34 research outputs found
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Progress in chemical treatment of LEU targets by the modified Cintichem process
Presented here are recent experimental results on tests of a modified Cintichem process for producing {sup 99}Mo from low enriched uranium (LEU). Studies were focused in three areas: (1) testing the effects on {sup 99}Mo recovery and purity of dissolving LEU foil in nitric acid alone, rather than in the sulfuric/nitric acid mixture currently used, (2) measuring decontamination factors for radionuclide impurities in each purification step, and (3) testing the effects on processing of adding barrier materials to the LEU metal-foil target. The experimental results show that switching from dissolving the target in the sulfuric/nitric mixture to using nitric acid alone should cause no significant difference in {sup 99}Mo product yield or purity. Further, the results show that overall decontamination factors for gamma emitters in the LEU-target processing are high enough to meet the purity requirements for the {sup 99}Mo product. The results also show that the selected barrier materials, Cu, Fe, and Ni, do not interfere with {sup 99}Mo recovery and can be removed during chemical processing of the LEU target
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Separation science and technology. Semiannual progress report, October 1993--March 1994
This document reports on the work done by the Separations Science and Technology Programs of the Chemical Technology Division, Argonne National Laboratory (ANL), in the period October 1993-March 1994. This effort is mainly concerned with developing the TRUEX process for removing and concentrating actinides from acidic waste streams contaminated with transuranic (TRU) elements. The objectives of TRUEX processing are to recover valuable TRU elements and to lower disposal costs for the nonTRU waste product of the process. Other projects are underway with the objective of developing (1) evaporation technology for concentrating radioactive waste and product streams such as those generated by the TRUEX process, (2) treatment schemes for liquid wastes stored are being generated at ANL, (3) a process based on sorbing modified TRUEX solvent on magnetic beads to be used for separation of contaminants from radioactive and hazardous waste streams, and (4) a process that uses low-enriched uranium targets for production of {sup 99}Mo for nuclear medicine uses
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Treatment of plutonium-bearing solutions: A brief survey of the DOE complex
With the abrupt shutdown of some DOE facilities, a significant volume of in-process material was left in place and still requires treatment for interim storage. Because the systems containing these process streams have deteriorated since shutdown, a portable system for treating the solutions may be useful. A brief survey was made of the DOE complex on the need for a portable treatment system to treat plutonium-bearing solutions. A survey was completed to determine (1) the compositions and volumes of solutions and heels present, (2) the methods that have been used to treat these solutions and heels in the past, and (3) the potential problems that exist in removing and treating these solutions. Based on the surveys and on the Defense Nuclear Facilities Safety Board Recommendation 94-1, design criteria for a portable treatment system were generated
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Processing of LEU targets for {sup 99}Mo production -- Dissolution of metal foil targets by alkaline hydrogen peroxide
In FY 1995, the authors started studies on a new process for dissolution of low-enriched uranium (LEU) targets for {sup 99}Mo production. In this process, an LEU metal foil target is dissolved in a mixture of sodium hydroxide and hydrogen peroxide, then {sup 99}Mo is recovered from the dissolved solution. They focused on the dissolution kinetics to develop a mechanistic model for predicting the products and the rate of uranium dissolution under process conditions. They thoroughly studied the effects of hydrogen peroxide concentration, sodium hydroxide concentration, and temperature on the rate of uranium dissolution. It was found that uranium dissolution can be classified into a low-base ( 0.2M) process. In the low-base process, both the equilibrium hydrogen peroxide and hydroxide concentrations affect the rate of uranium dissolution; in the high base process, uranium dissolution is a 0.25th order reaction with respect to the equilibrium hydrogen peroxide. The dissolution activation energy was experimentally determined to be 48.8 kJ/mol. Generally, the rate of uranium dissolution increases to a maximum as the hydroxide concentration is increased from 0.01 to about 1.5M, then it decreases as the hydroxide concentration is further increased. The alkalinity of the dissolution solution is an important factor that affects not only the dissolution rate, but also the amount of radioactive waste
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Optimization of magnetite carrier precipitation process for transuranic waste reduction
Transuranic (TRU) waste that is being generated at Argonne National Laboratory has a TRU activity ranging from 10{sup 2} to 10{sup 7} nCi/g with a wide variety of chemical compositions. Currently, the waste is stored in highly acidic solutions that must be neutralized for intermediate storage. A magnetite carrier precipitation process has been adapted to concentrate TRU isotopes in a noncorrosive solid phase. In this paper, the authors report the results of a series of laboratory tests done to optimize the process. The parameters they optimized included (1) magnetite concentration used to precipitate the TRUs from solution, (2) formation of magnetite (in situ or ex situ), (3) processing pH, and (4) temperature and mixing time of the carrier precipitation. They also studied the effects of anions, cations, and complexing agents in the waste solutions on the carrier precipitation and the effect of magnetite solids loading on the filtration equipment. An overview is given of the planned full-scale process, which will be operated in a glovebox
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Transformation of UREX effluents to solid oxides by concentration, denitration, and calcination.
A literature survey was conducted on the potential means for converting two aqueous streams from the UREX process to solid oxides. The UREX process is similar to the PUREX solvent-extractions process and is designed to separate U, Tc, and I from the transuranics and other fission products contained in spent fuel. It was determined that converting these waste streams to oxides is technically feasible by a combination of evaporation, denitration, and calcination
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Application of single ion activity coefficients to determine solvent extraction mechanism for components of high level nuclear waste
The TRUEX solvent extraction process is being developed to remove and concentrate transuranic (TRU) elements from high-level and TRU radioactive wastes currently stored at US Department of Energy sites. Phosphoric acid is one of the chemical species of concern at the Hanford site where bismuth phosphate was used to recover plutonium. The mechanism of phosphoric acid extraction with TRUEX-NPH solvent at 25{degrees}C was determined by phosphoric acid distribution ratios, which were measured by using phosphoric acid radiotracer and a variety of aqueous phases containing different concentrations of nitric acid and nitrate ions. A model was developed for predicting phosphoric acid distribution ratios as a function of the thermodynamic activities of nitrate ion and hydrogen ion. The Generic TRUEX Model (GTM) was used to calculate these activities based on the Bromley method. The derived model supports CMPO and TBP extraction of a phosphoric acid-nitric acid complex and a CMPO-phosphoric acid complex in TRUEX-NPH solvent
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Validation of the generic TRUEX model using data from TRUEX demonstrations with actual high-level waste
The main objective of the Generic TRUEX Model (GTM) is to calculate TRUEX solvent extraction flowsheets based on input of a specific feed and a specific set of process goals and constraints. The output will be (1) the compositions of all effluent streams, (2) the compositions of both phases (organic and aqueous) in each stage of the contacting equipment at steady state, and (3) estimates of the space and cost requirements for installing the flowsheet in a plant situation. Other options are available to calculate aqueous-phase speciation and thermodynamic activities, distribution ratios of extractable species, and solvent radiolytic and hydrolytic degradation. Calculation of these options is based on initial aqueous- and organic-phase compositions and other important variables supplied by the user. Three demonstrations of the TRUEX process have been run by Power Reactor and Nuclear Fuel Development Corp. (PNC) researchers at the Tokai Works using actual PUREX raffinates. A 19-stage mixer settler was used for the extraction and scrub sections, and a 16-to-19-stage unit for stripping. Stagewise data were collected on the behavior of nitric acid and several fission-product and actinide radioisotopes during these runs; Run 2 was the best documented and the one with which most comparisons were made. These data are important tools for validating predictions made by the GTM and understanding the intricacies of the TRUEX process. In this paper, results of the GTM calculations will be compared to the actual data published by PNC researchers. Differences between model predictions and experimental data were analyzed in terms of the process chemistry and demonstration conditions
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Separation Science and Technology semiannual progress report, October 1992--March 1993
This document reports on the work done by the Separations Science and Technology Section of the Chemical Technology Division, Argonne National Laboratory, in the period October 1992--March 1993. This effort is mainly concerned with developing the TRUEX process for removing and concentrating actinides from acidic waste streams contaminated with transuranic (TRU) elements. The objectives of TRUEX processing are to recover valuable TRU elements and to lower disposal costs for the nonTRU waste product of the process. Two other projects are underway with the objective of developing (1) evaporation technology for concentrating radioactive waste and product streams such as those generated by the TRUEX process and (2) treatment schemes for liquid wastes stored or being generated at Argonne
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Recovery of plutonium and americium from chloride salt wastes by solvent extraction
Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetrachloroethylene (TCE) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HCl, the Pu(III) is oxidized to the IV state with NaClO/sub 2/ and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed. 16 refs., 6 figs