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( ANL-91 Chemical Ter n 1i1gy Division Chemical Technology Division Chemical Technology Division Chemical Technology Division t..nem icd Chemical Chemical Chemical Chemical Chemical Chemical Chemical Chemical
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( ANL-91 Chemical Ter n 1i1gy Division Chemical Technology Division Chemical Technology Division Chemical Technology Division t..nem icd Chemical Chemical Chemical Chemical Chemical Chemical Chemical Chemical tecnnoiogy Div Vision Technology Division Technology Division Technology Division Technology Division Technology Division Technology Division Technology Division Technology Division Nuclear Technology Programs Semiannual Progress Report April-September 1989 ) Argonne National Laboratory. Argonne. llinois 6439 operated n; The University of Chicago for the United States Department of Energy under Contract W Eng-38 Chemical Technology Division Chemical Technology Division Chemical Technology Division Chemical Technology Division Argonne National Laboratory. with facilities in the states of Illinois and Idaho. is ow ned b :he United States gov-rnment. and operated by The University of Chicago under the provisions of a contract with the Department of Energy. DISCLAIMER This report was prepared as an account of w ork Aponsored h\ an agene, of the United States Go emment. Neither the Lnited States Government nor an% agency thereof. nor any of theiremploxees.makes an\ w arrant,. express or implied, or assumes any legal liability or responsibility% for the accurac%. completeness. or usefulness of an information. apparatus. product. or pro- Cese disclosed. or represents that its use would not infringe pri' ately owned nghts. Reference herein to an- specific commercial product. process. or,er ice b, trade name. trademark. manufacturer. or other' ise. does not necessaril\ constitute or impl% its endorsement, recommendation. or favoring b the United States Go ernment or any agenc% thereof. The iew, and opinions of authors expressed herein do not necessarily, state or reflect those of the United States Go.ernment or an'. agenc\, thereof. Reproduced from the best available copy. A' ailable to DOE and DOE contractors from the Office of Scientific and Technical Information P.O. Box 62 Oak Ridge. TN 3 31 Prices a% ailable from 615) 576-4u1. FTS 626-'4o1 A. ailable to the public from the %ational Technical Information Ser ice U.S. Department of Commerce 525 Port Roy al Road Springfield. VA 22161 Distribution Category: General, Miscellaneous, and Progress Reports (Nuclear) (UC-5) ANL--91/26 ANL-91/26 DE ARGONNE NATIONAL LABORATORY 97 South Cass Avenue Argonne, IL 6439 NUCLEAR TECHNOLOGY PROGRAMS SEMIANNUAL PROGRESS REPORT April-September 1989 Chemical Technology Division M. J. Steindler, Director J. E. Battles, Associate Director J. E. Harmon, Editor August 1991 Previous Reports in this Series October 1988-March 1989 ANL-9/16 April 1988-September 1988 ANL-9/15 October 1987-March 1988 ANL-89/29 MAASTER April 1987-September 1987 ANL D43TRt3D.7:QN C TH. 22 2~-'T $S UNL!MITEC TABLE OF CONTENTS A B ST R A C T SUMMARY APPLIED PHYSICAL CHEMISTRY... 1 A. Fission Product Vaporization from Core-Concrete Mixtures....iO 1. Introduction E xperim ental Results and Discussion Release during Severe Reactor Accidents C onclusions B. Thermophysical Properties of Metal Fuels Fuel-Cladding Compatibility U-Pu-Zr Phase Diagram Phase Relations in Fuel-Cladding Systems C. Desorption Characteristics of the LiA1O 2 -H 2 -H 2 (g) System Blank Experiments Measurements on LiAlO Future W ork D. Modeling of Tritium Behavior in Ceramic Breeder Materials Analysis of Results from Tritium Release Experiments Modeling of Tritium Inventory during Pulsed Operation Methods for Increasing Tritium Release E. Tritium Mass Transport in Ceramic Breeder Materials F. Interaction of Hydrogen with Li 2 Surfaces... 4 G. Blanket Design Studies Aqueous Salt Blanket Ceramic Solid Breeder Blanket H. Design of Breeding Blanket Interface I. Dosimetry and Damage Analysis Dosimetry Measurements in the Omega West Reactor Radiation Damage for High-Temperature Superconductors... 5 REFERENCES II. SEPARATION SCIENCE AND TECHNOLOGY Page 111 TABLE OF CONTENTS (contd) Page A. Generic TRUEX M odel Development M odel Enhancements SASSE Development B. Density Studies Density Correlation at Elevated Temperatures Verification of Density Predictions C. Estimation of Aqueous Electrolyte Activities at Elevated Temperatures D. Activity Measurements on A1(NO 3 ) E. Extraction Studies Oxalic Acid Plutonium (IV) Extraction Rare Earth Extraction Thorium, Neptunium, and Curium Extraction Iron Extraction Americium Extraction F. TRUEX-NPH Solvent Degradation Introduction Analysis of Experimental Results Previously Obtained for TRUEX-NPH Radiolysis and Hydrolysis Solvent W ashing G. Determination of Free Acid in Aluminum Nitrate Solutions H. Use of BAN and DIBAN Concentration of TRUEX Waste and Product Streams Distillation Precipitation Behavior J. Verification Studies Introduction Test Descrption Results and Discussion Contactor Heating Systems Contactor Cleaning/Decontamination K. Centrifugal Contactor Development Introduction Vibration Criteria Tests with 4-cm Contactors Consultation with W estinghouse Hanford iv TABLE OF CONTENTS (contd) L. Separation Processes to Treat Red W ater Analyses of Red Water Organic/Inorganic Separation REFERENCES Page III. HIGH-LEVEL WASTE/REPOSITORY INTERACTIONS A. Glass Studies for Yucca M ountain Project Unsaturated Test M ethod Parametric Experiments Static Leach Experiments Natural Analogs B. Spent Fuel Studies for Yucca M ountain Project Series 5 Spent Fuel Leach Tests Saturated Tests with Unirradiated U Unsaturated Tests with Unirradiated U C. Radiation Studies for Yucca M ountain Project Introduction Literature on Radiation Chemistry of Gaseous Ammonia Results and Discussion Conclusions D. Glass Studies for Defense Programs Product Consistency Test Defense W aste Glass Studies E. Detection and Speciation of Transuranic Elements LPAS Signal Response as a Function of Temperature Applicability to On-Line Detection of Uranyl in Nitric Acid Systems F. Experimental Validation of Performance Assessment for Repository Technology Program REFERENCES IV. PROGRAM ON PLUTONIUM RECOVERY FROM RESIDUES A. Introduction B. Process Development C. Calcium Reduction Experiments V TABLE OF CONTENTS (contd) 1. Slag Reduction Ash Heel Reduction D. Electrolytic Reduction of Calcium Oxide to Calcium Metal Page 1. Introduction Process Chem istry Experim ental Results and Discussion Conclusions E. Reference Electrod e Experim ents Zinc Electrode Electrode Response F. Calculations of Americium Distribution between Liquid Plutonium and M olten NaCI-KCI G. Design of Pyroprocessing Transfer Line Overall Design Heater Design and Testing Design Revision Tubing M anufacture A PPENDIX REFERENCES vi NUCLEAR TECHNOLOGY PROGRAMS SEMIANNUAL PROGRESS REPORT April-September 1989 ABSTRACT This document reports on the work done by the Nuclear Technology Programs of the Chemical Technology Division, Argonne National Laboratory, in the period April- September These programs involve R&D in three areas: applied physical chemistry, separation science and technology, and nuclear waste management. The work in applied physical chemistry includes investigations into the processes that control the release and transport of finsion products under accident-like conditions, the thermophysical properties of metal fuel -nd blanket materials of the Integral Fast Reactor, and the properties of selected matenals in environments simulating those of fusion energy systems. In the area of separation science and technology, the bulk of the effort is concerned with developing and implementing processes for the removal and concentration of actinides from waste streams contaminated by transuranic elements. Another effort is concerned with developing a process for separating the organic and inorganic constituents of the red-water waste stream generated in production of 2,4,6-trinitrotoluene. In the area of waste management, investigations are underway on the performance of materials in projected nuclear repository conditions to provide input to the licensing of the nation's high-level waste repositories. Applied Physical Chemistry SUMMARY Calculational and experimental efforts are underway to investigate the release of refractory fission products (Sr, Ba, and La) and uranium during the core-concrete interaction phase of a degraded-core accident. In the experimental effort, mixtures of urania, zirconia, and concrete are vapori d at 215 or 24 K into flowing H 2 or He-6% H 2 gas. Three different concretes having silica contents ranging from 7 to 69 wt % were used to reflect the known range of reactor-basemat compositions. The total mass of material that was vaporized was determined by weighing the condensates; the masses of individual elements were determined by chemical analyses of the condensates. The phases present in the heated mixtures were inferred from an electron probe microanalysis and X-ray diffraction analyses. Equilibrium calculations were also performed using the SOLGASMIX computer code and a thermodynamic data base containing 112 gaseous and 18 condensed species. The partial molar free energy of oxygen was calculated from the equilibrium oxygen pressure established in the high-temperature reaction zone between the gas and the sample. Using the experimental data, we estimated the release fractions to be expected in a severe nuclear-reactor accident for Sr, Ba, La, and U, and for total mass (aerosols). The release of these four elements was less than 1% for the basemat concrete of low silica content (7 wt %) and decreased to less than.1% for the basemat concrete of high silica content (69 wt %).These values are much lower than those reported in the literature. The total mass release was about.5% for all three concretes. Measurements are being performed to provide needed thermodynamic and transport property data for Integral Fast Reactor (IFR) fuels. As part of our investigation on fuel-cladding compatibility, we performed differential thermal analysis (DTA) experiments with mixtures of U-Pu-Zr fuel and steel 1 2 claddings (HT9, D9, and 316SS). The DTA curves on initial heating indicated solid-state transitions in the fuel at 6-7* C and an exothermal reaction forming more stable products (probably an Fe 2 Zr-like phase) at 12DC. On subsequent cooling, the DTA curve indicated primary precipitation at 122C and a freezing transition at 7*C. The DTA curves were used to estimate the onset-of-melting temperature with an accuracy of t 1-2 C. Scanning electron microscopy of the residues from these experiments indicated the following reaction sequence on heating: primary precipitation of a Fe 2 Zr-like phase, followed by secondary precipitation of Fe 2 U, followed by formation of Fe 2 U-FeU 6 eutectic. Other studies on metal fuel properties involved calculations of the Pu-U phase diagram, which will be used to refine the U-Pu-Zr phase diagram, and DTA experiments with U-Fe mixtures, which will be used to obtain a better understanding of the phase relations involved in fuel-cladding systems. A critical element in the development of a fusion reactor is the blanket for breeding tritium fuel. Several studies are underway with the objective of determining the feasibility of using lithium-containing ceramics as breeder material. In one such study, temperature programmed desorption (TPD) measurements are in progress to provide data that describe the kinetics of desorption of H 2 (g) and H 2 (g) from LiAlO 2 (s). Blank experiments (no LiAIO 2 present) were performed to gain information on the behavior of the empty stainless steel sample tube exposed to a flowing gas mixture, He- 1 ppm H 2-5 ppm H 2. Reactivity with the steel was demonstrated: H 2 was consumed and H 2 (g) was produced, and there was the suggestion of dissolution/reaction of H 2 (g) in the steel. Nevertheless, in the absence of H 2 (g), the sample tube can be stabilized so that useful measurements with H 2 (g) can be made. Fresh samples of LiAlO 2 (s) contain large amounts of adsorbed H 2. We obtained TPD spectra for LiAlO 2 samples that had been equilibrated with 2 ppm H 2 (g) at temperatures of 2, 3, 4, and 5* C. The spectra exhibited different shapes suggestive of differing and/or multiple processes taking place as H 2 is desorbed from the solid. In a continuing collaboration with Canadian researchers, we have been involved in analyzing tritium release data from CRITIC, an experimental study of tritium release from Li 2 O breeder material. Of particular interest is the tritium release at low temperatures. For temperature decreases to final temperatures below 45*C, the tritium release appeared to reach a plateau below the steady-state release. Initially, it was believed that this may have been a result of some instrumental problem, possibly due to a change in the sensitivity of the proportional counter used for tritium measurements at different temperatures or gas compositions. However, the data from the proportional counters and data from the scintillation counting of tritium collected in glycol traps indicated that release was below steady-state release for substantial periods of time. After considering other possible instrumental or measurement errors and ruling them out, we arrived at a hypothesis that the plateau region may be caused by a second phase of LiOT precipitating in the Li 2 O. In related work, we calculated the tritium inventory during pulsed operation for a lithium oxide blanket and investigated methods for increasing tritium release in breeder materials. With regard to the latter, our calculations indicated that doping will increase the lithium vacancy concentration in LiAlO 2, which would increase its tritium diffusivity. Within the solid breeder material, tritium may be found as LiOT, which may transport lithium (and tritium) to cooler parts of the blanket. This process may cause loss of lithium from the blanket, blocking of flow paths, and increase of the tritium inventory. Experiments were, therefore, undertaken to investigate the transport of LiOH from a lithium oxide solid breeder with a helium purge stream containing water vapor. The gas velocity was varied from 5 to 26 cm/s at 85 C. The fractional saturation product for H 2 O(g) diffusion from the flowing helium to the lithium oxide and LiOH(g) diffusion into the flowing helium was calculated from the experimental results. Our results were in general agreement with those reported by others. This work established conditions for calculating LiOH undersaturation in helium as a function of the blanket purge channel dimensions and velocity of the purge gas. 3 Earlier experiments indicated that substantial increase of tritium release occurs when a small amount of H 2 (about.1%) is added to the helium purge stream. We are using a quantum cluster approach with the extended Huckel method to investigate the interaction between H 2 (g) and the Li 2 O(1 1) surface. Three types of sites were investigated: midway between the nearest neighbor -, -Li, and Li-Li. We found that all three are sites for H 2 nondissociative adsorption. The binding energetics are around.2-.3 ev, and the equilibrium height is around 2 A. The International Thermonuclear Experimental Reactor (ITER) is an international project whose purpose is to produce a conceptual design of a tokamak reactor which can be used to test components for a prototype power reactor. Designs of each of the reactor systems are being generated by the ITER partners in the U.S., U.S.S.R., Japan, and the European community. In this report period, we prepared a design of the tritium processing system for the U.S. ITER aqueous salt blanket design and provided it to the staff at the Tritium Systems Test Assembly, the designated coordinator of the U.S. tritium systems design for ITER. This information was incorporated into a total tritium systems design for ITER, which includes the plasma exhaust processing system plus the blanket streams. Also, the effects of radiolysis and electrolytic decomposition were examined for the borated water shield and water coolant for a ceramic solid breeder. The Breeding Blanket Interface (BBI) is that system which performs the necessary processing and recycle of the tritium recovered from a fusion reactor blanket. Although considerable attention has been given to studies of tritium breeding and recovery and also many aspects of tritium processing, very little work has been done on the BBI. In this report period, preconceptual designs of the BBI were developed for two types of blankets: aqueous salt solution and solid breeder blanket. A summary of the major design features of the two BBI systems is given. In neutron dosimetry and damage analysis, fusion materials are being irradiated in a variety of facilities, including fission reactors, 14 MeV neutron sources, and higher energy accelerator-based neutron sources. We are determining the neutron energy spectrum, flux levels, and damage parameters for the materials irradiated in these facilities, along with exposure parameters for each irradiation. In this report period, neutron dosimetry and radiation damage calculations were performed for three short irradiations by Battelle Pacific National Laboratory in the Omega West Reactor at Los Alamos National Laboratory. The dosimetry capsules contained small wires of Ni, Fe, Ti,.1% Co-Al, and 8.2% Mn-Cu. The samples were irradiated at 2 C and a maximum neutron fluence of 8.3 x 119 n/cm 2. Data are reported for activation rates, neutron flux and fluence, and radiation damage parameters. Using the SPECOMP computer code, we also calculated the production of atomic displacement damage in the superconducting material YBa 2 Cu 3 O 7. The results have been added to the SPECTER computer code for routine calculation of damage in any specified neutron spectrum. Separation Science and Technology The Division's work in separation science and technology is mainly concerned with removing and concentrating actinides from waste streams contaminated with transuranic (TRU) elements by use of the TRUEX solvent extraction process. The extractant found most satisfactory for the TRUEX process is octyl (phenyl)-n,n-diisobutylcarbamoylmethylphosphine oxide, which is abbreviated CMPO. This extractant is combined with tributyl phosphate (TBP) and a diluent to formulate the TRUEX process solvent. The diluent is typically a normal paraffinic hydrocarbon (NPH) or a nonflammable chlorocarbon such as tetrachoroethylene (TCE). Another project was initiated to develop a process for converting the hazardous red-water waste stream from TNT manufacturers to forms that are readily disposable or acceptable for recycling. 4 The major effort involves development of a generic data base and modeling capability for the TRUEX solvent extraction process. The Generic TRUEX Model (GTM) will be directly useful for sitespecific flowsheet development directed to (1) establishing a TRUEX process for specific waste streams, (2) assessing the economic and facility requirements for installing the process, and (3) improving, monitoring, and controlling on-line TRUEX processes. Versions of the model are now available for use with Macintosh and IBM-compatible personal computers. A more powerful and faster second-generation GTM will be available in 199. Improvements will include the ability to calculate flowsheets and predict space and cost requirements for any type of solvent extraction equipment, including pulsed columns, and to estimate solvent degradation during processing caused by radiolysis and hydrolysis of the extractant (CMPO). One of the sections in the GTM, the Spreadsheet Algorithm for Stagewise Solvent Extraction (SASSE), was modified so that (1) it can now handle large amounts of other-phase carryover, (2) residence time is explicit rather than implicit, and (3) the volume of each phase in each stage in a TRUEX flowsheet can be specified. This modified worksheet is being used to evaluate a new model for solvent extraction in columns. In the future, it will be used to model time-dependent chemica
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