Post-Irradiation Examination of Array Targets - Part I

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 {sup 236}Pu content and to evaluate fission product release data. One array consisted of seven targets that contained 10 vol% NpO{sub 2} pellets, while the other array consisted of seven targets that contained 20 vol % NpO{sub 2} 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 NpO{sub 2} 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 {sup 85}Kr. These nine targets and the associated dosimeters were then chemically processed to measure the residual neptunium, total plutonium production, {sup 238}Pu production, and {sup 236}Pu 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

Sphere-Pac Evaluation for Transmutation

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

Transport of Radioactive Material by Alpha Recoil

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

Alpha Radiolysis of Sorbed Water on Uranium Oxides and Uranium Oxyfluorides

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 (UO{sub 3}, U{sub 3}O{sub 8}, and UO{sub 2}F{sub 2}) to evaluate the long-term storage characteristics of {sup 233}U. This report describes a similar study for alpha radiolysis. Uranium oxides and oxyfluorides (with {sup 238}U as the surrogate for {sup 233}U) were subjected to relatively high alpha radiation doses (235 to 634 MGy) by doping with {sup 244}Cm. The typical irradiation time for these samples was about 1.5 years, which would be equivalent to more than 50 years irradiation by a {sup 233}U 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 {sup 233}U oxides will not produce deleterious pressures or gaseous components that could compromise the long-term storage of these materials

User`s Manual for the SOURCE1 and SOURCE2 Computer Codes: Models for Evaluating Low-Level Radioactive Waste Disposal Facility Source Terms (Version 2.0)

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

Water Sorption and Gamma Radiolysis Studies for Uranium Oxides

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

A Gamma Radiolysis Study of UO{sub 2}F{sub 2} 0.4H{sub 2}O Using Spent Nuclear Fuel Elements from the High Flux Isotope Reactor

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%

Water Sorption and Radiolysis Studies for Neptunium Oxides

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

Recent improvements to the SOURCE1 and SOURCE2 computer codes

Performance assessments of low-level radioactive waste (LLW) disposal facilities often involve the use of computer codes to describe radionuclide releases from a waste form and the subsequent transport of radionuclides through the environment. The SOURCE1 and SOURCE2 computer codes are used to calculate radionuclide release rates (i.e., source terms) for LLW disposal facilities. These codes have been used to evaluate the source terms for Oak Ridge National Laboratory performance assessments. SOURCE1 is applicable to tumulus-type facilities, while SOURCE2 can be applied to silo, well-in-silo, well, and trench-type facilities. In addition to the calculation of radionuclide release rates, both SOURCE1 and SOURCE2 calculate the degradation of engineered barriers. This paper provides an overview of these codes and a description of recent improvements to the codes. Major improvements include incorporation of a new advective transport model into SOURCE1 and SOURCE2, development of a new model for SOURCE1 that calculates the degradation and failure of the tumulus pad and leachate collection system, improvement of routines for controlling water infiltration inputs, expansion of options for obtaining output summaries, and restructuring of SOURCE1 and SOURCE2 for sensitivity and uncertainty analyses. The status of code verification efforts is also presented

Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

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