[en] The Waste Form Degradation Process Model Report (Waste Form Degradation PMR) is one of nine PMRs that have the shared objective of describing the technical information that was used in the total system performance assessment (TSPA) conducted to evaluate the postclosure performance of a potential monitored geologic repository at Yucca Mountain. The TSPA-SR (site recommendation) will be used in the preparation of a document for a Secretarial decision on whether to recommend the Yucca Mountain site for development as a repository. The Waste Form Degradation PMR summarizes the results of investigations on the degradation of the radioactive spent nuclear fuel (SNF), high-level waste (HLW), dissolved radionuclides, and colloidal radionuclides. The culmination of these investigations was the construction of the Waste Form Degradation Model to predict the dissolved or colloidal radionuclides available for transport in the TSPA for SR (TSPA-SR). The Waste Form Degradation Model consists of eight major modeling/analysis components: (1) Radioisotope Inventory, (2) In-Package Chemistry, (3) Commercial Spent Nuclear Fuel (CSNF) Degradation, (4) CSNF Cladding Degradation, (5) U.S. Department of Energy (DOE) Spent Nuclear Fuel (DSNF) Degradation, (6) HLW Degradation, (7) Radioisotope Dissolved Concentration (solubility), and (8) Radioisotope Colloidal Concentration. These eight components are generally connected sequentially starting with the radioisotope inventory as input and ending with projected radioisotope dissolved and colloidal concentration. Within the analysis for the TSPA-SR analysis, the function of the Waste Form Degradation Model is to determine three outputs over time: (1) dissolved concentration, (2) reversible colloidal concentration, and (3) irreversible colloidal concentration of radionuclides. This paper will describe and summarize the eight components of the Waste Form Degradation Process Model. (author)

[en] A backfill system has been designed for the Waste Isolation Pilot Plant (WIPP) which will control the chemical environment of the post-closure repository to a domain where the actinide solubility is within its lowest region. The actinide solubility is highly dependent on the chemical species which constitute the fluid, the resulting pH of the fluid, and the oxidation state of the actinide which is stable under the specific conditions. The use of magnesium oxide (MgO) has the backfill material not only controls the pH of the expected fluids, but also effectively removes carbonate from the system, which has a significant impact on actinide solubility. The backfill selection process, emplacement system design, and confirmatory experimental results are presented

[en] A Waste Form Degradation Process Model has been developed as part of the total system performance assessment (TSPA) that will be used to evaluate the postclosure performance of a potential monitoring geologic repository at Yucca Mountain. For the Colloidal Radioisotope Concentration Component of this degradation model, the conceptualization directly used Yucca Mountain Project (YMP)-relevant experimental results from YMP-specific work and from the published literature. The conceptualization identified the availability and the stability of three categories of colloids: (1) existing colloids in the groundwater, (2) colloids generated during degradation of the waste, and (3) colloids generated during degradation of the disposal container. The faction of the colloidal radioisotope concentration component is to calculate the concentration of colloid-associated radionuclides that may be transported from the waste package. Attachment of radionuclides can range from instantaneously reversible to completely irreversibly bound attachment. This process was modeled using these two bounding end members. For waste form colloids, radionuclides can be incorporated into the substrate material before it is suspended as colloids. All colloid types may sorb radionuclides to form pseudocolloids, depending on the affinity of the colloid mineral substrate for a dissolved radionuclide. The contributions of each colloid type are summed to produce the mobile colloid source term for each important radionuclide. The models are based on laboratory results from waste form corrosion testing and testing of adsorption and desorption properties of Pu and Am on clay and iron-(hydr)oxide colloids. To the extent that the laboratory tests and test conditions represent anticipated repository conditions, the model is valid for calculating the colloid-associated radionuclide concentrations and colloid mass concentrations. This paper will describe and summarize the models and experimental studies used in the colloidal radioisotope concentration component of this waste form degradation process model. (author)

[en] The Waste Isolation Pilot Plant (WPP) is located in southeastern New Mexico and is being developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. A detailed performance assessment (PA) for the WIPP was carried out in 1996 and supports an application by the DOE to the U.S. Environmental Protection Agency (EPA) for the certification of the WIPP for the disposal of TRU waste. The 1996 WIPP PA uses a computational structure that maintains a separation between stochastic (i.e., aleatory) and subjective (i.e., epistemic) uncertainty, with stochastic uncertainty arising from the many possible disruptions that could occur over the 10,000 yr regulatory period that applies to the WIPP and subjective uncertainty arising from the imprecision with which many of the quantities required in the PA are known. Important parts of this structure are (1) the use of Latin hypercube sampling to incorporate the effects of subjective uncertainty, (2) the use of Monte Carlo (i.e., random) sampling to incorporate the effects of stochastic uncertainty, and (3) the efficient use of the necessarily limited number of mechanistic calculations that can be performed to support the analysis. The use of Latin hypercube sampling generates a mapping from imprecisely known analysis inputs to analysis outcomes of interest that provides both a display of the uncertainty in analysis outcomes (i.e., uncertainty analysis) and a basis for investigating the effects of individual inputs on these outcomes (i.e., sensitivity analysis). The sensitivity analysis procedures used in the PA include examination of scatterplots, stepwise regression analysis, and partial correlation analysis. Uncertainty and sensitivity analysis results obtained as part of the 1996 WIPP PA are presented and discussed. Specific topics considered include two phase flow in the vicinity of the repository, radionuclide release from the repository, fluid flow and radionuclide transport in formations overlying the repository, and complementary cumulative distribution functions used in comparisons with regulatory standards (i.e., 40 CFR 191, Subpart B)

[en] A backfill system has been designed for the WIPP which will control the chemical environment of the post-closure repository to a domain where the actinide solubility is within its lowest region. The actinide solubility is highly dependent on the chemical species which constitute the fluid, the resulting pH of the fluid, and oxidation state of the actinide which is stable under the specific conditions. The implementation of magnesium oxide (MgO) as the backfill material not only controls the pH of the expected fluids but also effectively removes the carbonate from the system, which has a significant impact for actinide solubility. The selection process, emplacement system, design, and confirmatory experimental results are presented