A.S. Icenhour
Office of Naval Research
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Archive | 2005
A.S. Icenhour
The movement of high-specific-activity radioactive particles (i.e., alpha recoil) has been observed and studied since the early 1900s. These studies have been motivated by concerns about containment of radioactivity and the protection of human health. Additionally, studies have investigated the potential advantage of alpha recoil to effect separations of various isotopes. This report provides a review of the observations and results of a number of the studies.
Radiochimica Acta | 2002
A.S. Icenhour; L.M. Toth; G. D. Del Cul; Laurence F. Miller
Summary The safe handling and storage of radioactive materials require an understanding of the effects of radiolysis on those materials. Radiolysis may result in the production of gases (e.g., corrosives) or pressures that are deleterious to storage containers. A study has been performed to address these concerns as they relate to the radiolysis of residual fluoride compounds in uranium oxides. Samples of UO2F2·xH2O and U3O8 (with ∼1.4 wt.% fluorine content) were irradiated in a 60Co source and in spent nuclear fuel (SNF) elements from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. Container pressures were monitored throughout the irradiations, and gas and solid samples were analyzed after the irradiations. The irradiation of UO2F2·xH2O produced O2 – with G(O2)-values ranging from 0.007 to 0.03 molecules of O2 produced per 100 eV. Neither F2 nor HF was produced by the irradiations. Chemical analyses of solid samples showed that some of the uranium was reduced from U(VI) to U(IV). A saturation damage limit for the UO2F2·xH2O was demonstrated by using the HFIR SNF elements, and the limit was found to be 7–9% at ∼108 rad/h). It is shown that the covalently bonded oxygen is more susceptible to radiation damage than is the ionically bonded fluorine. Irradiation of U3O8 (with ∼1.4 wt.% fluorine content) resulted in neither gas production nor a pressure increase. These experiments led to the conclusion that during long-term storage U3O8 is safe from overpressurization and the production of corrosives caused by gamma radiolysis of residual fluorides.
Nuclear Technology | 2004
A.S. Icenhour; L.M. Toth; Robert M. Wham; R. R. Brunson
Abstract Alpha radiolysis experiments have been performed on NpO2 that contains sorbed moisture. A high dose rate to the sample was achieved by spiking it with ~7000 ppm 244Cm during preparation. Pressure monitoring of sample containers showed that a low, steady-state pressure plateau is reached. This plateau indicates a situation in which the forward reaction (i.e., radiolysis of water) is equal to the back reaction (i.e., the reformation of H2O). In this technical note, a simple kinetic model that can be used for predicting steady-state pressures under practical conditions is described.
Other Information: PBD: 1 Apr 2000 | 2000
G. D. Del Cul; A.S. Icenhour; D.W. Simmons
The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.
Other Information: PBD: 27 Feb 2002 | 2002
A.S. Icenhour
During the development of a standard for the safe, long-term storage of {sup 233}U-containing materials, several areas were identified that needed additional experimental studies. These studies were related to the perceived potential for the radiolytic generation of large pressures or explosive concentrations of gases in storage containers. This report documents the results of studies on the sorption of water by various uranium oxides and on the gamma radiolysis of uranium oxides containing various amounts of sorbed moisture. In all of the experiments, {sup 238}U was used as a surrogate for the {sup 233}U. For the water sorption experiments, uranium oxide samples were prepared and exposed to known levels of humidity to establish the water uptake rate. Subsequently, the amount of water removed was studied by heating samples in a oven at fixed temperatures and by thermogravimetric analysis (TGA)/differential thermal analysis (DTA). It was demonstrated that heating at 650 C adequately removes all moisture from the samples. Uranium-238 oxides were irradiated in a {sup 60}Co source and in the high-gamma-radiation fields provided by spent nuclear fuel elements of the High Flux Isotope Reactor. For hydrated samples of UO{sub 3}, H{sub 2} was the primary gas produced; but the total gas pressure increase reached steady value of about 10 psi. This production appears to be a function of the dose and the amount of water present. Oxygen in the hydrated UO{sub 3} sample atmosphere was typically depleted, and no significant pressure rise was observed. Heat treatment of the UO{sub 3} {center_dot} xH{sub 2}O at 650 C would result in conversion to U{sub 3}O{sub 8} and eliminate the H{sub 2} production. For all of the U{sub 3}O{sub 8} samples loaded in air and irradiated with gamma radiation, a pressure decrease was seen and little, if any, H{sub 2} was produced--even for samples with up to 9 wt % moisture content. Hence, these results demonstrated that the efforts to remove trace moisture from U{sub 3}O{sub 8} are not necessary to avoid pressurization of stored uranium oxides caused by gamma-induced radiolysis. In fact, this system can tolerate several percent of sorbed moisture--most of which can be easily removed by heating to only 150 C. To complete the picture of the radiolytic response of uranium oxides that have sorbed moisture, alpha radiolysis experiments have been initiated.
Other Information: PBD: 24 Jan 2002 | 2002
A.S. Icenhour
The development of a standard for the safe, long-term storage of {sup 233}U-containing materials resulted in the identification of several needed experimental studies. These studies were largely related to the potential for the generation of unacceptable pressures or the formation of deleterious products during storage of uranium oxides. The primary concern was that these conditions could occur as a result of the radiolysis of residual impurities--specifically fluorides and water-by the high radiation fields associated with {sup 233}U/{sup 232}U-containing materials. This report documents the results from a gamma radiolysis experiment in which UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O was loaded in helium. This experiment was performed using spent nuclear fuel elements from the High Flux Isotope Reactor as the gamma source and was a follow-on to experiments conducted previously. It was found that upon gamma irradiation, the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O released 0{sub 2} with an initial G(O{sub 2}) = 0.01 molecule O{sub 2}/100 eV and that some of the uranium was reduced from U(VI) to U(IV). The high total dose achieved in the SNF elements was sufficient to reach a damage limit for the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O. This damage limit, measured in terms of the amount of the U(IV) produced, was found to be about 9 wt%.
Archive | 2005
A.S. Icenhour
The U.S. Department of Energy Advanced Fuel Cycle Initiative (AFCI) is sponsoring a project at Oak Ridge National Laboratory with the objective of conducting the research and development necessary to evaluate the use of sphere-pac transmutation fuel. Sphere-pac fuels were studied extensively in the 1960s and 1970s. More recently, this fuel form is being studied internationally as a potential plutonium-burning fuel. For transmutation fuel, sphere-pac fuels have potential advantages over traditional pellet-type fuels. This report provides a review of development efforts related to the preparation of sphere-pac fuels and their irradiation tests. Based on the results of these tests, comparisons with pellet-type fuels are summarized, the advantages and disadvantages of using sphere-pac fuels are highlighted, and sphere-pac options for the AFCI are recommended. The Oak Ridge National Laboratory development activities are also outlined.
Radiation Effects and Defects in Solids | 2004
A.S. Icenhour; L.M. Toth
Alpha radiolysis studies have been performed on uranium oxides and oxyfluorides (UO3, U3O8, and UO2F2) to evaluate the long-term storage characteristics of 233U. These uranium compounds (using 238U as the surrogate for 233U) were subjected to relatively high alpha radiation doses (235–634 MGy) by doping with 244Cm. The typical irradiation time for these samples was about 1.5 years, which would be equivalent to more than 50 years irradiation by a 233U sample. Both dry and wet (up to 10 wt% water) samples were examined in an effort to identify the gas pressure and composition changes that occurred as a result of radiolysis. This study shows that several competing reactions occur during radiolysis with the net effect of generating only very low pressures of hydrogen, nitrogen, and carbon dioxide from water, nitrate, and carbon impurities, respectively. In the absence of nitrate impurities, no pressures greater than 1000 Torr are generated. Usually, however, the oxygen in the air atmosphere over the oxides is consumed with the corresponding oxidation of the uranium oxide. In the presence of up to 10 wt% water, the oxides first show a small pressure rise followed by a net decrease due to the oxygen consumption and the attainment of a steady-state pressure at which the rate of generation of gaseous components is balanced by their recombination and/or consumption in the oxide phase.
Archive | 2004
A.S. Icenhour
Plans are to convert the {sup 237}Np that is currently stored as a nitrate solution at the Savannah River Site to NpO{sub 2} and then ship it to the Y-12 National Security Complex in Oak Ridge for interim storage. This material will serve as feedstock for the {sup 238}Pu production program, and some will be periodically shipped to the Oak Ridge National Laboratory (ORNL) for fabrication into targets. The safe storage of this material requires an understanding of the radiolysis of moisture that is sorbed on the oxides, which, in turn, provides a basis for storage criteria (namely, moisture content). A two-component experimental program has been undertaken at ORNL to evaluate the radiolytic effects on NpO{sub 2}: (1) moisture uptake experiments and (2) radiolysis experiments using both gamma and alpha radiation. These experiments have produced two key results. First, the water uptake experiments demonstrated that the 0.5 wt % moisture limit that has been typically established for similar materials (e.g., uranium and plutonium oxides) cannot be obtained in a practical environment. In fact, the uptake in a typical environment can be expected to be at least an order of magnitude lower than the limit. The second key result is the establishment of steady-state pressure plateaus as a result of the radiolysis of sorbed moisture. These plateaus are the result of back reactions that limit the overall pressure increase and H{sub 2} production. These results clearly demonstrate that 0.5 wt % H{sub 2}O on NpO{sub 2} is safe for long-term storage--if such a moisture content could even be practically reached.
Nuclear Technology | 2001
Guillermo D. Del Cul; A.S. Icenhour; Darrell W. Simmons
Abstract The Molten Salt Reactor Experiment (MSRE) site at Oak Ridge National Laboratory is being cleaned up and remediated. The removal of ~37 kg of fissile 233U is the main activity. Of that inventory, ~23 kg has already been removed as UF6 from the piping system and chemisorbed in 25 NaF traps. This material is in temporary storage while it awaits conversion to a stable oxide. The planned recovery of ~11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a uranium oxide (U3O8), which is suitable for long-term storage. The conversion of the MSRE material into an oxide presents unique problems, such as criticality concerns, a large radiation field caused by the daughters of 232U (an impurity isotope in the 233U), and the possible spread of the high-radiation field from the release of 220Rn gas. To overcome these problems, a novel process was conceived and developed. This process was specially tailored for providing remote operations inside a hot cell while maintaining full containment at all times to avoid the spread of contamination. This process satisfies criticality concerns, maximizes the recovery of uranium, minimizes any radiation exposure to operators, and keeps waste disposal to a minimum.