This report represents completion of milestone deliverable M2SF-22SN010309082 Annual Status Update for OWL, which is due on November 30, 2021 as part of the fiscal year 2022 (FY2022) work package SF-22SN01030908. This report provides an annual update on status of FY2021 activities for the work package “OWL - Inventory – SNL”. The Online Waste Library (OWL) has been designed to contain information regarding United States (U.S.) Department of Energy (DOE)-managed (as) high-level waste (DHLW), DOE-managed spent nuclear fuel (DSNF), and other wastes that are likely candidates for deep geologic disposal. Links to the current supporting documents for the data are provided when possible; however, no classified or official-use-only (OUO) data are planned to be included in OWL. There may be up to several hundred different DOE-managed wastes that are likely to require deep geologic disposal. This report contains new information on sodium-bonded spent fuel waste types and wastes forms, which are included in the next release of OWL, Version 3.0, on the Sandia National Laboratories (SNL) External Collaboration Network (ECN). The report also provides an update on the effort to include information regarding the types of vessels capable of disposing of DOE-managed waste.
In the planning for FY2020 in the U.S. DOE NE-81 Spent Fuel and Waste Science and Technology (SFWST) Campaign, the DOE requested development of a plan for activities in the Disposal Research (DR) Research and Development (R&D) over a five (5)-year period, and DOE requested periodic updates to this plan. The DR R&D 5-year plan was provided to the DOE based on the FY2020 priorities and program structure (Sassani et al., 2020) and represents a strategic guide to the work within the DR R&D technical areas (i.e., the Control Accounts), focusing on the highest priority technical thrusts. This FY2021 report is the first update to the DR R&D 5-year plan for the SFWST Campaign DR R&D activities. This 5-year plan will be a living document and is planned to be updated periodically to provide review of accomplishments and for prioritization changes based on aspects including mission progress, external technical work, and changes in SFWST Campaign objectives and/or funding levels (i.e., Program Direction). The updates to this 5-year plan will address the DR R&D that has been completed (accomplishments) and the additional knowledge gaps to be investigated, with any updates to the DR R&D priorities for the next stages of activities.
This report represents the milestone deliverable M4SF-21SN010309021 “Modeling Activities Related to Waste Form Degradation: Progress Report” that describes the progress of R&D activities of ongoing modeling investigations specifically on nuclear waste glass degradation, Density Functional Theory (DFT) studies on clarkeite structure and stability, and electrochemical modeling of spent nuclear fuel (SNF). These activities are part of the newly-created Waste form Testing, Modeling, and Performance work package at Sandia National Laboratories (SNL). This work package is part of the “Inventory and Waste Form Characteristics and Performance” control account that includes various experimental and modeling activities on nuclear waste degradation conducted at Oak Ridge National Laboratory (ORNL), SNL, Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL).
This report represents completion of milestone deliverable M2SF-21SN010309012 “Annual Status Update for OWL and Waste Form Characteristics” that provides an annual update on status of fiscal year (FY 2020) activities for the work package SF-20SN01030901 and is due on January 29, 2021. The Online Waste Library (OWL) has been designed to contain information regarding United States (U.S.) Department of Energy (DOE)-managed (as) high-level waste (DHLW), spent nuclear fuel (SNF), and other wastes that are likely candidates for deep geologic disposal, with links to the current supporting documents for the data (when possible; note that no classified or official-use-only (OUO) data are planned to be included in OWL). There may be up to several hundred different DOE-managed wastes that are likely to require deep geologic disposal. This draft report contains versions of the OWL model architecture for vessel information (Appendix A) and an excerpt from the OWL User’s Guide (Appendix B and SNL 2020), which are for the current OWL Version 2.0 on the Sandia External Collaboration Network (ECN).
The Online Waste Library (OWL) provides a consolidated source of information on Department of Energy-managed radioactive waste likely to require deep geologic disposal. With the release of OWL Version 1.0 in fiscal year 2019 (FY2019), much of the FY2020 work involved developing the OWL change control process and the OWL release process. These two processes (in draft form) were put into use for OWL Version 2.0, which was released in early FY2021. With the knowledge gained, the OWL team refined and documented the two processes in two separate reports. This report focuses on the change control process and discusses the following: (1) definitions and system components; (2) roles and responsibilities; (3) origin of changes; (4) the change control process including the Change List, Task List, activity categories, implementation examples, and checking and review; and (5) the role of the re lease process in ensuring changes in the Change List are incorporated into a public release.
The Online Waste Library (OWL) provides one consolidated source of information on Department of Energy-managed wastes likely to require deep geologic disposal. With the release of OWL Version 1.0 in fiscal year (FY) 2019, much of the FY2020 work involved developing the OWL change control process and the OWL release process. These two processes (in draft form) were put into use for OWL Version 2.0, which was released in early FY2021. With the knowledge gained, the OWL team refined and documented the two processes in two separate reports. This report addresses the release process starting with a definition of release management in Section 2. Section 3 describes the Information Technology Infrastructure Library (ITIL) framework, part of which includes the three different environments used for release management. Section 4 presents the OWL components existing in the different environments and provides details on the release schedule and procedures.
Disposal of large, heat-generating waste packages containing the equivalent of 21 pressurized water reactor (PWR) assemblies or more is among the disposal concepts under investigation for a future repository for spent nuclear fuel (SNF) in the United States. Without a long (>200 years) surface storage period, disposal of 21-PWR or larger waste packages (especially if they contain high-burnup fuel) would result in in-drift and near-field temperatures considerably higher than considered in previous generic reference cases that assume either 4-PWR or 12-PWR waste packages (Jové Colón et al. 2014; Mariner et al. 2015; 2017). Sevougian et al. (2019c) identified high-temperature process understanding as a key research and development (R&D) area for the Spent Fuel and Waste Science and Technology (SFWST) Campaign. A two-day workshop in February 2020 brought together campaign scientists with expertise in geology, geochemistry, geomechanics, engineered barriers, waste forms, and corrosion processes to begin integrated development of a high-temperature reference case for disposal of SNF in a mined repository in a shale host rock. Building on the progress made in the workshop, the study team further explored the concepts and processes needed to form the basis for a high-temperature shale repository reference case. The results are described in this report and summarized..
This report represents completion of milestone deliverable M2SF-19SNO10309013 "Online Waste Library (OWL) and Waste Forms Characteristics Annual Report" that reports annual status on fiscal year (FY) 2019 activities for the work package SF-19SN01030901 and is due on August 2, 2019. The online waste library (OWL) has been designed to contain information regarding United States (U.S.) Department of Energy (DOE)-managed (as) high-level waste (DHLW), spent nuclear fuel (SNF), and other wastes that are likely candidates for deep geologic disposal, with links to the current supporting documents for the data (when possible; note that no classified or official-use-only (OUO) data are planned to be included in OWL). There may be up to several hundred different DOE-managed wastes that are likely to require deep geologic disposal. This annual report on FY2019 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the beta testing and changes implemented from it. Also added are descriptions of the management/control processes for the OWL development, version control, and archiving. These processes have been implemented as part of the full production release of OWL (i.e., OWL Version 1.0), which has been developed on, and will be hosted and managed on, Sandia National Laboratories (SNL) systems. The version control/update processes will be implemented for updates to the OWL in the future. Additionally, another process covering methods for interfacing with the DOE SNF Database (DOE 2007) at Idaho National Laboratory on the numerous entries for DOE-managed SNF (DSNF) has been pushed forward by defining data exchanges and is planned to be implemented sometime in FY2020. The INL database is also sometimes referred to as the Spent Fuel Database or the SFDB, which is the acronym that will be used in this report. Once fully implemented, this integration effort will serve as a template for interfacing with additional databases throughout the DOE complex.
The safety case for deep borehole disposal of nuclear wastes contains a safety strategy, an assessment basis, and a safety assessment. The safety strategy includes strategies for management, siting and design, and assessment. The assessment basis considers site selection, pre-closure, and post-closure, which includes waste and engineered barriers, the geosphere/natural barriers, and the biosphere and surface environment. The safety assessment entails a pre-closure safety analysis, a post-closure performance assessment, and confidence enhancement analyses. This paper outlines the assessment basis and safety assessment aspects of a deep borehole disposal safety case. The safety case presented here is specific to deep borehole disposal of Cs and Sr capsules, but is generally applicable to other waste forms, such as spent nuclear fuel. The safety assessments for pre-closure and post-closure are briefly summarized from other sources; key issues for confidence enhancement are described in greater detail. These confidence enhancement analyses require building the technical basis for geologically old, reducing, highly saline brines at the depth of waste emplacement, and using reactive-transport codes to predict their movement in post-closure. The development and emplacement of borehole seals above the waste emplacement zone is also important to confidence enhancement.
This report describes the current status of the safety case for the deep borehole disposal (DBD) concept. It builds on the safety case presented in Freeze et al. (2016), presenting new information and identifying additional information needs for specific safety case elements. At this preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. Updated information is provided for the following safety case elements: * pre-closure basis and safety analysis, * post-closure basis and performance assessment, and * confidence enhancement. This research was performed as part of the deep borehole field test (DBFT). Based on revised U.S. Department of Energy (DOE) priorities in mid-2017, the DBFT and other research related to a DBD option was discontinued; ongoing work and documentation were closed out by the end of fiscal year (FY) 2017. This report was initiated as part of the DBFT and documented as an incomplete draft at the end of FY 2017. The report was finalized by Sandia National Laboratories in FY2018 without DOE funding, subsequent to the termination of the DBFT, and published in FY2019. iii
The U.S. Department of Energy is conducting research and development on generic concepts for disposal of spent nuclear fuel and high-level radioactive waste in multiple lithologies, including salt, crystalline (granite/metamorphic), and argillaceous (clay/shale) host rock. These investigations benefit greatly from international experience gained in disposal programs in many countries around the world. The focus of this study is the post-closure degradation and radionuclide-release rates for tristructural-isotropic (TRISO) coated particle spent fuels for various generic geologic repository environments.1,2,3 The TRISO particle coatings provide safety features during and after reactor operations, with the SiC layer representing the primary barrier. Three mechanisms that may lead to release of radionuclides from the TRISO particles are: (1) helium pressure buildup4 that may eventually rupture the SiC layer, (2) diffusive transport through the layers (solid-state diffusion in reactor, aqueous diffusion in porous media at repository conditions), and (3) corrosion5 of the layers in groundwater/brine. For TRISO particles in a graphite fuel element, the degradation in an oxidizing geologic repository was concluded to be directly dependent on the oxidative corrosion rate of the graphite matrix4, which was analyzed as much slower than SiC layer corrosion processes. However, accumulated physical damage to the graphite fuel element may decrease its post-closure barrier capability more rapidly. Our initial performance model focuses on the TRISO particles and includes SiC failure from pressure increase via alpha-decay helium, as exacerbated by SiC layer corrosion5. This corrosion mechanism is found to be much faster than solid-state diffusion at repository temperatures but includes no benefit of protection by the other outer layers, which may prolong lifetime. Our current model enhancements include constraining the material properties of the layers for porous media diffusion analyses. In addition to evaluating the SiC layer porosity structure, this work focuses on the pyrolytic carbon layers (inner/outer-IPyC/OPyC) layers, and the graphite compact, which are to be analyzed with the SiC layer in two modes: (a) intact SiC barrier until corrosion failure and (b) SiC with porous media transport. Our detailed performance analyses will consider these processes together with uncertainties in the properties of the layers to assess radionuclide release from TRISO particles and their graphite compacts.
TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 lam thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 degC). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10 -20 m 2 /s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10 -21 m 2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.
This report represents completion of milestone deliverable M2SF-18SNO10309013 "Inventory and Waste Characterization Status Report and OWL Update that reports on FY2018 activities for the work package (WP) SF-18SNO1030901. This report provides the detailed final information for completed FY2018 work activities for WP SF-18SN01030901, and a summary of priorities for FY2019. This status report on FY2018 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the two planned management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the Beta testing and changes implemented from it. There are two processes being planned in FY2018, which will be implemented in FY2019. One process covers methods for interfacing with the DOE SNF DB (DOE 2007) at INL on the numerous entries for DOE managed SNF, and the other process covers the management of updates to, and version control/archiving of, the OWL database. In FY2018, we have pursued three studies to evaluate/redefine waste form characteristics and/or performance models. First characteristic isotopic ratios for various waste forms included in postclosure performance studies are being evaluated to delineate isotope ratio tags that quantitatively identify each particular waste form. This evaluation arose due to questions regarding the relative contributions of radionuclides from disparate waste forms in GDSA results, particularly, radionuclide contributions of DOE-managed SNF vs HLW glass. In our second study we are evaluating the bases of glass waste degradation rate models to the HIP calcine waste form. The HIP calcine may likely be a ceramic matrix material, with multiple ceramic phases with/without a glass phase. The ceramic phases are likely to have different degradation performance from the glass portion. The distribution of radionuclides among those various phases may also be a factor in the radionuclide release rates. Additionally, we have an ongoing investigation of the performance behavior of TRISO particle fuels and are developing a stochastic model for the degradation of those fuels that accounts for simultaneous corrosion of the silicon carbide (SiC) layer and radionuclide diffusion through it. The detailed model of the TRISO particles themselves, will be merged with models of the degradation behavior(s) of the graphite matrix (either prismatic compacts or spherical "pebbles") containing the particles and the hexagonal graphite elements holding the compacts.
This report summarizes the 2018 fiscal year (FY18) field, laboratory, and modeling work funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign as part of the Sandia National Laboratories Salt Research and Development (R&D) and Salt International work packages. This report satisfies level-two milestone M2SF-18SNO10303031and comprises three related but stand-alone sections. The first section summarizes the programmatic progress made to date in the DOE-NE salt program and its goals going forward. The second section presents brine composition modeling and laboratory activities related to salt evaporation experiments, which will be used to interpret data collected during the heater test. The third section presents theoretical and numerical modeling work done to investigate the effects brine composition have on dihedral angle and the permeability of salt.
This report provides an update to Sassani et al. (2016) and includes: (1) an updated set of inputs (Sections 2.3) on various additional waste forms (WF) covering both DOE-managed spent nuclear fuel (SNF) and DOE-managed (as) high-level waste (HLW) for use in the inventory represented in the geologic disposal safety analyses (GDSA); (2) summaries of evaluations initiated to refine specific characteristics of particular WF for future use (Section 2.4); (3) updated development status of the Online Waste Library (OWL) database (Section 3.1.2) and an updated user guide to OWL (Section 3.1.3); and (4) status updates (Section 3.2) for the OWL inventory content, data entry checking process, and external OWL BETA testing initiated in fiscal year 2017.
ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal
The Deep Borehole Field Test (DBFT) is a planned multi-year project led by the US Department of Energy's Office of Nuclear Energy to drill two boreholes to 5 km total depth into crystalline basement in the continental US. The purpose of the first characterization borehole is to demonstrate the ability to characterize in situ formation fluids through sampling and perform downhole hydraulic testing to demonstrate groundwater from 3 to 5 km depth is old and isolated from the atmosphere. The purpose of the second larger-diameter borehole is to demonstrate safe surface and downhole handling procedures. This paper details many of the drilling, testing, and characterization activities planned in the first smaller-diameter characterization borehole.
The Waste Form Disposal Options Evaluation Report (SNL 2014) evaluated disposal of both Commercial Spent Nuclear Fuel (CSNF) and DOE-managed HLW and Spent Nuclear Fuel (DHLW and DSNF) in the variety of disposal concepts being evaluated within the Used Fuel Disposition Campaign. That work covered a comprehensive inventory and a wide range of disposal concepts. The primary goal of this work is to evaluate the information needs for analyzing disposal solely of a subset of those wastes in a Defense Repository (DRep; i.e., those wastes that are either defense related, or managed by DOE but are not commercial in origin). A potential DRep also appears to be safe in the range of geologic mined repository concepts, but may have different concepts and features because of the very different inventory of waste that would be included. The focus of this status report is to cover the progress made in FY16 toward: (1) developing a preliminary DRep included inventory for engineering/design analyses; (2) assessing the major differences of this included inventory relative to that in other analyzed repository systems and the potential impacts to disposal concepts; (3) designing and developing an on-line waste library (OWL) to manage the information of all those wastes and their waste forms (including CSNF if needed); and (4) constraining post-closure waste form degradation performance for safety assessments of a DRep. In addition, some continuing work is reported on identifying potential candidate waste types/forms to be added to the full list from SNL (2014 – see Table C-1) which also may be added to the OWL in the future. The status for each of these aspects is reported herein.
An important issue for present and future generations is the final disposal of spent nuclear fuel. Over the past over forty years, the development of technologies to isolate both spent nuclear fuel (SNF) and other high-level nuclear waste (HLW) generated at nuclear power plants and from production of defense materials, and low- and intermediate-level nuclear waste (LILW) in underground rock and sediments has been found to be a challenging undertaking. Finding an appropriate solution for the disposal of nuclear waste is an important issue for protection of the environment and public health, and it is a prerequisite for the future of nuclear power. The purpose of a deep geological repository for nuclear waste is to provide to future generations, protection against any harmful release of radioactive material, even after the memory of the repository may have been lost, and regardless of the technical knowledge of future generations. The results of a wide variety of investigations on the development of technology for radioactive waste isolation from 19 countries were published in the First Worldwide Review in 1991 (Witherspoon, 1991). The results of investigations from 26 countries were published in the Second Worldwide Review in 1996 (Witherspoon, 1996). The results from 32 countries were summarized in the Third Worldwide Review in 2001 (Witherspoon and Bodvarsson, 2001). The last compilation had results from 24 countries assembled in the Fourth Worldwide Review (WWR) on radioactive waste isolation (Witherspoon and Bodvarsson, 2006). Since publication of the last report in 2006, radioactive waste disposal approaches have continued to evolve, and there have been major developments in a number of national geological disposal programs. Significant experience has been obtained both in preparing and reviewing cases for the operational and long-term safety of proposed and operating repositories. Disposal of radioactive waste is a complex issue, not only because of the nature of the waste, but also because of the detailed regulatory structure for dealing with radioactive waste, the variety of stakeholders involved, and (in some cases) the number of regulatory entities involved.
Sandia National Laboratories has begun research on the potential use of deep boreholes for the dis¬posal of radioactive waste. Characterization activities will focus on measurements and samples that are important for evaluating the long-term iso¬lation capability of the deep borehole disposal (DBD) concept. Engineering demonstration activities will focus on providing data to evaluate the concept’s operational safety and practicality. Procurement of a scientifically acceptable deep borehole field test (DBFT) site and a site management contractor is now under way.
The amount of brine present in domal salt formation is far less than in bedded salts (e.g., 0.01 to 0.1% compared with 1 to 3%). In salt domes, shear deformation associated with diapirism has caused existing brine to coalesce, leading to flow and expulsion. Brine migration behavior was investigated in bedded salt at WIPP (Nowak and McTigue 1987, SAND87-0880), and in domal salt at Asse (Coyle et al. 1987, BMI/ONWI-624). Test methods were not standardized, and the tests involved large diameter boreholes (17 to 36 in. diameter) and large apparatus. The tested intervals were proximal to mined openings (within approximately 1 diameter) where in situ stresses are redistributed due to excavation. The tests showed that (1) brine inflow rates can range over at least 2 orders of magnitude for domal vs. bedded salt, (2) that brine inflow is strongly associated with clay and interbedded permeable layers in bedded salt, and (3) that measurement systems can readily collect very small quantities of moisture over time frames of 2 years or longer. Brine inflow rates declined slightly with time in both test series, but neither series approached a state of apparent depletion. This range of flow magnitude could be significant to repository design and performance assessment, especially if inflow rates can be predicted using stratigraphic and geomechanical inputs, and can be shown to approach zero in a predictable manner.
Deep Borehole Disposal (DBD) of radioactive waste has some clear advantages over mined repositories, including incremental construction and loading, enhanced natural barriers provided by deep continental crystalline basement, and reduced site characterization. Unfavorable features for a DBD site include upward vertical fluid potential gradients, presence of economically exploitable natural resources, presence of high permeability connection from the waste disposal zone to the shallow subsurface, and significant probability of future volcanic activity. Site characterization activities would encompass geomechanical (i.e., rock stress state, fluid pressure, and faulting), geological (i.e., both overburden and bedrock lithology), hydrological (i.e., quantity of fluid, fluid convection properties, and solute transport mechanisms), chemical (i.e., rock and fluid interaction), and socioeconomic (i.e., likelihood for human intrusion) aspects. For a planned Deep Borehole Field Test (DBFT), site features and/or physical processes would be evaluated using both direct (i.e., sampling and in-hole testing) and indirect (i.e., surface and borehole geophysical) methods for efficient and effective characterization. Surface-based characterization would be used to guide the exploratory drilling program, once a candidate DBFT site has been selected. Borehole based characterization will be used to determine the variability of system state (i.e., stress, pressure, temperature, petrology, and water chemistry) with depth, and to develop material and system parameters relevant for numerical simulation. While the site design of DBD could involve an array of disposal boreholes, it may not be necessary to characterize each borehole in detail. Characterization strategies will be developed in the DBFT that establish disposal system safety sufficient for licensing a disposal array.
While deep borehole disposal of nuclear waste should rely primarily on off-the-shelf technologies pioneered by the oil and gas and geothermal industries, the development of new science and technology will remain important. Key knowledge gaps have been outlined in the research roadmap for deep boreholes (B. Arnold et al, 2012, Research, Development, and Demonstration Roadmap for Deep Borehole Disposal, Sandia National Laboratories, SAND2012-8527P) and in a recent Deep Borehole Science Needs Workshop. Characterizing deep crystalline basement, understanding the nature and role of deep fractures, more precisely age-dating deep groundwaters, and demonstrating long-term performance of seals are all important topics of interest. Overlapping deep borehole and enhanced geothermal technology needs include: quantification of seal material performance/failure, stress measurement beyond the borehole, advanced drilling and completion tools, and better subsurface sensors. A deep borehole demonstration has the potential to trigger more focused study of deep hydrology, high temperature brine-rock interaction, and thermomechanical behavior.
Options for disposal of the spent nuclear fuel and high level radioactive waste that are projected to exist in the United States in 2048 were studied. The options included four different disposal concepts: mined repositories in salt, clay/shale rocks, and crystalline rocks; and deep boreholes in crystalline rocks. Some of the results of this study are that all waste forms, with the exception of untreated sodium-bonded spent nuclear fuel, can be disposed of in any of the mined disposal concepts, although with varying degrees of confidence; salt allows for more flexibility in managing high-heat waste in mined repositories than other media; small waste forms are potentially attractive candidates for deep borehole disposal; and disposal of commercial SNF in existing dual-purpose canisters is potentially feasible but could pose significant challenges both in repository operations and in demonstrating confidence in long-term performance. Questions addressed by this study include: is a " 'one-size-fits-all ' repository a good strategic option for disposal?" and "do some disposal concepts perform significantly better with or without specific waste types or forms? " The study provides the bases for answering these questions by evaluating potential impacts of waste forms on the feasibility and performance of representative generic concepts for geologic disposal.
An analysis of the Waste Isolation Pilot Plant (WIPP) colloid model constraints and parameter values was performed. The focus of this work was primarily on intrinsic colloids, mineral fragment colloids, and humic substance colloids, with a lesser focus on microbial colloids. Comments by the US Environmental Protection Agency (EPA) concerning intrinsic Th(IV) colloids and Mg-Cl-OH mineral fragment colloids were addressed in detail, assumptions and data used to constrain colloid model calculations were evaluated, and inconsistencies between data and model parameter values were identified. This work resulted in a list of specific conclusions regarding model integrity, model conservatism, and opportunities for improvement related to each of the four colloid types included in the WIPP performance assessment.
Published results of performance assessments for deep geologic disposal of high-level radioactive waste and spent nuclear fuel provide insight into those aspects of the waste form that are potentially important to the long-term performance of a repository system. Alternative waste forms, such as might result from new technologies for processing spent fuel and advances in nuclear reactor design, have the potential to affect the long-term performance of a geologic repository. This paper reviews relevant results of existing performance assessments for a range of disposal concepts and provides observations about how hypothetical modifications to waste characteristics (e.g., changes in radionuclide inventory, thermal loading, and durability of waste forms) might impact results of the performance assessment models. Disposal concepts considered include geologic repositories in both saturated and unsaturated environments. Specifically, we consider four recent performance assessments as representative of a range of disposal concepts. We examine the extent to which results of these performance assessments are affected by (i) thermal loading of the waste proposed for disposal; (ii) mechanical and chemical lifetime of the waste form; and (iii) radionuclide content of the waste. We find that peak subsurface temperature generally is a constraint that can be met through engineering solutions and that processing of wastes to reduce thermal power may enable more efficient use of repositories rather than improved repository performance. We observe that the rate of radionuclide release is often limited by geologic or chemical processes other than waste form degradation. Thus, the effects on repository performance of extending waste-form lifetime may be relatively small unless the waste form lifetime becomes sufficiently long relative to the period of repository performance. Finally, we find that changes to radionuclide content of waste (e.g., by separation or transmutation processes) do not in general correspond to proportional effects on repository performance. Rather, the effect of changes to radionuclide content depends on the relative mobility of various radionuclides through the repository system, and consequently on repository geology and geochemistry.
This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.
This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within a few meters of the repository, that can be reasonably characterized. Within a few centuries after waste emplacement, overburden pressures will seal fractures, resaturate the dehydrated zones, and provide a repository setting that strongly limits radionuclide movement to diffusive transport. Coupled hydrogeochemical transport calculations indicate maximum extents of radionuclide transport on the order of tens to hundreds of meters, or less, in a million years. Under the conditions modeled, a shale repository could achieve total containment, with no releases to the environment in undisturbed scenarios. The performance analyses described here are based on the assumption that long-term standards for disposal in clay/shale would be identical in the key aspects, to those prescribed for existing repository programs such as Yucca Mountain. This generic repository evaluation for shale is the first developed in the United States. Previous repository considerations have emphasized salt formations and volcanic rock formations. Much of the experience gained from U.S. repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, is applied here to scoping analyses for a shale repository. A contemporary understanding of clay mineralogy and attendant chemical environments has allowed identification of the appropriate features, events, and processes to be incorporated into the analysis. Advanced multi-physics modeling provides key support for understanding the effects from coupled processes. The results of the assessment show that shale formations provide a technically advanced, scientifically sound disposal option for the U.S.