Leslie Kevin Felker

Chemistry of metal chloride complexes in aprotic systems

Abstract A study of metal chloride solubility in aprotic solvents has been initiated. These solvent systems have very low hydrogen ion activities and thus allow chloride ion activities which are much higher than those attainable in water. The high chloride ion activities can be generated by the dissolution of soluble salts, such as calcium chloride or sodium chloride, in the aprotic media. Metals that normally form aqueous-insoluble chlorides or exist as cations in aqueous solutions (e.g., Ag, Pb, Cd, or Au) may be readily dissolved in aprotic solvent systems as anionic chloride complexes or as solvation complexes. To understand such systems, we constructed ternary phase diagrams for dimethylsulfoxide-water-calcium chloride (DMSO-H2O-CaCl2) and DMSO-H2O-NaCl systems. These diagrams were used to establish the solution regions available for the solubilization of Pb, Ag, Au, Cd, Cu, Zn, and Al. Measurements of Pb and Ag solubilities in the DMSO-H2O-CaCl2 system gave concentrations as high as ∼1.5 and ∼4).9 M...

Separation of Americium, Curium, and Plutonium from Irradiated Targets

Abstract The Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL) processes highly irradiated targets for the Mark 42 program to separate Am, Cm, and Pu. The target feed material for each assembly was 3.3 kg of plutonium (78% 239Pu) that was irradiated at the Savannah River Site to yield approximately 100 g each of 243Am and 244Cm, and 100-g quantities of 242Pu for special DOE projects. The REDC has plans to process ten of these target assemblies over the next few years. The first assembly has been dissolved, and approximately 1/4 of this material has been used to test the processing flowsheet. Various aqueous processes developed at the REDC over the past years were utilized to dissolve the target segments, separate the bulk of the impurities from the transuranics, separate the plutonium from the transplutonium actinides, and separate the rare earth fission products from the Am-Cm. The separation of the Am-Cm products to the desired purity levels presented new proces...

Reestablishing the Supply of Plutonium-238

The US Department of Energy has presented a plan to use existing reactors at Oak Ridge National Laboratory (ORNL) and Idaho National Laboratory (INL) and processing facilities at ORNL, modified as needed, to produce Pu. The basic capabilities that need to be put into place to produce new Pu are (1) neptunium storage and transport, (2) target fabrication, (3) target irradiation, and (4) chemical processing of irradiated targets to recover Pu. Neptunium currently in storage at INL will be shipped to ORNL during CY 2015. The target design has progressed to a prototypic target design that is expected to be used for production. Initial chemical processing experiments have shown successful recovery of neptunium and plutonium, but overall product purity has not been as high as desired.

Dissolution and Separation of Aluminum and Aluminosilicates

Selection of an aluminum alloy for target cladding affects post-irradiation target dissolution and separations. Recent tests with aluminum alloy 6061 yielded greater than expected precipitation in the dissolver, forming up to 10 wt.% solids of aluminum hydroxides and aluminosilicates. Aluminosilicate dissolution presents challenges in a number of different areas, including metals extraction from minerals, flyash treatment, and separations from aluminum alloys. We present experimental work that attempts to maximize dissolution of aluminum metal in caustic, along with silicon, magnesium, and copper impurities, through control of temperature, the rate of reagent addition, and incubation time. Aluminum phase transformations have been identified as a function of time and temperature, using X-ray diffraction. Solutions have been analyzed using wet chemical methods and X-ray fluorescence. These data have been compared with published calculations of aluminum phase diagrams. Approaches are given to enhance the dissolution of aluminum and aluminosilicate phases in caustic solution.

Fermium Purification Using Teva™ Resin Extraction Chromatography

Abstract The Radiochemical Engineering Development Center at Oak Ridge National Laboratory processes irradiated targets to recover the transplutonium actinides for research and industrial users. In a typical processing campaign, dekagram quantities of curium are recovered for recycle into targets for subsequent irradiation and processing; decigram quantities of californium are recovered for fabrication into neutron sources; and milligram quantities of einsteinium and berkelium as well as picogram quantities of fermium are recovered for distribution to the research community. The transcurium actinides are separated in a series of chromatographic elutions using a cation-exchange resin and ammonium α-hydroxyisobutyrate as the eluant. The fermium fraction from these final purification runs still contains significant amounts of rare earth fission products, such as yttrium, dysprosium, and holmium. In the most recent campaign, a process using a TEVA™ resin extraction chromatography column was developed and test...

Aluminum Target Dissolution in Support of the Pu-238 Program

Selection of an aluminum alloy for target cladding affects post-irradiation target dissolution and separations. Recent tests with aluminum alloy 6061 yielded greater than expected precipitation in the caustic dissolution step, forming up to 10 wt.% solids of aluminum hydroxides and aluminosilicates. We present a study to maximize dissolution of aluminum metal alloy, along with silicon, magnesium, and copper impurities, through control of temperature, the rate of reagent addition, and incubation time. Aluminum phase transformations have been identified as a function of time and temperature, using X-ray diffraction. Solutions have been analyzed using wet chemical methods and X-ray fluorescence. These data have been compared with published calculations of aluminum phase diagrams. Temperature logging during the transients has been investigated as a means to generate kinetic and mass transport data on the dissolution process. Approaches are given to enhance the dissolution of aluminum and aluminosilicate phases in caustic solution.

THE ABSORPTION SPECTRUM OF UOCL2 IN MOLTEN CHLORIDE SALTS

Abstract The absorption spectrum of UOCl2 in molten KCl-MgCl2 salts has been measured and compared with that of the related UCl4 spectrum at temperatures up to 932°C and melt compositions of 60–40, 34–66, and 0–100 mol %, respectively, KCl-MgCl2. The species UOCl2 is an important intermediate in the conversion of UO2 to UCl4 and its spectrum provides a means of monitoring the reaction, and other similar reactions, in situ. The solubility of UOCl2 has been determined from absorption spectra and has been found to be 10 to 25 times higher than reported earlier with mole fractions as high as 0.0019 in pure MgCl2 at 932°C.

The absorption spectrum of UOCl{sub 2} in molten chloride salts

Abstract The absorption spectrum of UOCl2 in molten KCl-MgCl2 salts has been measured and compared with that of the related UCl4 spectrum at temperatures up to 932°C and melt compositions of 60–40, 34–66, and 0–100 mol %, respectively, KCl-MgCl2. The species UOCl2 is an important intermediate in the conversion of UO2 to UCl4 and its spectrum provides a means of monitoring the reaction, and other similar reactions, in situ. The solubility of UOCl2 has been determined from absorption spectra and has been found to be 10 to 25 times higher than reported earlier with mole fractions as high as 0.0019 in pure MgCl2 at 932°C.

Fluorine gettering by activated charcoal in a radiation environment

Activated charcoal has been shown to be an effective gettering agent for the fluorine gas that is liberated in a radiation environment. Even though activated charcoal is a commonly used getter, little is known about the radiation stability of the fluorine-charcoal product. This work has shown that not only is the product stable in high gamma radiation fields, but also that radiation enhances the capacity of the charcoal for the fluorine. The most useful application of this work is with the Molten Salt Reactor Experiment (MSRE) fuel salt because the radioactive components (fission products and actinides) cause radiolytic damage to the solid LiF-BeF/sub 2/-ZrF/sub 4/-UF/sub 4/ (64.5, 30.3, 5.0, 0.13 mol %, respectively) resulting in the liberation of fluorine gas. This work has also demonstrated that the maximum damage to the fuel salt by approx.3 /times/ 10/sup 7/ R/h gamma radiation is approximately 2%, at which point the rate of recombination of fluorine with active metal sites within the salt lattice equals the rate of fluorine generation. The enhanced reactivity of the activated charcoal and radiation stability of the product ensures that the gettered fluorine will stay sequestered in the charcoal.