Electron Microscopy Characterization of Tc-Bearing Metallic Waste Forms- Final Report FY10

The DOE Fuel Cycle Research & Development (FCR&D) Program is developing aqueous and electrochemical approaches to the processing of used nuclear fuel that will generate technetium-bearing waste streams. This final report presents Pacific Northwest National Laboratory (PNNL) research in FY10 to evaluate an iron-based alloy waste form for Tc that provides high waste loading within waste form processing limitations, meets waste form performance requirements for durability and the long-term retention of radionuclides and can be produced with consistent physical, chemical, and radiological properties that meet regulatory acceptance requirements for disposal

Nanoscale oxygen defect gradients in UO2+x surfaces.

Oxygen defects govern the behavior of a range of materials spanning catalysis, quantum computing, and nuclear energy. Understanding and controlling these defects is particularly important for the safe use, storage, and disposal of actinide oxides in the nuclear fuel cycle, since their oxidation state influences fuel lifetimes, stability, and the contamination of groundwater. However, poorly understood nanoscale fluctuations in these systems can lead to significant deviations from bulk oxidation behavior. Here we describe the use of aberration-corrected scanning transmission electron microscopy and electron energy loss spectroscopy to resolve changes in the local oxygen defect environment in [Formula: see text] surfaces. We observe large image contrast and spectral changes that reflect the presence of sizable gradients in interstitial oxygen content at the nanoscale, which we quantify through first-principles calculations and image simulations. These findings reveal an unprecedented level of excess oxygen incorporated in a complex near-surface spatial distribution, offering additional insight into defect formation pathways and kinetics during [Formula: see text] surface oxidation

Nanoscale oxygen defect gradients in UO_2+x surfaces

Oxygen defects govern the behavior of a range of materials spanning catalysis, quantum computing, and nuclear energy. Understanding and controlling these defects is particularly important for the safe use, storage, and disposal of actinide oxides in the nuclear fuel cycle, since their oxidation state influences fuel lifetimes, stability, and the contamination of groundwater. However, poorly understood nanoscale fluctuations in these systems can lead to significant deviations from bulk oxidation behavior. Here we describe the use of aberration-corrected scanning transmission electron microscopy and electron energy loss spectroscopy to resolve changes in the local oxygen defect environment in&nbsp;</p

Nanoscale Oxygen Defect Gradients in the Actinide Oxides

Oxygen defects govern the behavior of a range of materials spanning catalysis, quantum computing, and nuclear energy. Understanding and controlling these defects is particularly important for the safe use, storage, and disposal of actinide oxides in the nuclear fuel cycle, since their oxidation state influences fuel lifetimes, stability, and the contamination of groundwater. However, poorly understood nanoscale fluctuations in these systems can lead to significant deviations from bulk oxidation behavior. Here we describe the first use of aberration-corrected scanning transmission electron microscopy and electron energy loss spectroscopy to resolve changes in the local oxygen defect environment in UO$_2$ surfaces. We observe large image contrast and spectral changes that reflect the presence of sizable gradients in interstitial oxygen content at the nanoscale, which we quantify through first principles calculations and image simulations. These findings reveal an unprecedented level of excess oxygen incorporated in a complex near-surface spatial distribution, offering new insight into defect formation pathways and kinetics during UO$_2$ oxidation.Comment: 26 pages, 12 figure

Coupling the Mixed Potential and Radiolysis Models for Used Fuel Degradation

The primary purpose of this report is to describe the strategy for coupling three process level models to produce an integrated Used Fuel Degradation Model (FDM). The FDM, which is based on fundamental chemical and physical principals, provides direct calculation of radionuclide source terms for use in repository performance assessments. The G-value for H2O2 production (Gcond) to be used in the Mixed Potential Model (MPM) (H2O2 is the only radiolytic product presently included but others will be added as appropriate) needs to account for intermediate spur reactions. The effects of these intermediate reactions on [H2O2] are accounted for in the Radiolysis Model (RM). This report details methods for applying RM calculations that encompass the effects of these fast interactions on [H2O2] as the solution composition evolves during successive MPM iterations and then represent the steady-state [H2O2] in terms of an “effective instantaneous or conditional” generation value (Gcond). It is anticipated that the value of Gcond will change slowly as the reaction progresses through several iterations of the MPM as changes in the nature of fuel surface occur. The Gcond values will be calculated with the RM either after several iterations or when concentrations of key reactants reach threshold values determined from previous sensitivity runs. Sensitivity runs with RM indicate significant changes in G-value can occur over narrow composition ranges. The objective of the mixed potential model (MPM) is to calculate the used fuel degradation rates for a wide range of disposal environments to provide the source term radionuclide release rates for generic repository concepts. The fuel degradation rate is calculated for chemical and oxidative dissolution mechanisms using mixed potential theory to account for all relevant redox reactions at the fuel surface, including those involving oxidants produced by solution radiolysis and provided by the radiolysis model (RM). The RM calculates the concentration of species generated at any specific time and location from the surface of the fuel. Several options being considered for coupling the RM and MPM are described in the report. Different options have advantages and disadvantages based on the extent of coding that would be required and the ease of use of the final product

Technetium Waste Form Development - Progress Report

Analytical electron microscopy using SEM and TEM has been used to analyze a ~5 g. ingot with composition 71.3 wt% 316SS-5.3 wt% Zr-13.2 wt% Mo-4.0 wt% Rh-6.2 wt% Re prepared at the Idaho National Laboratory. Four phase fields have been identified two of which are lamellar eutectics, with a fifth possibly present. A Zr rich phase was found distributed as fine precipitate, ~10µm in diameter, often coating large cavities. A Mo-Fe-Re-Cr lamellar eutectic phase field appears as blocky regions ~30µm in diameter, surrounded by a Fe-Mo-Cr lamellar eutectic phase field, and that in turn is surrounded by a Zr-Fe-Rh-Mo-Ni phase field. The eutectic phase separation reactions are different. The Mo-Fe-Re-Cr lamellar eutectic appears a result of austenitic steel forming at lower volume fraction within an Mo-Fe-Re intermetallic phase, whereas the Fe-Mo-Cr lamellar eutectic may be a result of the same intermetallic phase forming within a ferritic steel phase. Cavitation may have arisen either as a result of bubbles, or from loss of equiaxed particles during specimen preparation

Radiolysis Process Model

Assessing the performance of spent (used) nuclear fuel in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO2, is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water (including OH• and H• radicals, O2-, eaq, H2O2, H2, and O2) that may increase the waste form degradation rate and change radionuclide behavior. H2O2 is the dominant oxidant for spent nuclear fuel in an O2 depleted water environment, the most sensitive parameters have been identified with respect to predictions of a radiolysis model under typical conditions. As compared with the full model with about 100 reactions it was found that only 30-40 of the reactions are required to determine [H2O2] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation

Radiolysis Process Model

Assessing the performance of spent (used) nuclear fuel in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO2, is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water (including OH• and H• radicals, O2-, eaq, H2O2, H2, and O2) that may increase the waste form degradation rate and change radionuclide behavior. H2O2 is the dominant oxidant for spent nuclear fuel in an O2 depleted water environment, the most sensitive parameters have been identified with respect to predictions of a radiolysis model under typical conditions. As compared with the full model with about 100 reactions it was found that only 30-40 of the reactions are required to determine [H2O2] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation

Getters for improved technetium containment in cementitious waste forms.

A cementitious waste form, Cast Stone, is a possible candidate technology for the immobilization of low activity nuclear waste (LAW) at the Hanford site. This work focuses on the addition of getter materials to Cast Stone that can sequester Tc from the LAW, and in turn, lower Tc release from the Cast Stone. Two getters which produce different products upon sequestering Tc from LAW were tested: Sn(II) apatite (Sn-A) that removes Tc as a Tc(IV)-oxide and potassium metal sulfide (KMS-2) that removes Tc as a Tc(IV)-sulfide species, allowing for a comparison of stability of the form of Tc upon entering the waste form. The Cast Stone with KMS-2 getter had the best performance with addition equivalent to ∼0.08wt% of the total waste form mass. The observed diffusion (Dobs) of Tc decreased from 4.6±0.2×10-12cm2/s for Cast Stone that did not contain a getter to 5.4±0.4×10-13cm2/s for KMS-2 containing Cast Stone. It was found that Tc-sulfide species are more stable against re-oxidation within getter containing Cast Stone compared with Tc-oxide and is the origin of the decrease in Tc Dobs when using the KMS-2