Publications

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Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

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Sandia National Laboratories (SNL) and Taiwan's Institute for Nuclear Energy Research (INER) have teamed together to evaluate several candidate sites for Low-Level Radioactive Waste (LLW) disposal in Taiwan. Taiwan currently has three nuclear power plants, with another under construction. Taiwan also has a research reactor, as well as medical and industrial wastes to contend with. Eventually the reactors will be decomissioned. Operational and decommissioning wastes will need to be disposed in a licensed disposal facility starting in 2014. Taiwan has adopted regulations similar to the US Nuclear Regulatory Commission's (NRC's) low-level radioactive waste rules (10 CFR 61) to govern the disposal of LLW. Taiwan has proposed several potential sites for the final disposal of LLW that is now in temporary storage on Lanyu Island and on-site at operating nuclear power plants, and for waste generated in the future through 2045. The planned final disposal facility will have a capacity of approximately 966,000 55-gallon drums. Taiwan is in the process of evaluating the best candidate site to pursue for licensing. Among these proposed sites there are basically two disposal concepts: shallow land burial and cavern disposal. A representative potential site for shallow land burial is located on a small island in the Taiwan Strait with basalt bedrock and interbedded sedimentary rocks. An engineered cover system would be constructed to limit infiltration for shallow land burial. A representative potential site for cavern disposal is located along the southeastern coast of Taiwan in a tunnel system that would be about 500 to 800 m below the surface. Bedrock at this site consists of argillite and meta-sedimentary rocks. Performance assessment analyses will be performed to evaluate future performance of the facility and the potential dose/risk to exposed populations. Preliminary performance assessment analyses will be used in the site-selection process and to aid in design of the disposal system. Final performance assessment analyses will be used in the regulatory process of licensing a site. The SNL/INER team has developed a performance assessment methodology that is used to simulate processes associated with the potential release of radionuclides to evaluate these sites. The following software codes are utilized in the performance assessment methodology: GoldSim (to implement a probabilistic analysis that will explicitly address uncertainties); the NRC's Breach, Leach, and Transport - Multiple Species (BLT-MS) code (to simulate waste-container degradation, waste-form leaching, and transport through the host rock); the Finite Element Heat and Mass Transfer code (FEHM) (to simulate groundwater flow and estimate flow velocities); the Hydrologic Evaluation of Landfill performance Model (HELP) code (to evaluate infiltration through the disposal cover); the AMBER code (to evaluate human health exposures); and the NRC's Disposal Unit Source Term -- Multiple Species (DUST-MS) code (to screen applicable radionuclides). Preliminary results of the evaluations of the two disposal concept sites are presented.

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The proposed Yucca Mountain repository is anticipated to be the first facility for long-term disposal of commercial spent nuclear fuel and high-level radioactive waste in the United States. The facility, located in the southern Nevada desert, is currently in the planning stages with initial exploratory excavations completed. It is an underground facility mined into the tuffaceous volcanic rocks that sit above the local water table. The focus of the work described in this paper is the development of radionuclide absorbers or 'getter' materials for neptunium (Np), iodine (I), and technetium (Tc) for potential deployment in the repository. 'Getter' materials retard the migration of radionuclides through sorption, reduction, or other chemical and physical processes, thereby slowing or preventing the release and transport of radionuclides. An overview of the objectives and approaches utilized in this work with respect to materials selection and modeling of ion 'getters' is presented. The benefits of the 'getter' development program to the United States Department of Energy (US DOE) are outlined.

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The conceptual structure of the 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) is described. This structure involves three basic entities (EN1, EN2, EN3): (i) EN1, a probabilistic characterization of the likelihood of different futures occurring at the WIPP site over the next 10,000 years; (ii) EN2, a procedure for estimating the radionuclide releases to the accessible environment associated with each of the possible futures that could occur at the WIPP site over the next 10,000 years; and (iii) EN3, a probabilistic characterization of the uncertainty in the parameters used in the definition of EN1 and EN2. In the formal development of the 1996 WIPP PA, ENI is characterized by a probability space (Sst, Gst, pst) for stochastic (i.e. aleatory) uncertainty; EN2 is characterized by a function f that corresponds to the models and associated computer programs used to estimate radionuclide releases: and EN3 is characterized by a probability space (Ssu, Gsu, psu) for subjective (i.e. epistemic) uncertainty. A high-level overview of the 1996 WIPP PA and references to additional sources of information are given in the context of (Sst, Gst, pst), f and (Ssu, Gsu, psu).

The Waste Isolation Pilot Plant (WIPP) is under development by the US Department of Energy (DOE) for the geologic disposal of transuranic waste. The construction of complementary cumulative distribution functions (CCDFs) for total radionuclide release from the WIPP to the accessible environment is described. The resultant CCDFs (i) combine releases due to cuttings and cavings, spallings, direct brine release, and long-term transport in flowing groundwater; (ii) fall substantially to the left of the boundary line specified by the US Environmental Protection Agency's (EPA's) standard 40 CFR 191 for the geologic disposal of radioactive waste; and (iii) constitute an important component of the DOE's successful Compliance Certification Application to the EPA for the WIPP. Insights and perspectives gained in the performance assessment (PA) that led to these CCDFs are described, including the importance of: (i) an iterative approach to PA; (ii) uncertainty and sensitivity analysis; (iii) a clear conceptual model for the analysis; (iv) the separation of stochastic (i.e. aleatory) and subjective (i.e. epistemic) uncertainty; (v) quality assurance procedures; (vi) early involvement of peer reviewers, regulators, and stakeholders; (vii) avoidance of conservative assumptions; and (viii) adequate documentation.