Publications

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This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United States repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, helps define a clear strategy for a heat-generating nuclear waste repository in salt.

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Regulatory compliance and effective management of the nation`s TRU waste requires knowledge about the constituents present in the waste. With limited resources, the DOE needs a cost-effective characterization program. In addition, the DOE needs a method for predicting the present and future analytical requirements for waste characterization. Thus, a strategy for predicting the present and future waste characterization needs that uses current knowledge of the TRU inventory and prioritization of the data needs is presented.

NEFTRAN-S was developed by Sandia National Laboratories for the United States Environmental Protection Agency as part of a program providing technical support for re-promulgation of the standard 40 CFR 191. The code is intended to provide realistic estimates of releases to the environment that could result from disposal of radioactive waste in geologic subsurfaces. One of the geologic environments that will be considered by the EPA in their analyses is unsaturated tuff. The information given in this report is intended to provide a conceptual model for the NEFTRAN-S code for calculations involving a generic site in unsaturated tuff. Information about the phenomena expected to dominate transport and methods for modeling transport in an unsaturated medium are presented. NEFTRAN-S calculations using this conceptual model are compared to TOSPAC calculations for three possible infiltration rates. TOSPAC is the code currently used in performance assessment for an unsaturated tuff site at Yucca Mountain in Nevada. 14 refs., 21 figs., 22 tabs.

This document describes the NEFTRAN-S computer code and is intended to provide the reader with enough information to use the code. NEFTRAN-S was developed for the United States Environmental Protection Agency for the assessment of ground-water flow and radionuclide transport from radioactive waste disposal in geologic formations. NEFTRAN-S is a successor to the NEFTRAN code. The code was developed in conjunction with NEFTRAN-2, which was developed recently for the United States Nuclear Regulatory Commission. As a result, some of the features contained in NEFTRAN-2 have been included in NEFTRAN-S. In particular, NEFTRAN-S includes an exponential-leach-rate source, decoupled time steps for source and transport, and an option for inputting pore-water velocities. Features unique to NEFTRAN-S include a user-friendly format for use on personal computers and coupling with statistical sampling and analysis using the SUNS software shell. This document was written to provide a comprehensive discussion of the NEFTRAN-S code including its history, the theory, its use and examples of possible applications. Minimal reference to previous documents is intended. 25 refs., 132 figs., 30 tabs.