ho H, \O H H 2,5-Dimethylfuran CHgz 0 CHz H H Furan 2-methyltetrahydrofuran in both dihydrofuran and cyclohexene suggests that tetrahydrofuran gives higher yields than the 2-methyl compound. Evaporation of the irradiated mixtures indicated that over 20% of the original materials had been converted to nonvolatile residue at doses of about 4 % 10® rads. Initial yields in terms of consump~ tion of dihydrofuran in mixtures with methyltetra- hydrofuran were estimated at G values (molecules per 100 ev) of about 10. Infrared spectra indicated the formation of small quantities of hydroxy and carbonyl compounds. Mixtures of telogens with furan and 2 5-dimethyl- furan gave very little 1:1 or higher telomer, probably because of resonance stabilization of the furan ring. Further work will involve identifying the products indicated by gas chromatography and measurement of yields at elevated temperatures. 15. Fluoride Studies for Other ORNL Programs THE CHEMISTRY OF CHROMIUM IN THE FLUORIDE VOLATILITY PROCESS B. J. Sturm R. E. Thoma In earlier versions of the Fluoride Volatility Process, volatile UF _ was obtained by fluorina- tion of fluoride melts in which solid fue!l elements had been dissolved by treatment with HF. For zirconium-’ aluminum-based® fuel elements the method worked well because the available low- melting salt mixtures permitted the fluorination at temperatures of 600°C or lower, where corrosion was not severe. Fuel elements with matrices and claddings of stainless steel proved less tractable; a process in which UF _ is recovered by volatiliza- tion from a fluidized bed instead of a melt® looks promising for these fuels. Uranium hexafluoride from both processes is purified by adsorption- desorption cycles on beds of NaF, In the fluidized-bed procedure the fuel elements are supported in a fluidized alumina powder, which acts as a heat transfer medium, and are decom- posed by treatment with HF-0, mixtures. Uranium is then recovered as UF 5 by fluorination. When chromium, an important constituent of stainless steel, reacts with the HF-0O, mixture, it forms nonvolatile oxides and fluorides. On fluorination, chromium is oxidized to higher valence, and large quantities of its volatile fluorides and oxyfluorides form. These compounds might volatilize with and contaminate the recovered UF , Little is known of these volatile compounds of chromium, Reactor Chemistry Division researches performed or - (38 lohem. Tech. Div. Ann. Progr. Rept. May 31, 1963, ORNL-3452, p. 26, 2Chem. Tech. Div. Ann. Progr. Rept. May 31, 1564, ORNL-3627, p. 40. SChem. Tech. Div. Ann. Information Meeting Program, Nov. 10--12, 1965, pp. 5~7. 125 in support of the ORNL Fluoride Volatility Process program have, accordingly, been devoted to the chromium compounds which may affect the new method. Synthesis of CrF4 Freesenergy estimates® indicated that the re- action AgF 4+ CiF, 2 CF , + AgF , AF = =29 keal , might be used for synthesizing CeF,. In laboratory tests, we found that dense blue-black CiF, vapors were evolved above 400°C, the boiling point of the compound, when equimolar amounts of the re- actants were present or when there was an excess of AgF . Reaction of Chromium Fluerides with NaF Since UF_ is purified by absorption on NaF beds, the behavior of chromium fluorides with NaF was studied, Two compounds, 3NaF - CrF, and SNaF - 3CiF |, formed during the crystallization of molten NaF-CtF, mixtures containing less than 50 mole % CrF,. The former compound (ap- parently a structural analog of cryolite) exists in two crystal modifications. The high-temperature form melted reproducibly at 1090°C and exhibited a range of solid-solution stoichiometry, The low- temperature form is biaxial, with an average re- fractive index of 1.411. X-ray diffraction pattems from 5NaF~3CrF3 powder suggest that the coms pound, melting incongruently at 800°C, is iso- morphous with chiolite (5NaF=3A1F3). In the 4A‘ Glassner, The Thermochemical Prog’:)erties of the Oxides, Fluorides, and Chlovides at 25007 K, ANL-5750 (1957). composition region 0 to 50 mole % CrF,, the sys- tem NakF-CrF | has two invariant points: a eutectic at 14 mole % CrF, and 873°C and a peritectic near 42 mole % CrF , and 800°C, An NaF-CiF phase was synthesized by expos- ing granulated NaF to CrF, vapor at 400 to 500°C in a controlled-atmosphere glove box. This product had a face-centered cubic structure with a lattice constant of 7.90 A, Since it is isostructural with the high-temperature form of KZCIFG,S pound is believed to be Na,CrF . the coms- Behavior of Cr02F2 Chromium oxyfluoride has been prepared® by reaction of CoF, and CrO, and has been exposed in the vapor state for 0.5 hr at 80°C to several of the materials expected to contact the Fluoride Volatility Process gas streams. Sodium fluoride reacted with the CrO I, to produce an (as yet unidentified) yellow-orange compound. Aluminum fluoride, UF,, and UO, showed no evidence of reaction. Equilibrium data’~? for the reaction 2HF () + Cr0,(s) & CrO,F (&) + H,0(8) suggest that AG for solid CrO,F, at 298°K should be near — 189 kcal/mole. From this value one predicts that CrOF should oxidize CeF, to CrF, uo, to UO,, and (by 2 kcal per gram atom of F) UF, to UF,. No tests of the first two reactions have been made; UF4 was not oxidized to UF5 during the relatively mild exposures above. PREPARATION OF LiF SINGLE CRYSTALS BY THE MODIFIED STOCKBARGER METHOD R. G. Ross R. E. Thoma As part of the AEC Pure Materials Program, sustained efforts have been made within the Re SH. C. Clark and Y. N. Sodana, Can. J. Chem. 42, 50 (1964). 6G. D. Flesch and H. J. Svec, J. Am. Chem. Soc. 80, 3189 (1958). A, Engeibrecht and A, V. Grosse, J. Am. Chem. Sac. 74, 5262 (1952). 89. A, Munter, O. T, Sepli, and R. A. Kossatz, Ind. Eng. Chem. 39, 427 (1947). 9R. A, Oriani and C. P. Smyth, J. Am. Chem. Soc. 70, 125 (1948). 126 actor Chemistry Division to develop techniques and apparatus required for the production of very pure, large (350 g) single crystals of LilF with selected isotopic ratios. In a previous report10 we indicated that by refinement of standardized techniques’! for preparing lithium fluoride single crystals, we could routinely produce such crystals in which metallic impurity concentration does not exceed 30 ppb. Crystals were grown from molten LiF in capsules of grade ‘A’ nickel, The single contaminant detected in the LiF product by the best analytical methods available is manganese, its source is the nickel capsule wall, from which Mn® and/or Mn2* diffuse into the melt. The principal application of these LiF crystals has been; in testing theoretical models telating the variation of the thermal conductivity of the crystals with their ®Li-71.i ratio. For this purpose, workers at the Cornell Materials Science Center!? have found that crystals must be sufficiently pure that the impurity does not mask the isotopic effect. While adequate for most studies which require very pure crystals of LiF, material which contains as much as 1 ppm Mn?" is insufficiently pure for testing theoretical models under consideration. In attempts to improve the purity of LiF crystals further, we have modified the Stockbarger ap- paratus to grow LiF crystals in a capsule lined with laboratory-grade platinum, has proved to be very effective in improving the chemical purity of crystalline LiF. A 270-g crys- tal of LiF, designated ORNI.-11, and containing 99,99 at. % ‘Li, was grown in the platinum-lined capsule. As in previous ORNIL preparations, the crystal was vacuum annealed and handled sub- sequently only in vacuum or inert atmosphere, This innovation The crystal was found to contain a lower concen- tration of heavy-metal contaminants than is cure rently detectable by activation analysis. 'That is the most likely contaminant, is <1,86 parts per billion, and the crystal is purer than any of our previous efforts. to say that the concentration of manganese, 10¢, F. Weaver et al., The Production of LiF Single Crystals with Selected Isotopic Ratios of Lithium, ORNL-3341 (March 1964), llp B, Thoma et al., Reactor Chem. Div. Ann. Progr. Rept, Jan. 31, 1965, ORNL-3789, p, 323. 12p, D, Thacher, Thermal Conductivity Studies of Phonon Scattering by Boundaries and Isotopes in Lithium Fluoride Crystals (thesis). Report 369, Ma- terials Science Center, Cornell Univ,, Ithaca, N.Y. (June 1965), 16. Chemical Support for the Controlled Thermonuclear Research Program R. A. Strehlow Whether controlled thermonuclear devices with useful power densities are feasible is a question which can be conveniently divided into two prin- cipal parts. The first part asks whether a plasma fuel element can be constructed; that is, whether the physics of plasma confinement can be solved’ in a practical way. The second part, which is more obscure, asks whether there is any fundamental bar {o construction of a thermonuclear reactor if the confinement problem is solvable. We have continued to study chemical aspects of both parts of this question. Our studies in support of the confinement prob- lems are directed toward development of tech- niques to diagnose the quality of the plasma environment in experimental devices and to assist in attempts to improve this quality. Such pursuits have dominated our chemical studies.!'™® In addi- tion, an extensive literature survey and an experi- mental study have been devoted to possible prob- lems of fritium inventory and the fuel cycle of a thermonuclear reactor. Some of our special equip- ment was also used to assist in the design (by ORNL Isotopes Division personnel) of an isotope separator of a new and improved type. Each of these portions of our effort is described briefly in the following. lReactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNI~3789. 2rhermonuclear Div. Semiann. Progr. Rept. Apr. 30, 1964, ORNL-3652, pp. 13845, 3Thermonuclear Div. Semiana. Progr. Rept. Oct. 31, 1964, ORNL=3760, p. 95, 127 YACUUM ANALYSIS IN AN EXPERIMENTAL PLASMA DEVICE R. A, Strehlow A simple mass-spectrometric assembly has been installed and operated routinely to yield a continu- ing record of vacuum conditions during operation of the plasma device DCX-2. About 2500 mass spectta have been obtained in this study since installation of this residual-gas analyzer in Feb- ruary 1965. Such data have proved helpful in assaying various pumping and operating parameters of DCX-2; such species as acetylene, trichloro- ethylene, methane, ethane, alcohols, and vapors of metals (such as Zn) can be detected unambig- uously. The mass spectra alone are insufficient to yield a real diaguosis of the DCX-2 vacuum environments since, in many cases, the origin of the observed species is more important than the fact of its presence. Accordingly, a study is under way of the effect of processes within DCX-2 upon the mass spectrum obtained from pertinent materials. For this study a second mass-analyzer assembly, similar to that on DCX-2, has been installed on a simple, large vacuum system whose construction materials, assembly methods, pumps, pump .oils, etc., are similar to those of DCX-2. This assembly pemmits study of the effects of various analyzer instrument parameters (electron-~accelerating volt- age and ion-source assembly procedure, for example) on the spectrum observed. More important, how- ever, it permits study of the effect of processes such as electron and ion bombardment, temperature fluctuations, titanium getter evaporation, and others used in DCX-2 on the mass spectrum ob- served. With this assembly, mass-spectrometric studies have been made of the effect of warming liquid- nitrogen-cooled surfaces,'*? of hot-filament reac- tions and electron bombardment of contaminated surfaces,? and of reactions of titanium in vacuum systems,®> Recent studies have dealt with the formation of products from bombardment with hydro- gen ions during glow discharges within the as- sembly. These studies suggest that acetylene is observed, in addition to methane, only when a trisiloxane oil is present in the bombarded materials. Acetylene, previously observed in DCX-2 during ion injection of I12+ into the assembly, is now be- lieved to be evidence of contamination by this pump-oil component, INTERACTION OF TRITIUM WITH THERMONUCLEAR-REACTOR MATERIALS S. S. Kirslis For a thermonuclear reactor to be feastble, the holdup of tritium in the metal of the containment vessel (probably molybdenum or tungsten) and in that of the ion sources (alloys of nickel or iron) must not result in excessive tritium inventory and decay losses. The equilibrium solubility of hydro- gen in these metals at low pressures is low. How- ever, the metals can be cathodically ‘‘surcharged”’ with hydrogen to amounts of the order of 100 c¢m? (NTP)/g; this is sufficient to cause, in some cases, blistering and cracking of the metal. In order to judge more clearly whether the metal in a reactor under bombardment by energetic tritium atoms might be surcharged with tritium, a literature study of the various possible surface and interior diffusion processes has been made; this survey will be published separately. The 50- to 100-kev pected to penetrate the metal to their range depth of 1 to 2 x 107* cm. They presumably diffuse inward (and also back to the entrance surface) according to normal diffusion laws. If there is no appreciable barrier to their egress at the sur- tritium atoms may be ex- face, the concentration of tritium atoms just in- side the metal will be low or zero. If all this is true, the steady-state concentration of tritium in D, M. Richardson and R. A. Strehlow, Trans. Natl. Vacuum Symp., 10th, 1963, p. 97. 128 the metal will have a peak at the range depth and will fall off to low values at the bombarded sur- face and at the far surface of the metal. The average concentration in the metal would be half the peak concentration. If a barrier to egress of tritium from the bombarded surface exists, the whole steady-state concentration curve would cor- respondingly rise to higher tritium concentrations, The factors determining the peak concentration of tritium are the flux of atoms to the metal, the dif- fusion constant of tritium in the metal, and the boundary condition for the degassing of the bom- barded metal surface. Information on hydrogen ditffusion in the metals of interest was derived mainly f{rom published permeability and solubility studies. For informa- tion on the degassing boundary condition, the literature was searched in the fields of chemisorp- tion and desorption, catalysis, hydrogen-electrode behavior, and hydrogen-atom recombination, Other peitinent information was obtained from published studies of ion bombardment, spuftering of metal surfaces, and saturation of metals by proton or deuteron bombardment. The conclusions of the literature study are as follows: 1. Hydiogen is easily degassed at low pressures and moderate temperatures, even from metals (such as tungsten) for which the chemisorption heats are high. Cathodic surcharging depends on the poisoning of the metal surface for hydrogen-atom recombi- nation. In a gaseous environment on clean metals, hydrogen-atom recombination is rapid. Diffusion rates for hydrogen in molybdenum at 600 to 800°C are sufficiently rapid to deliver injected atoms rapidly back to the bombarded surface, thus maintaining a very low (and from a practical viewpoint, negligible) concentration of hydrogen in the metal, The same is expected to be true for tritium. If the hydrogen concen- tration will indeed be as low as estimated from the simple diffusion model, it should cause no hydrogen embrittlement problems. Since the diffusion constants for nickel and iron are much higher than those of molybdenum, high internal hydrogen concentrations are not expected even at much lower temperatures. Some recent work on deuteron bombardment of metals indicates higher of deuterium just inside the metal surface than concentrations 129 predicted by the simple model. Some evidence indicates this may be the effect of surface contamination on the metal, HYDROGEN SURCHARGING OF MOLYBDENUM IN A GLOW DISCHARGE D. M. Richardson A study of hydrogen occlusion by a molybdenum cathode in a glow discharge has been conducted. The principal objective of this work was to assess the significance of hydrogen surcharging to design considerations for themmonuclear reactors. Initial studies had indicated a possible hydrogen content of one or more atmospheric cubic centimeters per cubic centimeter of metal after bombardment in a discharge. Though this concentration of hydrogen is less than 0.1 at. %, even this low value could, if it were typical of the concentration in a large fraction of the reactor, present serious problems to the reactor designer. The work described here and the literature study summarized above, how- ever, have led to the conclusion that, at elevated temperatures, occlusion of hydrogen is not large. At lower temperatures it appears that surface contamination can matkedly impede the recombina~ tion of hydrogen during even low-enetgy bombard- ment; very high hydrogen concentrations in the metal are able, therefore, to be achieved. For these studies, a 45-cm length of molybdenum wire 1 mm in diameler was used as a cathode in a cylindrical tank with a volume of 200 liters. The ends of the tank were electrically shielded with sheet Teflon, The center of the tank was cone nected to a vacuum system through a 6-in. isolation gate valve and a gate valve with a small orifice which had a speed of 2 x 107* the speed of the vacunm-system manifold., This allowed a steady introduction of hydrogen gas through a palladium leak to the glow discharge chamber at a pressure of 1072 to 1 torr with continuous mass analysis of the effluent gas from this chamber. The gas during the usual discharge contained not more than 1 part per thousand of nonhydrogen impurity. The conditions for the usual low-temperature discharge are summarized in Table 16.1. The total bombardment for the conditions shown was 12 amp-sec or about 1 to 3 atm-cm® of hydrogen gas. After opening the valve and pumping down to a Table 16.1. Conditions Employed in Experiments with Glow Discharges Wire volume, cm® 0.35 Wire area (apparent), em? 14 Pressure, torrs 0.25 Discharge current, ma 10 Discharge time, min 20 Applied voltage to cathode, v 500 pressure of about 1 x 1077 torr, the wire was heated to redness. Mass-spectrometric and ione gage observations were made of the hydrogen evolved by the depassing process. Pressute rises of as much as 6 x 10™* torr were observed during the 4-sec degassing procedure, Since the system pumping speed for hydrogen was about 1000 liters/ sec, this pressure rise corresponded to about 2 atm-cm® of hydrogen, Stringent cleaning of the wire and ion bombard- ment of the chamber walls for several hours, fol- lowed by repetition of the discharge, led to a value of only 0.017 atm-cm®. Continued repetitions of the discharge-degas cycle led to increasing amounts of hydrogen being occluded. Since the wire was not heated past 1000°C, a gradual increase in surface contamination could have been responsible for the very high values. Several attempts were made at higher current densities (and consequent high temperatures) to determine the amount of hydrogen occluded during a usual discharge. None of these attempts led to an evolution of hydrogen in detectable amounts, Molybdenum surfaces which are clean or which are heated do not seem to occlude large quantities of hydrogen during low~intensity bombardment with protons, MEASUREMENT OF GAS LOAD FROM SOURCE OF ELECTROMAGNETIC SEPARATOR R. A. Stehlow Scientists of the ORNL Isotopes Division are designing an improved electromagnetic isotope separator. This separator, a considerably modified calutron, is to have separate differential pumping systems for the source, collector, and main-vacuums- tank regions. For design of these pumping systems, 130 it was necessary to know the gas load upon each; that from the source was judged to require experi- mental measurement. ORNL electromagnetic separators use, where possible, vapor of the chloride of the element whose isotopes are to be separated., This chloride is generated within the source assembly by chlori- nation with CCl, of the appropriate metal or oxide. The gas load from the source assembly, therefore, consists almost entirely of unreacted CCl1 . A flow-rate meter with a constant (regulated) pressure and variable volume, capable of operation with condensable pases and at low pressures, was adapted for this study., This device, designed to measure flow rates as low as 10™° torr-liter/min for the thermonuclear support propram, was installed near the source region of a calutron. The CCI4 flow rate from this unit was monitored during a run in which the isotopes of cerium were being sep- arated. The measured flow rate, higher by nearly an order of magnitude than that previously esti- mated, has been used in design of the differential pumping system for the source of the new unit, ELECTRON ENERCY 5.4 15.3 14.7 14.9 ION INTENSITY 2.7 ELECTRON ENERGY FOR ETHAME FRAGMENTS (ev) 12.5 2.2 APPEARANCE-POTENTIAL MEASUREMENTS FROM TIME-OF-FLIGHT MASS SPECTROMETRY J. D. Redman Polymeric species are commonly observed in the vapor of simple and of mixed metal halides, We hope to assist in interpreting mass spectra from such halide vapors by determining appearance potentials of the various ions formed by electron impact with the wvapor species evolving from an effusion cell. A preliminary stage of this work has been completed with the construction of a retarding-potential circuit and its application to the study, with a time-of-flight mass spectrometer, This assessment was considered necessary, even though application of retarding-potential-difference methods of various gases to assess its reliability. to this type of spectrometer has been demonstrated, to determine whether gases introduced at the low pressures corresponding to the expected flux of salt-vapor species would yield appearance poten-~ tials with satisfactory precision. ORNL- DWG £6-980 FOR KRYPTCN (ev) 15.5 15.7 15.2 6.7 o \goa —— ) \\ A e ors S| s 0.44 ){/ ! : ! L 7 | | —— 1 134 133 135 3.7 139 44 143 145 Fig. 16.1. Normalized lonization-Efficiency Curves for Krypton and for C2H4+, CoHg +, and C?_Hé'+ from Ethane. The assembly was standardized by a series of careful appearance-potential (AP) measurements using krypton. We obtained the value 13.51 % 0,04 ev from these measurements; this is less by 0.5 ev than the generally accepted value obtained by other investigators.>”'! We have ascribed this discrepancy to (reproducible) contact potentials and, perhaps, other systematic errors in vur as- C. E, Melton and W, H, Hamill, J. Chem. Phys. (to be published). ®R. E. Fox et 2l., Phys. Rev. 89, 555 (1L953), D, C, Frost and C. A, MeDowell, Proc, Roy. Soc, (London), Ser. A 232, 227 (1955). 8]. F., Burns, Nature 192, 651 (1961). Y. Kaneko, J, Phys. Soc. Japan 16, 1587 (1961). 195, N, Foner and B. H. Hall, Phys. Rev. 122, 512 (1961). g, 1, Field and J. L. Franklin, Electron Impact Phenomena, Academic, New York, 1957, 131 sembly. Appearance potentials for other gases were obtained from ionization-efficiency curves, normalized for this systematic deviation at onset, obtained in the standardization with krypton. Typical normalized ionization-efficiency plots are shown in Fig. 16.1 for krypten and ethane frag- ments. Our values for other appearance potentials are compared in Table 16.2 with those obtained by others. We fail to check the published values for the C2H4+ and C21-15+ fragments from butane, but agree quite satisfactorily with published values for all others studied. Since the operating source pressure was main- tained at 5 x 1077 to 1 x 107 % torr in these studies, we believe that vapor fluxes from effusion cells containing metal halides will prove adequate for the planned investigation. It is likely that temper- ature control of the effusion-cell system will be of prime importance, 132 Table 16.2. lonization Potentials for Fragment Peaks of Argon, Nitrogen, Ethane, and Butane Mormalized for Yariation of Krypton AP from Literature Value Number of Fragment Reactant D . . Ionization Potentials (ev) Reference eterminations Art Ar 6 15.65 + 0.15 16.00 + 0.10 16.32 + 0.14 This work Art Ar 15.74 a ar’t Ar 15.77 b , * , 6 15.65 = 0.18 16.85 + 0.19 This work + , ) 15.60 b C,H, * C,H, 5 12.28 + 0.08 12.76 + 0,20 13.42 + 0.10 This work + . C,H, C,H, 12,10 12.70 13.20 c + C,H, C,H, 12.10 5 JHo * C,H, 3 12.80 + 0.2 13.39 + 0,14 13.78 + 0.23 This work + JH, C,H, 12.10 12.55 13.40 c + C,H, C,H, 12.80 b C,H, * C,H, 4 11.66 * 0.17 12.00 + 0,10 12.84 + 0.15 This work + C,H, C,H_ 11.60 12.10 12.65 c + C 1, C,H, 11,60 b , 4* C,H,, 2 12.26 * 0.24 12.86 + 0,1 13,52 + 0.3 This work + H, CH 11.40 B C2H5+ C,H, 3 12.75 +0.25 13.27 + 0,02 13.77 +0.09 This work + C,H, C,H, 12.10 b + e C,H, C,H 1 11.40 11.60 12.10 This work + C,H, C,H 11.50 11.80 12.20 c + C,H, C,H 11.70 B + \ ,{, _J‘._ - C,H,, C,H 4 10.68 + 0.22 11.11 + 0.20 This work + C,H, C,H,, 10.60 11.00 c + C,H,, C,H,, 10.80 b 2C. E. Mclton and W, H. Hamill, J. Chem. Phys. (to be published). bE. H. Field and J. L. Franklin, Electron Impact Phenomena, Academic, New York, 1957, °C, E. Melton and W, H. Hamill, J. Chem. Phys. 41(7), 546 (1964), Part V Nuclear Safety 17. Nuclear Safety Tests in Major Facilities FiSSION PRODUCTS FROM FUELS UNDER REACTOR-TRANSIENT CONDITIONS G. W. Parker R. A. Lorenz J. G. Wilhelm* Miniature fuel elements are melted in a special assembly in the TREAT to study the release of radioactive material when the fuel cladding and the fuel are melted or vaporized rapidly, as in a nuclear accident resulting from a reactor tran- The program attempts to measure and inter- pret the effect upon fission product release. of fuel type, cladding, atmosphere during the tran- sient, inventory of fission products, and charac- teristics of the transient. The extent of reaction of the cladding and of the UO _ fuel with the cover- ing atmosphere is also determined. sient. Behavior on Melting in Steam Analysis of two experiments which used atmos- pheres of 1000 psia steam (285°C) has been re- ported elsewhere.? One experiment used a spec- imen of UQ, with stainless steel cladding; during exposure in TREAT the cladding melted, oxidized extensively, slumped in a spongy mass, and ad- hered to the unmelted UO, pellets. Fission product release from fuel and cladding was much less than that in an earlier experiment, which used previously unirradiated stainless-steel-clad UO_ in 45-psi argon atmosphere, where the cladding flowed completely off of the unmelted U0, pellets. Appar- ently, the layer of oxidized cladding formed by the 10n assignment from Karlsruhe Center for Nuclear Research and Development, Karlsruhe, West Germany. 2. WL Parker, R. A. Lorenz, and J. G. Wilhelm, Nucl. Safety Program Semiann. Progr. . Rept. June 30, 19635, ORNL-3843, pp. 39-67. 135 high-pressure steam provided an effective barrier to fission product release. The second experiment nsed a specimen with Zircaloy-2 cladding on the UO, fuel; after the exposure, fuel and cladding were found to be fragmented and dispersed within the alumina crucible. Direct comparison of fission product release from the stainless-steel- and the Zircaloy- clad specimens cannot be made since plugged tubing prevented release of steam from the auto- clave after the transient with the Zircaloy-clad specimen. However, the distribution of fission ptoducts within the fuel autoclave showed that the release was greater for the Zircaloy-clad specimen; this must be considered at present to be a result of the fragmentation. Behovior on Melting Under Water The first two of a series of experiments have been performed with the TREAT reactor in order to study the release of fission products from fuel melted underwater. The experiments used stainless- steel- and Zircaloy-2-clad UQ, fuel specimens irradiated to 18 Mwd/metric ton. Aninitial specimen temperature of 70°C was used, and operating and reactor-transient conditions for the two experiments were identical. Heat input to the fuel during the transients. was approximately 500 cal per gram of UO2 (50‘?0: greater than any previously used); the U0, was heated to well above its melting point. The experiments were designed to simulate tran- sient accidents with water-cooled reactors in which steam and water are expelled from the core vessel. Valves were opened immediately after the transient, and the transient-generated steam and a purge of argon gas were permitted to flow through a con- denser, water-collection traps, filters, and a charcoal bed into a gas-collection tank. The fuel- containing autoclave was then electrically heated 136 to 350°C maximum for 1 hr to slowly boil excess . 2 . o vesers water out of the fuel autoclave. ' The release and distribution of fission products were essentially the same in the two experiments. The fuel specimens melted completely (Figs. 17.1 and 17.2). Fission product release to the fuel autoclave was high, but transport of fission products out of the autoclave by steam release and argon purge was relatively small. The '?°Te, '*7Cs, and 13! carried out of the fuel autoclave ranged from 2 to 7%. The transported release of non- volatile materials (°°Zr, '*'Ce, and UO ) was only 0.1%. Approximately 0.005% of the 311 reached the filter papers and charcoal bed. Fig. 17.1. Puffy-Appearing Cake of Melted Fuel and Cladding from Stainless-Steel-Clad UO?_ Melted Under Water by Transient Heat Input of 504 cal per Gram of UO2 in TREAT Experiment 7 (Front Malf of Crucible Remowved). ' . PHOTO B1912 Fig. 17.2. End Caps ond Melted Fuel and Cladding in Sample Holder from TREAT Experiment 8 in Which Zircaloy-2-Clad UC)2 Was Melted Under Water by Transient Heat Input of 511 cal per Gram of U02 (Flux Monitor Capsule in Place on Back Half of Crucible). An unmelted fuel specimen is shown dlongside for comparison. 137 Most of the transient-generated steam condensed in the fuel autoclave, and only a small amount was immediately discharged through the condenser. The water inside the fuel autoclave was an excellent trap for the fission products since slow evapora- tion of the remaining water (simulating accident after-heat) did not transport large amounts of fission products. Approximately 1 liter of hydrogen was produced in each experiment by reaction between c¢ladding and water; this resulted in a chemically reducing atmosphere for most fission products. Metallographic examination of portions of the melted fuel samples showed the formation of a eutectic phase in the specimen with stainless steel cladding. The major phase in the sample from the Zircaloy-2-clad specimen was a solid solution of approximately 75% UO, and 25% Zr:()z. Two additional experiments were recently per- formed with fuel specimens and reactor transients similar to those described above. The initial temperature was increased and the argon purge was eliminated in order to have quicker and more com- plete release of steam from the fuel autoclave in the first minute following the transients. Examina- tion of these experiments is in progress. FISSION PRODUCTS FROM SIMULATED LOSS-OF-COOLANT ACCIDENTS IN ORR W. E. Browning, Jr. C. E. Miller, Jr. W. H. Montgemery B. F. Roberts R. P, Shields 0. W. Thomas? A. F. Roemer? J. G. Wilhelm® Release of fission products and their subsequent behavior under a variety of conditions which might occur in loss-of-coolant accidents to nuclear re- actors is the subject of a continuing study. The loss-of-coolant accidents are simulated in a versa- tile experimental assembly in the Oak Ridge Research Reactor.®*® This assembly permits 3General Engineering and Construction Division. 4Analytical Chemistry Division. SW. E. Browning, Jr., ef al., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1961, ORNL-3127, pp. 149~52; Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1962, ORNI.-3262, pp. 172-76. ‘w. E. Browning, Jr., et al., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, p. 263. attainment of controlled temperatures (through nuclear heating of the small fuel specimen) to and above the melting point of UO, with any of sev- eral atmospheres. The release and the subsequent deposition of eight fission products (I, Te, Cs, Ru, Sr, Ba, Zr, and Ce) are determined in each of the experiments. Results of these investigations are published in detail elsewhere. ” The atmospheres that have been investigated include helium, moist helium, steam-helium-hydro- gen mixtures, dry air, moist air, and steam-air mixtures, Of the eight elements measured, only iodine and ruthenium showed significant variations in behavior with changes in atmosphere. lodine and ruthenium were both more volatile in a moist air than in any other atmosphere studied and were transported farther under these conditions. lodine passed through absolute filters but was retained on charcoal; ruthenium was more easily filtered in the volatile form (from moist air) than in the less volatile form. Ruthenium was not volatile in an atmosphere containing appreciable guantities of steam. The release of cesium from the 1000°C zone was lowest in experiments in which steam was a component of the atmosphere. Cladding material near the heated fuel is an abundant, strongly reducing reagent and can be expected to affect fission product behavior. Stain- less steel appears to retain ruthenium and, under oxidizing conditions, to lower the melting point of UO,. Experiments with Zircaloy cladding are in progress. Out-of-pile experiments* on the effects of stainless steel and Zircaloy cladding have shown that the efficiency of trapping iodine by steam condensation is increased by the presence of these materials. Three experiments have been performed to study the effect of high burnup of fuel (>>20,000 Mwd/ton) on fission product behavior. Suchfuel was examined metallographically and was found to be typical of high-burnup fuel used in power reactors of advanced design. Cesium and ruthenium were the only elements affected by burnup; the fractional release of ruthenium decreased with increasing burnup, while that of cesium increased. Effects of maximum temperature of the fuel are also being investigated. In these experiments, 7W. E. Browning, Jr., et al., Nucl. Safety Program Semiann. Progr. Reptf. June 30, 1965, ORNL-3843, pp. 3—39; and Nucl. Safety Program Semiann. Progr. Rept. Dec. 31, 1965, in preparation. performed under oxidizing conditions in which the maximum fuel temperature was maintained at ap- proximately 2000°C, the UO appeared to have melted due to the formation of a [ow-melting eutectic with stainless steel oxide. The releases of the eight elements in such intermediate-temperature experiments were similar to releases in experiments in which complete melting occurred at approximately 2900°C. Future experiments are planned at less than 2000°C and under reducing conditions. The primary purpose of these in-pile experiments is, of course, to permit prediction of behavior of the fission products in possible reactor accidents with reasonably assumed conditions. Accordingly, assessment of the form and the physical and chemical characteristics of the released materials is an essential part of the program. data with a fibrous filter suggest that particles of two size ranges are important in transport of fission product activity. Analysis by composite diffusion tubes shows three nonelemental iodine species in the gas. Studies with composite diffusion tubes have also shown that desorption of iodine from the apparatus after a simulated accident is small; only about 0.1% of the original inventory was desorbed, form. Preliminary and that primarily in a nonelemental The aging of fission product aerosols following release from the fuel may substantially alter the form and therefore the behavior of the fission products as a function of time. In order to investi- gate this effect, a simulated reactor containment vessel is being built into the in-pile facility, Future experiments with this extended facility will treat aerosol aging. FISSION PRODUCTS FROM HIGH-BURNUP UO, G. W, Parker R. A. Lorenz W. M. Martin C. J. Barton G. E. Creek In the first experiment of a series designed piimarily to test the effect of burnup on release and behavior of fission products, a re-irradiated specimen of stainless-steel-clad UO, previously irradiated to 1000 Mwd/ton was melted in the Containment Mockup Facility (CMF).® The specimen was melted by induction heating in a steam-air atmosphere; the released fission products were aged in a stainless steel tank filled with pres- surized steam-air mixture. Distributions of released ‘‘real’’ fission products are compared in Table 17.1 with those for simulated fission products from previous experiments in the CMF. The comparison tends to validate the use of simulated high-burnup UO fuel. Deposition of fission products before they reached the stainless steel aging tank was probably affected by the different furnace-tube geometry of this experiment 8G. W. Parker et al., Reactor Chem. Div, Ann. Progr. Rept. Jan. 31, 1965, ORNL-3789, p. 251. Table 17.1. Distribution of Fission Products Released from Simulated Fuel (8) and from High-Burnup U02 (HB) Fission Products Found (% of total inventory) L . Cesium Tellurium Ruthenium Strontium Iodine ocation o N e _ HB S HB S HB S 1B S HB S Fumace tube and duct 25.0 3.0 18.1 0.3 Q.26 0.07 0.047 0.01 15.0 10.1 Tank walls and deposition samples 8.5 15.1 12.4 6.9 0.054 0.35 0.015 0.04 27.0 33.8 Condensate 12.6 43.6 0.7 0.8 0.001 0.12 0.005 0.0003 57.6 56.0 Filters 2.1 0.8 0.5 0.45 0.002 0.0006 0.0001 0.0004 0.35 0.05 Total release from fuel 48.2 62.5 31.7 8.4 0.32 0.54 0.07 0.05 ~100 ~100 from that of earlier experiments. However, few differences in the distribution of fission products reaching the tank were noted. The volatile fission products, lodine, cesium, and tellurium, were the only activities other than rare pases reaching the tank in significant amounts; a large fraction of these activities remained on the tank wall or was collected in the condensate after 3 hr aging in the tank with gradually decreasing pressure and temperature. Two different types of samplers were used to study concentrations of gas-borne activity in the pressurized tank during the aging period. The two samplers gave comparable results, and the total airborne gamma activity (mainly from iodine, cesium, and tellurium) indicated by these samplers agreed well with gross-activity values from a collimated ion chamber adjacent to the side of the tank. Figure 17.3 shows the variation with time in concentration of the various gas-borne activities. ORNL-—-WG 66-254 N T RE PRODUCT INVENTORY IN TANK ATMOSPHI = 5 \ | D000 N e 2 . U e A L. < 5r i~ | =z it OO0 e e e e T . Lt 5 ‘ . 0.000M — 1 ; — 0 20 40 o0 80 100 120 140 TIME AFTER MELTING (min) Fig. 17.3. Composition of the Containment Mockup Facility Tank Atmosphere as Determined by Gas Samples. 139 Measurements were made of deposition rates of seven fission products on four different surfaces (stainless steel, carbon steel, and vinyl paint) during the experiment. Jodine showed a significant variation in rate of deposition on the different surfaces, but the others, with the possible exception of a faster rate for molybdenum on carbon steel and vinyl paint, showed no signif- icant variations. ( aluminum, THE CONTAINMENT RESEARCH INSTALLATION G. W. Parker W. J. Martin A schematic diagram of the Containment Research Instaliation (CRI) and a brief description of its principal features were given in the previous progress report.” All the major components of the CRI have been received, and construction is approaching completion. The primary simulation vessel, equipped with external resistance heaters (21-kw total capacity), has been mounted in the hot cell, and the major piping between the furnace and the containment vessel has been installed (see Fig. 17.4). Installation of the containment vessel has been delayed by a warpage problem (now solved) that temporarily prevented insertion of one of the re- movable liners. The liners will be painted with either an epoxy coating that is of interest to the LOFT facility at the NRTS or with a Bakelite coating of the type being considered for use in the Containment Systems Ezxperiment at Hanford. Stainless-steel-clad UO_ fuel specimens have been successfully melted in the Containment Mockup Facility using the pressurized induction furnace arrangement that will be used in the CRI. Modifica- tions of the furnace fittings to allow remote loading and unloading of highly irradiated fuel are presently being worked out. Wiring and instrumentation of the CRI, involving instrument panels on two floor levels, are about 75% complete. Preliminary testing of the CRI is expected to begin early in 1966. ' °G. W. Parker et al., Reactar Chem. Div. Ann. Progr, Rept. Jan. 31, 1965, ORNL-3789, pp. 259-61. 140 PHOTO 81557 . e —»‘n""lf?‘f"" R o LD Fig. 17.4. Primary Simulation Vessel of Containment Research Installation Mounted in Hot Cell. ANALYSIS OF PLANS FOR SCALE.UP IN NUCLEAR SAFETY PROGRAM C. E. Miller, Jr. W. E. Browning, Jr. The Nuclear Safety Program of the United States includes firm plans for large-scale experiments — of which the LOFT!? is the most ambitious — in which a reactor core would be destroyed and fis- sion product release and behavior would be meas- ured. Such experiments will be exceedingly complex and costly and will, obviously, be few in number. It will be essential, therefore, to gain from them the maximum information possible and to be able — with the help of well-designed smaller-scale studies — to apply this information to possible accidents under conditions other than those tested. At the request of AEC Washington we have attempted an evaluation of the research and development program at ORNL and elsewhere and of the varicus scaled-up experiments for which planning is firm. 11,12 are As a result of this evaluation, two reports in the final stages of preparation. We conclude that experiments to study fission product behavior in containment and in gas-cleaning systems are well scaled over the range between small laboratory experiments and LOFT, but that experiments to study the release and the transport of fission products from fuels are not well scaled, We suggest that intermediate-scale experiments of the otrder of 1 and 10% of the size of LOFT are 1OT. R. Wilson et al., An Engineering Test Program to Investigaie a Loss-of-Coolant Accident, ID0O-17049 (October 1964), e g, Mitler, Jr., and W. E. Browning, Jr., A4n Analysis of the Adequacy of Planned Experiments fo Test fthe Effects of Scale-Up in Reactor Accidents, Part [: Scale-Up in the U.5. Nuclear Safety Program, ORNL-3501, to be published. Vi, wm Miller, Jr., and W. E. Browning, Jr., 4n Analysis of the Adeguacy of Planned Experiments. to Test the Effects of Scale-Up in Reactor Accidents, Part I: Recommended Additional Nuclear Safety Scale- Up Experiments, in preparation. 141 needed. These intermediate experiments should measure fission product release from the fuel and study fission product behavior in the primary system following release. We would propose for consideration four experi- ments to bridge the pap between those described in the previous sections of this chapter and the LOFT assembly. These are: A. Tests at 1% of LOFT power level: 1. Investigation of fission product release and transport in the Containment Research Installation'?® primary simulator using a dummy core with irradiated fuel elements. These would be melted in place by central- tungsten-resistor methods. Investigation of fission product release and transport in an experimental research re- actor using a multiple-pin subassembly with irradiated fuel pins. These would be melted in place using nuclear heating induced by the parent reactor. B. Tests at 10% of LOFT power level: l. Investigation of fission product release and transport in the Containment Systems Exper- iment'* primary simulator using a dummy fuel core with irradiated fuel elements. These would be melted in place by cedtral- tungsten-resistor methods. 2. Investigation of fission product release and transport using a fuel subassembly of irradiated fuel elements in the LOFT f[acil- ity. These would be melted by nuclear self-heating induced by neutrons from the LOFT parent core. HG. W. Parker and W. J. Martin, Nucl. Safety Program Semiann. FProgr. Rept. June 30, 1965, ORNL-3843, po, G297, 14 G. J. Rogers, Program for Containment Systems Ex- periment, HW-83607 (September 1964). 18. Laboratory-Scale Supporting Studies RETENTION OF RADIOIODINE AND METHYL I0DIDE BY ACTIVATED CARBONS W. E. Browning, Jr. F. V. Hensle R. D. Ackley R. E. Adams J. E. Attrill? J. D. Dake? G. W. Parker D. C. Evans? G. E. Creek A. Ferreli® Its relatively high volatility and its great bio- logical hazard combine to make radioiodine a most important fission product in nuclear safety con- Elemental I, is relatively easy to trap and hold; the great convenience and safety of ambient-temperature activated-charcoal beds have led to their general adoption for removing iodine from air or gas streams. An increasing awareness, however, that a significant fraction of the iodine liberated in a nuclear accident might be present as organic iodides (notably methyl iodide) which are poorly retained by common activated charcoals has caused concern in the field for several years. Careful experiments have served only to con- firm that methyl iodide is poorly retained by standard charcoals under high humidity, Removal of methyl iodide by Pittsburgh type PCB activated charcoal, 12 x 30 mesh (unimpregnated for CH,I reaction), was investigated at 24°C; gas chroma- tography was used to determine the CHBI that passed the bed. The inlet methyl iodide concen- tration was varied (inversely with air velocity) from 8 to 80 mg of CH,I per cubic meter of air to provide a constant introduction rate, Time periods over which useful removal efficiencies were ob- gsiderations. 1Analytical Chemistry Division. 2Co-op student, University of Tennessee, 3Noncitizen guest on assignment from the Italian National Committee for Nuclear Energy. 142 served varied from less than 0.2 hr for a relative humidity near 100%, an air velocity of 100 fpm, and a bed depth of 1 in., to about 100 hr for a relative humidity of less than 3%, an air velocity of 10 fpm, and a bed depth of 3 in.**> Removal of methyl iodide and of elemental iodine by a variety of charcoals was investigated with the airecharcoal system at around 100°C and with the air stream about 80% saturated with steam. In these tests the iodine which passed the bed was determined by radioactivity measurements. Elemental iodine was trapped with fairly high efficiency, but methyl iodide was not. Both radiation detection and gas chromatography were used to test a method based on dehumidification for improving the efficiency of methyl iodide removal from steam-air mixiures. This method was observed to be successful but would require a rather complex moisture-removal sys;t*em.6 The recent obsetvation’ "2 that beds of a com- mercial charcoal (Mine Safety Appliances Company No. 85851) which is impregnated with several percent by weight of iodine will effectively remove methyl iodide is, therefore, advance. Subsequent tests have shown that other charcoals when impregnated with I, or KI are equally effective, We confirmed the observations a most important ‘w. E. Browning, Jr., et al., Nucl. Safefy Program Semiann. Progr. Rept. June 30, 1965, ORNIL-3843, pp. 139—-43, *R. E. Adams et al., Nucl., Safety Program Semiann. Progr. Rept, Dec. 31, 1965, to be issued. 6W. E. Browning, Jr., et al., Nucl. Safety Program Semiann. Progr., Rept., June 30, 1965, ORNIL.-3843, pp. 143—-48, 7G. W. Patker and Alberto Ferreli, Nucl. Safety Program Semiann. Progr. Repft, June 30, 1965, ORNL- 3843, pp. 194-210, 8G. W. Parker et al., Nucl. Safefy Program Semiann, Progr. Rept. Dec. 31, 1965, to be issued. IR, E. Adams et al., Nucl., Safety Program Semianit. Progr. Rept. Dec. 31, 1965, to be issued. ORNL~DWG 65 -11435RZ 00 10 : —— L O Yo PENETRATION 0.0 Q.00 2E72S UNTREATED 25725 + 1, (10 mg/g) 25725 + £, 0. (50 mg/q) PCB -+ 1, (70 mg/g) 25725 + TE.D. (850 mg/q}+KI (10 mg/y) BCB + KI (10 mgsy) PCB-FKI WO mg/g)t 12(10 g /) 2D = TRIETHYLENEDIAMING 0,000 0 G 02 Q.3 RESIDENCE TIME (sec) Fig. 18.1. at 25°C and 70% Relative Humidity with o 20
989.99% on the 21M. D, Silverman and W. E. Browning, Jr., Science 143, 572 (1964). 220, D. Silverman et al.,, Nwucl. Safety Program Semiann. Progr. Repf. Dec. 31, 1965, to be issued, B present address: Dow Chemical Company, Midland, Mich, 4w, E. Browning, Jr., RE. E. Adams, and G. L. Kochanny, Jr., Reactor Chem. Div. Ann. Proge. Rept. Jan. 31, 1965, ORNL~-3789, p. 288, QSW. E. Browning, Jr., et al., Nucl. Safety Program Semiann. Progr. Rept. June 30, 1965, ORNL~3843, pp. 148--56. Flanders filter, while the more penettating group is trapped at a slightly lower efficiency of >99,9%. Current using aerosols generated from UO, clad with either neutron-irradiated stain- less steel or Zircaloy have produced results similar to the earlier studies. Similar values were ob- tained when air velocity ranged from 3.2 to 13 fpm and when dry helium or argon was substituted for the air atmosphere. However, in a recent series of experiments26 in which the relative humidity of the air sweep ranged from approximately 46 to 100%, the filter efficiency was noted to drop to as low as 93%. Fibrous-filter deposition profiles and photomicro- graphs of an aerosol produced in a 46% relative humidity experiment indicate that the filtration experiments characteristics of the particles of the aerosol are changed. Future experiments will attempt to de- termine the role which water vapor plays in this observed reduction in filter efficiency, IGNITION OF CHARCOAL ADSORBERS BY FISSION PRODUCT DECAY HEAT B. F. Roberts R. P. Shields W. E. Browning, Jr. C. E. Miller, Jr. In the event of a nuclear accident in which fise sion products are released into the containment system, the charcoal adsorbers used for removal of 26R. E, Adams, J. S. Gill, and W. E, Browning, Jr., Nucl, Safety Program Semiann. Progr. Rept. Dec. 31, 1965, to be issued, 146 iodine are subject to heating by decay of iodine adsorbed on the charcoal. The heat generated in pieces of charcoal containing the greatest amount of adsorbed iodine may raise their temperature to the ignition point. In order to simulate such an event, an in-pile experiment that utilizes the f{uel- melting facility in the Qak Ridge Research Re- actor has been designed, In this experimeat the decay of short-lived iodine adsorbed on the char- coal generates heat. The objective of this ex- periment is to measure and compare the ignition temperature of the charcoal with and without fis- sion product adsorption. A comparison of the cal- culated decay-heat load on the front surface of the charcoal in the experimental adsorber with those of adsorbers at various reactors indicates that the ignition effects in the reactor adsorbers will be smaller than those in the experiment. After the mechanical design and plans for operat- ing the experiments were shown to be adequate by tests in a laboratory mockup,?” an initial series of in-pile experiments was performed.?® In these experiments the ignition temperature was studied as a function of both the air flow and the decay- heat loading., Preliminary results of these studies indicate that the charcoal ignition temperature is not greatly affected by the decay heat of adsorbed fission products, Tw. E. Browning, Jr., et al., Nucl. Safety Program Semiann. Progr. Repfi. June 30, 1965, ORNL-3843, pp. 15667, 28W. E. Browning, Jr., et al., Nucl. Safety Program Semiann., Progr. Rept. Dec. 31, 1965, to be issued. AUTHOR(s) Bacarella, A, L., and A, L. Bution Baes, C, F., Jr. Baes, C, F,, Jr., N. J. Meyer, and C. E. Roberts Bamberger, C. E. L., H. ¥. McDuffie, and C. ¥. Baes, ]Jr. Barton, C. J., and W. B. Cottrell Browning, W. E., Jr. Brunton, G, D. Burns, J. H. Burns, J. H., and W. R, Busing Cantor, 8, Carroll, R. M. Carroll, R, M., R. B. Perez, and O. Sisman Caveol]l, R, M,, and P. E. Reagan Carroll, R. M., and O. Sisman Publications JOURNAL ARTICLES TITLE Electrochemical Measurements on Zirconium and Zircaloy-2 at Elevated Temperatures, 2. 200-300°C The Reduction of Potentiometric Hydrolysis Data The Hydrolysis of Thorium(IV) at 0..95° Purification of Beryllium by Acetylacetone- EDTA Solvent Extraction — Procedure and Chemistry Fission Product Refease and Transport Removal of Radiviodine from Gases Crystal Structure of Zirconium Tetrafluoride Crystal Structure of Hexagonal Sodium Neodymium Fluoride and Related Compounds Crystal Structures of Rubidium Lithium Fluoride and Cesium Lithium Fluoride The Vapor Pressures of Beryllium Fluoride and Nickel Fluoride Fisgion Product Release from Fuels Release éf Fission Gas During Fissioning of UC)2 Techuniques for In-Pile Measurements of Fission Gas Release In-Pile Fission-Gas Release from Single Crystal UO:7 147 PUBLICATION J. Electrochem. Soc. 112(6), 546 (1965) Inorg. Chem. 4, 588 (19653) Inorg. Chem. 4, 518 (1965) Nucl. Sci. Eng. 22(1), 14 (1965) Nucl. News 8{7), 25 (1965) Nucl. Safety 6(3) 272 (1965) J. Inorg. Nucl. Chem. 27, 1173 (1965) Inorg. Chem. 4, 881 (1965) Inorg. Chem. 4, 1510 (1965) J. Chem. Eng. Data 10{3), 237 (1965) Nucl. Safety 7(1), 34 (19653) J. Am. Ceram. Soc. 48(2), 55 (1965) Nucl. Sci. Eng. 21(2), 141 (1965) Nucl. Sci. Eng. 21(2), 147 (1963) AUTHOR(s) Carroll, R. M., and O. Sisman de Bruin, H. J., and G. M. Watson Fuller, E. L., H., F. Holmes, and C. H. Secoy Hoimes, H, F,, C. S. Shoup, Jr., and C, H. Secoy Johnson, G. L., M. J. Kelly, and D, R, Cuneo Keilholtz, G. W., J. E. Lee, Jr., R. E. Moore, and R. L.. Hamner Malinauskas, A. P,, and F, L. Carlsen, Jr. Marshall, W, L., and Ruth Slusher Miller, C, E., Jr., and W, ¥, Hillsmeier Osborne, M. F,, E. L. Long, Jt., and J. G. Morgan Overholser, L.. G., and J. P, Blakely Parkinson, W, W., Jr., C. D. Bopp, D. Binder, and J. E, White Quist, A, S,, and W, L., Marshall Quist, A, S,, W. L. Marshall, and . R. Jolley 148 TITLE In-Pile Fission-Gas Release from Fine Grain UO2 Self-Diffusion of Beryllium in Unirradiated Beryllium Ozxide Gravimetric Adsorption Studies of Thorium Dioxide Surfaces with a Vacuum Microbalance FElectrokinetic Phenomena at the Thorium Oxide---Aqueous Solution Interface Reactions of Aqueous Thorium Nitrate Solutions with Hydrogen Peroxide Behavior of BeO Under Neution Irradiation Gas Transport Characteristics of Uraniume- Fueled Graphites Aqueous Bystems at High Temperature, XV, Solubility and Hydrolytic Instability of Magnesium Sulfate in Sulfuric Acid-Water and Deuterosulfuric Acid—Deuterium Oxide Solutions, 200° to 350°C Second AEC Conference on Radigactive Fallout from Nuclear Weapons Tests Performance of Prototype EGCR Fuel Under Extreme Conditions Oxidation of Graphite by Low Concentrations of Water Vapor and Carbon Dioxide in Helium A Comparison of Fast Neutron and Gamma Irradiation of Polystyrene., Part I. Cross- L.inking Rates Assignment of Limiting Equivalent Conduct- ances for Single Ions to 400° Estimation of the Dielectric Constant of Water to 800° Electrical Conductances of Aqueous Solutions at High Temperature and Pressure. [I, The Conductances and lonization Constants of Sulfuric Acid—Water Solutions from 0—800°C and at Pressures Up to 4000 Bars PUBLICATION J. Nucl. Mater. 17(4), 305 {1965) J. Nucl. Mater. 14, 239 (1964) Vacuum Microbalance T'ech. 4, 109 {1965) J. Phys. Chem. 69, 3148 (1965) J. Inorg. Nucl. Chem. 27, 1787 (1965) J. Nucl. Mater. 14, 87 (1964) Am. Ceram. Soc. Bull. 44, 251 (1965) J. Chem. Eng. Data 10, 353 (1965) Nucl. Safety 6(3), 283 (1965) Nucl. Sci. Eng. 22(4), 420 (1965) Carbon 2, 385 (1965) J. Phys. Chem. 69, 828 (19653) J. Prys. Chem. 69, 2084 (1945) O J. Phys. Chem, 69, 3165 (1965) J- Phys. Chem. 69, 2726 (1965) AUTHOR(s) Reagan, P. E., J. G. Morgan, and Q. Sisman Robbins, G. D., R. E. Thoma, and H. Insley Singh, A. J., R. G. Ross, and R. E. Thoma Thema, R. E., H. Insley, H. A, Friedman, and G. M. Hebert Adams, R. E., and W. E, Browning, Jr. Adams, R. E., W, E. Browning, Jr., Wm, B, Cottrell, and . W, Parker Baumann, C, D, Browning, W, E., Jr., and M. E, Davis Brunton, G. D., H. Insley, T. N. McVay, and R, E. Thoma Clark, W. E., L. Rice, and D. N, Hess English, J. L., and J. C. Griess Haynes, V., O., . H, Sweeton, and D, E, Tidwell Jenks, G, H. 149 TITLE Fission-Gas Release from Pyrolytic-Carbons Coated Fuel Particles During Irradiation at 2000 to 2500°F Phase Equilibria in the System CstZrF4 Vacuum Disiillation of LiF The Condensed System LiF-NaF-ZrF4 — Phase Equilibria and Crystallographic Data REPORTS ISSUED Iodine Vapor Adsorption Studies for the NS “Savannah® Project The Release and Adsorption of Methy! Iodide in the HFIR Meaximum Credible Accident {rradiation Effects in the EGCR Fuel A Standard Surface for Fission Product Deposition Experiments Crystallographic Data for Some Metal Fluorides, Chlorides, and Oxides Evaluation of Hastelloy I and Other Corrosion-Resistant Structural Materials for a Continuvons Centrifuge in a Multi- purpose Fuel-Recovery Plant Laboratory Corrosion Studies for the High Flux Isclope Reactor Irradiation Deta for the Armmy PM Fuel Experiment in the ORR Pressurized-Water Loop for ORR Cycle 52 Irradiation Data for the Army PM Fuel Experiment in the ORR Pressurized-Water Loop for ORR Cycles 53 and 54 Effects of Reactor Operation on HFIR Coolant PUBLICATION Nucl. Sci. Eng. 23, 215 {1965) J. Inorg. Nucl, Chem. 27, 559 (1965) J. Appl. Phys. 36(4), 1367 (1965) J. Chem. Eng. Data 10(3), 219 (1965) ORNI.-3726 (February 1965) ORNL=TM=1201 (October 19635) ORNL.-3504 {June 1965) ORNL-TM-1304 (October 1065) ORNL~3761 (February 1065) ORNL=3787 (April 1965) ORNLTM-102%2 (June 1965) ORNI~-TM-083 (I"ebruary 1963) ORNL-TM-1034 (April 1965) ORNI.-3848 (October 1965) AUTHOR(s) Jenks, G. H., E. G, Bohlmann, and J. C. Griess Keilholtz, G. W., and C. J. Barton Mason, E. A,, and A, P. Malinauskas Mathews, A, L., and C, I, Baes Rabin, 8, A., J. W, Ullmann, E, L.. T.ong, Jr., M. F. Osborne, and A, E. Goldman Redman, J. D, Reed, S. A., and J. C. Movers Savage, H. W,, E. L. Compere, W. R. Huntley, R. E. MacPherson, and A, Taboada Singh, A. J., R. G. Ross, and R, E. Thoma Thoma, R, E, Yeatts, L. B,, Jr., and W, T. Rainey, ]Jr. 150 TITLE An Evaluation of the Chemical Problems Associated with the Aqueous Systems in the Tungsten Water Moderated Reactor, Addenda 1 and 2 Behavior of Iodine in Reactor Containment The Effect of Accommodation on Thermal Transpiration: Limitations of the **Dusty= Gas®® Model in the Description of Surface Scattering Oxide Chemistry and Thermodynamics of Molten Lithium Fluoride--BeryIlium Fluoride by Equilibration with Gaseous Water— Hydrogen Fluoride Mixtures Irradiation Behavior of High Burup ThOQ— 4.45% UO2 Fuel Rods A Literature Review of Mass Spectrometric— Thermochemical Technique Applicable to the Analysis of Vapor Species over Solid Inorganic Materials Estimation of Annual Operating and Maintenance Costs of Dual Putpose Nuclear- Electric Sea Water Conversion Stations SNAP=8 Corrosion Program Quarterly Progress Report for Period Ending November 30, 1964 SNAP-8§ Corrosion Program Quarterly Progress Repott for Period Ending February 28, 1965 SNAP-8 Corrosion Program Quarterly Progress Report for Period Ending May 31, 1965 Zone Melting of Inorganic Fluorides Rare-Farth Ralides Purification of Zirconium Tetraflyoride PUBLICATION ORNL-TM-278 (March 1965) ORNL-NSIC-4 (February 1965) ORNL-3796 (April 1965) ORNL-TM-1129 (May 1965) ORNI.-3837 (October 1965) ORNL-TM-089 (June 1965) ORNL-TM-1057 (April 1965) ORNI.-3784 (March 1965) ORNI.-3823 (June 1965) ORNI.-3859 (September 1965) ORNL-3658 (July 1965) ORNL-3804 (May 1965) ORNL-TM=1292 (November 1965) AUTHOR(s) Adams, R. E., and W. E. Browning, Jr. Browning, W. E., Jr., R. E. Adams, K. D. Ackley, M. E, Davis, and J. E. Attrill Carroll, R. M., and O. Sisman Compere, E. L., and J. E. Savolainen Davis, M. E., W. E. Browning, Jr., G. E. Creek, G. W. Parker, and L. F. Parsly Joncich, M. J., and . F. Holmes Kelly, M, J. Miller, C. E., Jr., W. E. Browning, Jr., R. P. Shields, and B. F. Roberts Parker, G. W., R. A, Loerenz, and J. G. Wilhelm 151 BOOKS AND PROCEEDINGS TITLE Removal of Iodine and Velatile Iodine Compounds from Air Systems by Activated Chearcoal Identity, Character, and Chemical Behavior of Vapor Forms of Radioiodine Fission-Gas Release During Fissioning in Uranium(IV) Oxide The Chemistry of Hydrogen in Liquid Alkali Metal Mixtures Useful as Nuclear Reactor Coolants. The Deposition of Fission-Product Iodine on I. NaK-78 Structural Surfaces Heat Effects at Electrodes (work performed at University of Tennessee) Removal of Rare Earth Fission Products from Molten Salt Reactor Fuels by Distillation In-Pile Fission Product Release Experiments - Atmosphere Effects Release of Fuels During Transient Accidents Simulated Fission Products from Reactor in TREAT PUBLICATION Proc. Intern. Symp. Fission Product Release and Transport Under dccident Conditions, Qak Ridge, Tenn., Apr. 5.-7, 1963, CONF-650407 (1965) Proc. Intem, Symp. Fission Product Release and Transport Under Accident Conditions, Qak Ridge, Tetn., Apr, 57, 1965, CONF-650407 (L1965) Trans. Am. Nucl. Soc, 8(1), 22 (1965) Trans. Am. Nucl. Soc. 8(1), 170 (1965) Proc. Intern. Symp. Fission Product Release and Transport Under Accident Condirions, Qak Ridge, Tenn., Apr, 5.-7, 1965, CONF-650407 (1965) Proc. 1st Australian Conf, Electrochem., Sydney and Hobart, Ausiralia, February 1963, Pergamoen, New York, 1964 Trans. Am. Nucl. Soc. 8(1), 170 (1965) Proc. Intern. Symp. Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tenn., Apr. 5.7, 1965, CONF-650407 (1965) Proc, Intem. Symp. Fission Product Release and Transport Under Accident Conditions, Qak Ridge, Tenn., Apr. 5.7, 1965, CONF-650407 (19653) AUTHOR(s}) Parker, G. W., W, J. Martin, G. E. Creek, and C, J. Barton Perez, R. B. Roberts, B, F., W. E, Browning, Jr., R. P. Shields, and C. E. Miller, Jr. Mathews, A. L. 152 TITLE Behavior of Radioiedine in the Containment Mockup Facility A Dynamic Method for In-Pile Fission-Gas Release Studies Effects of Atmosphere on Behavior of Fission Products Released by In-Pile Melting of UO2 THESIS Oxide Chemisfry and Thermodynamics of Molten Lithium Fluoride--Beryllium Fluoride by Eqguilibration with Gaseous Water~Hydrogen Fluoride Mixtures PUBLICATION Proc. Intern. Symp. Fission Product Release and Transport Under Accident Conditions, OQak Ridge, Tenn., Apr. 5—-7, 1965, CONF-650407 {1963) Trans. Am. Nucl. Soc. 8(1), 22 (1965) Trans. Am. Nucl. Soc. 8(1), 139 (1965) Thesis submitted in partial ful~ fillment of the requirements for Ph.D. degree, University of Mississippi, May 1965 Papers Presented at Scientific and Technical Meetings AUTHOR(s) Adams, R. E,, and W, E, Browning, ]Jr. Bacarella, A. L. Baes, C, F., I Baes, C, F., Jr., N. J. Meyer, and C. E, Roberts Bennett, R, L., H. L. Hewmphill, and W. T. Rainey, Jr. Bennett, R, L., H. L. Hemphill, W. T. Rainey, Jr., and G. W. Keilholtz Blakely, J. P., and .. G. Overholser Blankenship, F. F., il. F. McDuffie, R®. E. Thoma, and W. R. Grimes Bohimann, E. G., and F. A, Posey TITLE Removal of lodine and Volatile lodine Compounds from Air Systems by Activated Charcoal Anodic Film Growth on Zirconium at Tempers atures from 200--300°C The Chewmistry and Thermodynamics of Molten Salt Reactor Fluoride Solutions Acidity Measurements at Elevated Temper= atures. 2. Thorium(IV) Hydrolysis, 0—95°C Thermal EMF Drift of Refractory Metal Thermocouples in Pure and Slightly Contaminated Helium Atmospheres Stdbility of Thermoelectric Materials in a Helium-~Graphite Environment Oxidation of AT]J Graphite by Low Concentra- tions of Water Vapor and Carbon Dioxide in Helium Molten Fluorides as Nuclear Reactor Fuels Aluminum and Titanium Corresion in Saline Waters at Elevated Temperatures 153 PLACE PRESENTED International Symposiam on Fission Product Release ard Transport Under Accident Conditions, Dak Ridpe, Tenn., Apr, 5~7, 1965 Electrochemical Society Meeting, Buffalo, N.Y., Oct. 1114, 1965 TAEA Symposium on Thermo= dynamics with Emphasis on Nuclear Materials and Atomic Transport in Solids, Vienna, Austria, July 2227, 1965 American Chemical Society, Detroit, Mich., Apr, 49, 1965 AEC High-Temperature Thermome etry Meeting, Washington, D.C,, Feb. 24-26, 1965 AEC High-Temperature Thermom- etry Meeting, Washington, D.C., Feb. 2426, 1965 Conference on Carbon, 7th Biennial, Cleveland, June 21-25, 1965 Electrochemical Society, San Francisco, Calif,, May 9--13, 1965 1st International Symposium on Water Desalination, Washington, D.C., Oct. 3—9, 1965 AUTHOR(s) Browning, W. E., Jr., R. E. Adams, R. D, Ackley, M. E. Davis, and J. E. Attrill Buras, J. H. Buras, J. H., and E. K. Gordon Carroll, R. M. Carroll, K. M., and PP. E. Reagan Carroll, R. M., and O, Sisman Compere, E. L., and J. E. Savolainen Dabbs, J. W. T., F. J. Walter, and G. W. Parker Davis, M. E,, W, E, Browning, Jr., G. E. Creek, G. W, Parker, and L. F. Parsly Davis, R. J. Grimes, W. R. 154 TITLE Identity, Character, and Chemical Behavior of Vapor Forms of Radioiodine Crystal Structure Analysis Applied to Inorganic Fluorides Refinement of the Structure of Lithium Tetrafluoroberyllate The Behavior of Fission-Gas in Fuels In-Pile Performance of High-Temperature Thermocouples Fission-Gas Release During Fissioning in Uranium(IV) Oxide The Chemistry of Hydrogen in Liquid Alkali Metal Mixtures Useful as Nuclear Coolants. 1. NaK-78 Saddle Point Rotational States from Resonance Fission of Oriented Nuclei The Deposition of Fission-Product Iodine on Structural Surfaces Oxide Growth and Capacitance on Preirradiated Zircaloy-2 Chemistry of Molten Fluoride Reactor Systems Molten Fluorides as Reactor Material The Chemistry of the Molten Salt Reactor PLACE PRESENTED International Symposium on Fission Product Release and Transport Under Accident Conditions, Qak Ridge, Tenn., Apr. 5-7, 1965 ORINS Traveling Lecture Program, University of Arkansas, Fayetteville, Dec. 6, 1965 American Crystallographic Association, Gatlinburg, Tenn., June 27—July 2, 1965 AIME Conference on Radiation Effects, Ashville, N.C,, Sept. 8-10, 1965 AEC High-Temperature Thermometry Meeting, Washing- ton, D.C., Feb. 24--26, 1965 American Nuclear Society, Gatlinburg, Tenn., June 2124, 1965 American Nuclear Society, Gatlinburg, Tenn. June 21-24, 1945 IAEA Symposium on the Physics and Chemistry of Fission, Salzburg, Austria, March 2226, 1965 International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tenn., Apr. 5-7, 1965 Electrochemical Society, San Francisco, Calif,, May 6-—13, 1965 Gordon Research Conference on Molten Salts, Meriden, N.H., Aug. 30—Sept. 3, 1965 Chemistry Seminar at Argonne National Laboratory, Nov, 12, 1965 Nuclear Engineering Colloquium at Brookhaven National Lab- oratory, Dec. 2, 1965 AUTHOR(s) Grimes, W. R, Karraker, R. H., and R. E. Thoma Kelly, M. J. Keyser, R. M. Marshall, W. L. Marshall, W. L.,, and E. V. jones Mathews, A. L., and C, F, Baes, Jr. McDuffie, H, F., and R. P. Atkinson Miller, C. E., Jr., W. E. Browning, Jr., R. P. Shields, and 3. F. Roberts Parker, G. W, R. A. Lorenz, and J. G. Wilhelm Parker, G. W., W. J. Martin, G. E. Creek, and C. J. Barton Parkinson, W. W. 155 TITLE Molten Fluorides as Fuels, Coolants, and Blankets for Nuclear Reactors Sodium Fluoride—Scandium Fluoride Phase Equilibria Removal of Rare Earth Fission Products from Molten Salt Reactor Fuels by Distillation Experiments in Chemonuclear Synthesis at ORNI. Solubility and HKlectrical Conductance Meas- urements of Some Aqueous Electrolytes of Interest to Oceanography Second Dissociation Constant of Sulfuric Acid from Solubilities of Calcium Sulfate in Sule furic Acid Solutions Oxide Chemistry and Thermodynamics of Molten Lithium Fluoride~Beryilium Fluoride The Development of an Information Center for Molten Salt Chemistry In-Pile Fission Product Release Experiments — Atmosphere Effecis Release of Fission Products from Reactor Fuels During Transient Accidents Simulated in TREAT Behavior of Radiociodine in the Containment Mockup Facility The Effect of Radiation on Plastics and Rubber PLACE PRESENTED Chemistry Seminar at Purdue University, Lafayette, Ind., Dec. 9, 1965 American Chemical Society, SW-SK Regional Meeting, Memphis, Tenn., Dec. 2—4, 1965 American Nuclear Society, Gatlinburg, Tenn., June 21-24, 1965 Chemenuclear Review Meeting at Onchicta Conference Center, Tuxedo, N.Y., Oct, 19--20, 1963 Conference on Kinetics and Mechanism in Aqueous Inorganic Systems, Miami, Fla., Nov, 18--23, 1965 American Chemical Society, Detroit, Mich., Apr. 4--9, 1965 American Chemical Society, SW-SE Regional Meeting, Memphis, Tenn., Dec. 2~4, 1965 Tennessce Academy of Science, Qak Ridge, Tenn., Dec. 10--11, 1965 International Symposium on Fission Product Release and Transport Under Accident Conditions, Qak Ridge, Tenn., Apr. 57, 1965 International Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tenn., Apr, 5--7, 1965 International Symposinm on Fission Product Release and Transport Under Accident Conditiofis, QOak Ridge, Tenn., Apr. 5--7, 1965 U.5. Army Nuclear Science Seminar, Oak Ridge, Tenn., July 27, 1965 AUTHOR(s) Parkinson, W, W,, and W. D. Burch Perez, R. B., and G. M. Watson Quist, A. S. Quist, A, S., and W. L. Marshall Reed, S. A. Roberts, B. F., wW. E, Browning, ]Jr., R. P. Shields, and C. E. Miller, Jr. Rutherford, J. L., J. P. Blakely, and I.. G, Overholser Secoy, C. H,, H. F. Holmes, C. S. Shoup, and E. L., Fuller Soldano, B, A. Thoma, R. E, 156 TITLE Synthesis of Organics Through Isotopic Decay Radiation A Dynamic Method for In-Pile Fission=-Gas Release Studies Electrical Conductances of Aqueous Potassium Bisulfate and Sulfuric Acid Solutions to 800°C and 4000 Bars Electrical Conductances of Aqueous Potas=- sium Bisulfate and Sulfuric Acid Solutions to 800°C and 4000 Bars Some Chemical and Materials Problems in Sea Water Conversion Plants Effects of Atmosphere on Behavior of Fission Products Released by In-Pile Melting of UO2 Oxidation of Fueled and Unfueled Graphite Spheres by Steam The Interaction of Water with the Surface of Ceramic Oxides The Osmotic Behavior of Agueous Solutions at Elevated Temperatures Methods of Investigating High Temperature Phase Equilibria Rare Earth Trifluoride Systems PLACE PRESENTED Annual Contractors Conference, AEC Division of Isotopes Development, Brookhaven Mational Laboratory, Oct. 78, 1965 American Nuclear Society, Gatlinburg, T'enn., June 21-24, 1965 Karlsruhe Technische Hochschule, Karlsruhe, Germany, Sept. 13—15, 1965 Central Electricity Research l.aboratories, L.eatherhead, England, Sept. 21, 1965 Risgd Research Establishment, Risg (Roskilde), Denmark, Sept. 27, 1965 CITCE Symposium on Electro= chemistry at High Temperatures, Budapest, Hungary, Sept. 5~10, 1965 American Chemical Society Regional Meeting, Dayton, Chio, Dec. 14, 1965 American Nuclear Society, Gatlinburg, Tenn., June 21-24, 1965 Conference on Carbon, 7th Biennial, Cleveland, June 21-25, 1965 International Conference on Tropical Oceanography, Miami Beach, Fla., Nov. 17—24, 1965 ORINS Traveling L.ecture Program, Furman University, Greenville, S,C,, Nov. 1, 1965 ORINS Traveling Lecture Program, Furman University, Greenville, 5.C,; Mar, 1, 1965 Chemistry Seminar at Arpgonne National Laboratory, May 21, 1965 AUTHOR(s) Truitt, Jack, J. ©. Stiegler, and R. B. Evans IIi Vanderzee, C. E.,l and E. L. Fuller, Jr. Watson, G. M. Yeatts, L. B., Jr., and W, L. Marshall 157 TITLE Actinide Migration in Pyrocarbons as Influenced by Actinide and Defect Concentrations Thermochemistry and Kinetics of the Hydrolytic Decomposition of Ammonium Carbamate and Ammonium Dithiocarbamate The Economics of Nuclear Power Chemical Aspects of Nuclear Safety Physico-Chemical Problems of Nuclear Reactors Solubilities of Calcium Hydroxide and Saturation Behavior of Calcium Hydroxide — Calcium Carbonate Mixtures in Aqueous Sodium Nitrate Seolutions at High Temperatures IResearch performed at University of Nebraska, Lincoln. PLACE PRESENTED American Ceramic Society, Philadelphia, May 2--5, 1963 First Midwest Regional Meeting, American Chemical Society, Kansas City, Mo., Nov. 4-5, 1965 AEC International Exhibit at San Salvador, El Salvador, Mar, 1922, 1665 AEC International Exhibit at San Salvador, El Salvador, Mar. 19..22, 10965 ORINS Traveling Lecture Program, Tarleton State College, Stephenville, Tex,, Deec. 6, 1965 American Chemical Society, SW-SE Regional Meeting, Memphis, Tenn,, Dec. 2—4, 1965 T 9-.58. 59, 60. 61. 62. 63-77. 78. 79. 80. 61, 82, 83. 84. 85. 86. 87. 88. 89, 90. 91. 92. 93. 94. Q5. 96. 97. 98. 99, 100. 101. 102. 103. 104. 105. 106. 107. 108. 109. 159 INTERNAL DISTRIBUTION Bioclogy Library _ Central Research Library Reactor Division Library Laberatory Shift Supervisor ORNL Y-12 Technical Library Document Reference Section Laboratory Records Department L.aboratory Records, ORNL R.C. . E. Larson M. Weinberg G. MacPherson E. Boyd R. Bruce L H . Culler . Jordan . Briggs . Cottrell . Fraas . Krous . Lane Milier . Trauger . Whitman . Beall Billingten . Ferguson Frye, Jr. . Kelley . Taylor . Bredig . Corbin . Cunningham . Crawford . King . Lyon . Parker Skinner . White . Williams PEETAAME-TFEZMEI-C00YVQO0PERPEAX>ZIPOF>ETTTOT>O0O ITHAPIAIMOMOTS->>0EEC NnO-wmzZx 110. 111 112, 113. 114. 115. 116. 117. 118. 119. 120, 121. 122. 123, 124. 125. 126. 127. 128. 129. 130. 131. 132. 133. 134. 135. 136. 137. 138. 139. 140. 141. 142. 143, 144, 145. 146. 147. 148. 149, 150, 151. 152. 153. X PCUIARDYPEOrEBONEPALEOOICMALIMAIACOEOOPO0MROTME OUIXA>PIPLPETEQOMMI US>V X OITOmMmMUOS~rrmmImIT e A TTMOFBr-r< ORNL-3913 UC.4 — Chemistry TID-4500 (47th ed.) . Grimes . Bohlmann . McDuffie . Watson . Blankenship . Secoy . Baes, Jr. . Bacarella Barton . Bopp . Browning, Jr. . Brunton . Burns antor . Carroll . Compere Davis . English . Evans 1] . Fuller, Jr, . Griess . Holmes . Jenks . Keilholtz Kelly Kirslis . L.orenz . Malinauskas . Marshall . Miller . Moore . Morgan . Overholser . Parker . Parkinson Quist . Reed . Richardson . Romberger . Savage . Sears . Shaffer Shor ., Silvermon 160 154, O, Sisman 177. H. Insley (consultant) 155. B. A. Soldano 178. H. R. Jolley (consultant) 156. R. A. Strehlow 179. E. V. Jones (consultant) 157. F. H. Sweeton 180. T. N. McVay (consultant) 158. R. E. Thoma 181. G. Mamantov (consultant) 159-166. G. C. Warlick 182. J. L. Margrave {consultant) 167. C. F. Weaver 183. E. A. Mason (consultant) 168. L. B. Yeatts 184. R. F. Newton (consultant) 169. L. Brewer (consultant) 185. R. B. Perez (consultant) 170. J. W. Cobble (consultant) 186. J. E. Ricci (consultant) 171. F. Daniels (consultant) 187. Howard Reiss (consultant) 172. R. W. Dayton (consultant) 188. G. Scatchard {consultant) 173, P. H. Emmett (consultant) 189. D. A. Shirley {consultant) 174. H. S. Frank (consultant) 190. H. Steinfink (consultant) 175. N. Hackerman (consultant) 191. R. C. Vogel (consultant) 176. D. G. Hill (consultant) 192. T. F. Young (consultant) EXTERNAL DISTRIBUTION 193. D. F. Bunch, Health Physics Branch, AEC, Washington 194. Paul E. Field, Dept. of Chemistry, Virginia Polytechnic Institute 195. L. R. Zumwalt, General Atomic Division, General Dynamics Corp., San Diego, Calif. 196. Research and Development Division, AEC, ORO 197. Reactor Division, AEC, ORO 198. Assistant General Manager for Research and Development, AEC, Washington 199. Division of Research, AEC, Washington 200. Division of |sotopes Development, AEC, Washington 201. Assistant General Manager for Reactors, AEC, Washington 202. Division of Reactor Development and Technology, AEC, Washington 203. Space Nuclear Propulsion Office, AEC, Washington 204. J. A. Swartout, 270 Park Ave., New York 17, New York 205-533. Given distribution as shown in TID-4500 (47th Ed.) under Chemistry category (75 copies -- CFSTI)