Ernest Hardin

Waste package degradation from thermal and chemical processes in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

Abstract This paper summarizes modeling of waste container degradation in performance assessments conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. As understanding of the Yucca Mountain disposal system increased, modeling of container degradation evolved from a component of the source term in 1984 to a separate module describing both container and drip shield degradation in 2008. A thermal module for evaluating the influence of higher heat loads from more closely packed, large waste packages was also introduced. In addition, a module for evaluating drift chemistry was added in later PAs to evaluate the potential for localized corrosion of the outer barrier of the waste container composed of Alloy 22, a highly corrosion-resistant nickel–chromium–tungsten–molybdenum alloy. The uncertainty of parameters related to container degradation contributed significantly to the estimated uncertainty of performance measures (cumulative release in assessments prior to 1995 and individual dose, thereafter).

Overview of Total System Model Used for the 2008 Performance Assessment for the Proposed High-Level Radioactive Waste Repository at Yucca Mountain Nevada.

Abstract A summary is presented for the total system model used to represent physical processes associated with the seven scenario classes (i.e., nominal conditions, early waste package (WP) failure, early drip shield (DS) failure, igneous intrusive events, igneous eruptive events, seismic ground motion events and seismic fault displacement events) considered in the 2008 performance assessment for the proposed repository for high-level radioactive waste at Yucca Mountain, Nevada. The total system model estimates dose to an exposed individual resulting from radionuclide movement through the repository system and biosphere. Components of the total system model described in this presentation include models for (i) climate analysis, (ii) land surface infiltration and associated unsaturated zone flow, (iii) multi-scale thermal hydrology and engineered barrier system (EBS) thermal–hydrologic environment, (iv) EBS physical and chemical environment, (v) WP and DS degradation, (vi) drift seepage and drift wall condensation, (vii) waste form degradation and mobilization, (viii) water and radionuclide movement in the EBS and underlying unsaturated and saturated zones, (ix) radionuclide movement in the biosphere and resultant human exposure, and (x) processes specific to early WP and DS failures, intrusive and eruptive igneous events, and seismic ground motion and fault displacement events.

Thermal-Hydrology Simulations of Disposal of High-Level Radioactive Waste in a Single Deep Borehole

Simulations of thermal-hydrology were carried out for the emplacement of spent nuclear fuel canisters and cesium and strontium capsules using the PFLOTRAN simulator. For the cesium and strontium capsules the analysis looked at disposal options such as different disposal configurations and surface aging of waste to reduce thermal effects. The simulations studied temperature and fluid flux in the vicinity of the borehole. Simulation results include temperature and vertical flux profiles around the borehole at selected depths. Of particular importance are peak temperature increases, and fluxes at the top of the disposal zone. Simulations of cesium and strontium capsule disposal predict that surface aging and/or emplacement of the waste at the top of the disposal zone reduces thermal effects and vertical fluid fluxes. Smaller waste canisters emplaced over a longer disposal zone create the smallest thermal effect and vertical fluid fluxes no matter the age of the waste or depth of emplacement.

Used Fuel Management System Architecture and Interface Analyses

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications.Copyright

Proceedings of the scientific visit on crystalline rock repository development.

.................................................................................................................................. iii ACKNOWLEDGEMENTS ........................................................................................................... iv TABLE OF CONTENTS ................................................................................................................ v 1.0 OVERVIEW ............................................................................................................................. 1 2.0 PARTICIPANTS ...................................................................................................................... 2 3.0 PRESENTATIONS................................................................................................................... 3 3.1 Welcome ...................................................................................................................... 3 3.2 Status Reports .............................................................................................................. 3 3.2.1 Czech Republic ............................................................................................. 3 3.2.2 Germany ....................................................................................................... 4 3.2.3 Korea ............................................................................................................ 5 3.2.4 Lithuania ....................................................................................................... 5 3.2.5 Mexico .......................................................................................................... 6 3.2.6 Pakistan ........................................................................................................ 6 3.2.7 Poland ........................................................................................................... 6 3.2.8 Romania ........................................................................................................ 6 3.2.9 Slovak Republic ........................................................................................... 7 3.2.10 Spain ............................................................................................................................ 7 3.2.11 Switzerland ................................................................................................... 8 3.2.12 United Kingdom ........................................................................................... 8 3.2.13 USA .............................................................................................................. 9 3.2.14 Discussion .................................................................................................... 9 3.3 Development of a Repository Programme ................................................................ 13 3.3.1 Overview of Repository Programme Development ................................... 13 3.3.2 Technical Bases .......................................................................................... 13 3.3.3 Design Concepts ......................................................................................... 14 3.3.4 Safety Assessment Methodology ............................................................... 16 3.3.5 Performance Assessments and R&D Prioritization .................................... 17 3.4 Site Selection and Site Characterization ................................................................... 18 Proceedings of Scientific Visit on Crystalline Rock Repository Development February 28, 2013 vi 4.0 FINALE .................................................................................................................................. 25 5.0 SESSION PHOTOS ................................................................................................................ 27 6.0 JOSEF GALLERY TOUR ...................................................................................................... 29 7.0 REFERENCES ....................................................................................................................... 31 APPENDIX A: PROGRAMME FOR THE SCIENTIFIC VISIT ON CRYSTALLINE ROCK REPOSITORY DEVELOPMENT ............................................................................................. A-1 APPENDIX B: PARTICIPANTS ............................................................................................... B-1 APPENDIX C: PRESENTATION SLIDES ............................................................................... C-