[en] During the development of a standard for the safe, long-term storage of ²³³U-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, ²³⁸U was used as a surrogate for the ²³³U. 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 ⁶⁰Co source and in the high-gamma-radiation fields provided by spent nuclear fuel elements of the High Flux Isotope Reactor. For hydrated samples of UO₃, H₂ 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 UO₃ sample atmosphere was typically depleted, and no significant pressure rise was observed. Heat treatment of the UO₃ · xH₂O at 650 C would result in conversion to U₃O₈ and eliminate the H₂ production. For all of the U₃O₈ samples loaded in air and irradiated with gamma radiation, a pressure decrease was seen and little, if any, H₂ was produced--even for samples with up to 9 wt % moisture content. Hence, these results demonstrated that the efforts to remove trace moisture from U₃O₈ 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

[en] Plans are to convert the ²³⁷Np that is currently stored as a nitrate solution at the Savannah River Site to NpO₂ 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 ²³⁸Pu 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 NpO₂: (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 H₂ production. These results clearly demonstrate that 0.5 wt % H₂O on NpO₂ is safe for long-term storage--if such a moisture content could even be practically reached

[en] During FY 2001, two arrays, each containing seven neptunium-loaded targets, were irradiated at the Advanced Test Reactor in Idaho to examine the influence of multi-target self-shielding on ²³⁶Pu content and to evaluate fission product release data. One array consisted of seven targets that contained 10 vol% NpO₂ pellets, while the other array consisted of seven targets that contained 20 vol % NpO₂ pellets. The arrays were located in the same irradiation facility but were axially separated to minimize the influence of one array on the other. Each target also contained a dosimeter package, which consisted of a small NpO₂ wire that was inside a vanadium container. After completion of irradiation and shipment back to the Oak Ridge National Laboratory, nine of the targets (four from the 10 vol% array and five from the 20 vol% array) were punctured for pressure measurement and measurement of ⁸⁵Kr. These nine targets and the associated dosimeters were then chemically processed to measure the residual neptunium, total plutonium production, ²³⁸Pu production, and ²³⁶Pu concentration at discharge. The amount and isotopic composition of fission products were also measured. This report provides the results of the processing and analysis of the nine targets

[en] Site-specific radiological performance assessments are required for the disposal of low-level radioactive waste (LLW) at both commercial and US Department of Energy facilities. This work explores source term modeling of LLW disposal facilities by using two state-of-the-art computer codes, SOURCEI and SOURCE2. An overview of the performance assessment methodology is presented, and the basic processes modeled in the SOURCE1 and SOURCE2 codes are described. Comparisons are made between the two advective models for a variety of radionuclides, transport parameters, and waste-disposal technologies. These comparisons show that, in general, the zero-order model predicts undecayed cumulative fractions leached that are slightly greater than or equal to those of the first-order model. For long-lived radionuclides, results from the two models eventually reach the same value. By contrast, for short-lived radionuclides, the zero-order model predicts a slightly higher undecayed cumulative fraction leached than does the first-order model. A new methodology, based on sensitivity and uncertainty analyses, is developed for predicting intruder scenarios. This method is demonstrated for ¹³⁷Cs in a tumulus-type disposal facility. The sensitivity and uncertainty analyses incorporate input-parameter uncertainty into the evaluation of a potential time of intrusion and the remaining radionuclide inventory. Finally, conclusions from this study are presented, and recommendations for continuing work are made

[en] 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

[en] The development of a standard for the safe, long-term storage of ²³³U-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 ²³³U/²³²U-containing materials. This report documents the results from a gamma radiolysis experiment in which UO₂F₂ · 0.4H₂O 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 UO₂F₂ · 0.4H₂O released 0₂ with an initial G(O₂) = 0.01 molecule O₂/100 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 UO₂F₂ · 0.4H₂O. This damage limit, measured in terms of the amount of the U(IV) produced, was found to be about 9 wt%

[en] 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

[en] The radiolysis of sorbed water and other impurities contained in actinide oxides has been the focus of a number of studies related to the establishment of criteria for the safe storage and transport of these materials. Gamma radiolysis studies have previously been performed on uranium oxides and oxyfluorides (UO₃, U₃O₈, and UO₂F₂) to evaluate the long-term storage characteristics of ²³³U. This report describes a similar study for alpha radiolysis. Uranium oxides and oxyfluorides (with ²³⁸U as the surrogate for ²³³U) were subjected to relatively high alpha radiation doses (235 to 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 that only very low pressures of hydrogen, nitrogen, and carbon dioxide are generated from the water, nitrate, and carbon impurities, respectively, associated with the oxides. 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 where the rate of generation of gaseous components is balanced by their recombination and/or consumption in the oxide phase. These results clearly demonstrate that alpha radiolysis of either wet or dry ²³³U oxides will not produce deleterious pressures or gaseous components that could compromise the long-term storage of these materials

[en] 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

[en] The SOURCE1 and SOURCE2 computer codes calculate source terms (i.e. radionuclide release rates) for performance assessments of low-level radioactive waste (LLW) disposal facilities. SOURCE1 is used to simulate radionuclide releases from tumulus-type facilities. SOURCE2 is used to simulate releases from silo-, well-, well-in-silo-, and trench-type disposal facilities. The SOURCE codes (a) simulate the degradation of engineered barriers and (b) provide an estimate of the source term for LLW disposal facilities. This manual summarizes the major changes that have been effected since the codes were originally developed