Publications

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Commercial generation of energy via nuclear power plants in the United States (U.S.) has generated thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE) (Nuclear Waste Policy Act of 1982). Any repository licensed to dispose of the SNF must meet requirements regarding the long-term performance of the repository. In evaluating the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store the SNF but were not designed for permanent disposal. As part of a multiyear plan that is currently being developed for the DOE, a two-phase study has been initiated to examine the potential consequences, with respect to long-term repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consists of generating an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or include criticality in a performance assessment (PA) as appropriate. The Phase I approach will be used to guide the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the approach created during Phase I. The study discussed herein focuses on the consequences of criticality in a DPC; it does not address the probability of occurrence of a criticality event. This approach examines two types of criticality events for SNF disposed of in a single type of DPC: a steady-state criticality and a transient criticality. The steady-state critical event is characterized by a relatively low constant power output over 10,000 years, while the transient critical event is characterized by a power spike that lasts on the order of seconds. Possible effects of the criticality are an increase in the radionuclide inventory; an increase in temperature; and a change in the chemistry inside the waste package, along with a change in radionuclide solubilities, fuel degradation rates, and steel corrosion rates. Additionally, for transient criticality the possibility of mechanical damage to the engineered and natural barriers also exists.

The most straightforward concept for disposal of large, heavy packages containing commercial spent nuclear fuel (CSNF) in a repository in bedded salt, would be to emplace them directly on the floor in emplacement tunnels. In-tunnel axially aligned horizontal emplacement would minimize excavated volume and avoid drilling of large-diameter emplacement boreholes. A similar concept was proposed in Germany for direct disposal of POLLUX^® canisters. The repository would be constructed at a depth of 500 to 1,000 m for isolation from the surface, and for sufficient overburden stress to ensure creep reconsolidation of repository openings. It could entail modular panels of emplacement tunnels arranged on headings oriented in cardinal directions from a central core, to accommodate the estimated 140,000 MTU total U.S. CSNF inventory. To do so, the overall area of the repository layout would be approximately 20 km². Many layouts are possible, but the approach should be modular, excavation should be deferred for as long as possible to avoid maintenance, and the emplacement areas should share support facilities and shafts. Vertical shafts would be used in accordance with mining practice in sedimentary basins. Large diameter shafts would be needed for ventilation exhaust and waste transport, with smaller shafts for waste salt removal, men & materials, and ventilation intake. The spacing between disposal tunnels as estimated from thermal modeling, seeks to limit the maximum average areal thermal load in the panels to 11 W/m² to control long-term heat buildup in the host rock. Peak salt temperature would occur within a few years and would be dominated by each waste package locally, simplifying thermal management. There would be some flexibility to decrease the package spacing or increase the emplacement thermal power limit Backfilling emplaced waste packages immediately with mine-run crushed salt would provide shielding and expedite reconsolidation. This arrangement would isolate adjacent waste packages from one another by the intervening backfill, especially after it reconsolidates and its properties approach those of intact salt. After the repository is fully loaded and the performance confirmation program is complete, activities to permanently close the repository would be initiated. During closure operations all openings in the host salt would be backfilled, then shafts would be sealed, and boreholes plugged. Plans for the Waste Isolation Pilot Plant (WIPP) show how sealing and plugging could be done. A monitoring program could continue for 50 years or longer after repository closure. With an emplacement thermal power limit of 10 kW per waste package, nearly all the CSNF that is projected to be produced by the current fleet of reactors in the U.S. could be emplaced over a period of approximately 50 years starting in calendar 2048. No barriers to implementation in a reasonable timeframe have been identified from this generic analysis. Engineering challenges include: 1) shaft construction; 2) a very-large capacity shaft hoist; 3) overpack design; 4) a transport-emplacement-vehicle (TEV) for transporting waste packages once they are underground and emplacing them remotely; and 5) remotely operated equipment for emplacing backfill. The method of shaft construction would depend on site-specific conditions, and could involve freezing the subsurface. A shaft hoist with payload capacity of 175 MT seems technically and economically feasible based on development work in Germany, and it would be the largest hoist of its kind. The function of disposal overpacks would be to provide reliable containment during repository operations, which could be accomplished using a corrosion allowance material such as a low-carbon steel. Development effort would be needed to determine overpack thickness (e.g., 7 to 20 cm) that can resist corrosion and loading from salt creep, to rovide containment throughout the repository operational period. The transport-emplacement vehicle (TEV) would be similar to previous concepts, particularly one option proposed for a Yucca Mountain repository. It would move over a rough salt floor on independently driven and steered wheels, and carry heavy shielding in addition to a waste package. By analogy to the safety case for the WIPP in New Mexico, human intrusion is likely to be the dominant mode of radionuclide release from the repository. Treatment of human intrusion for a CSNF repository in salt could depend on promulgation of site-specific changes in the regulations. Radionuclide release and migration would be quite limited for undisturbed conditions. There may be opportunities for improved understanding of salt performance with waste heating, based on future in situ testing in an underground salt research laboratory. This report also discusses a developing area of salt rock mechanics that involves low-stress, low strain-rate creep that might cause large, heavy waste packages to slowly sink. Site-specific sampling, testing, and modeling would be used to determine if the mechanism is important enough to merit consideration in design, or inclusion in performance assessment. Part of engineering design and postclosure safety assessment for a CSNF repository in salt would be to implement a methodology to show that the probability of a criticality event in the repository, when waste packages eventually breach and are flooded, is less than the probability screening threshold for performance assessment. In the methodology, a criticality analysis would be performed for waste packages in the repository, incorporating measures that could be introduced as needed to limit reactivity, for example using fuel selection and loading rules, and crediting the absorption of thermal neutrons by natural chlorine in the environment. A similar analysis has been underway for CSNF stored in dual-purpose canisters. Ideally the strategy would be developed prior to actually loading SNF assemblies into canisters used in waste packages for disposal.

This report presents a generic (i.e., site-independent) preliminary plan for drilling, testing, sampling, and analyzing data for a deep characterization borehole drilled into crystalline basement for the purposes of assessing the suitability of a site for deep borehole disposal (DBD). 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. This report presents a possible sampling, testing, and analysis campaign that could be carried out as part of a future project to quantify geochemical, geomechanical, geothermal, and geohydrologic conditions encountered at depths up to 5 km in crystalline basement.

This report presents a preclosure radiological safety assessment for deep borehole disposal (DBD) of nuclear wastes. The primary purpose of the safety assessment is to identify risk factors for disposal operations, to aid in design for an engineering demonstration of technology for DBD. The assessment is based on a conceptual design for disposal packages and borehole systems that was developed previously. It considers operational steps that could be used for actual DBD, with internal and external initiating off-normal events, to develop insights that can be applied to an engineering demonstration that would be performed without using any form of nuclear waste. 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

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The overall DOE R&D strategy for DPC disposition includes a significant effort directed toward consequence screening to determine if engineered solutions discussed above are needed. Work to develop injectable filler technology will continue. The disposal criticality control features approach, and zone loading, have not been investigated since the EPRI studies in 2008-2009. The utility of such measures would be maximized by implementing them soon. This ongoing study is motivated by comparative cost analysis [7] that showed the potential cost savings using the control rods/blades approach, compared to repackaging (comparing the two most technically mature options for DPC disposition and retaining the low-probability criticality screening objective) would be approximately $2 million per DPC.

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This report is the deliverable M2SF-18SN010305026 FY18 Summary Update on the Feasibility of Direct Disposal of SNF in Existing DPCs. It reports on work done throughout fiscal year (FY) 2018, on work planned at the beginning of that FY, consisting of R&D activities for: 1) injectable fillers that could be used in dual-purpose canisters to prevent postclosure criticality in a geologic repository, and 2) as-loaded DPC data gathering and criticality. The work reported here was performed by Sandia National Laboratories and Oak Ridge National Laboratory. Appropriate attribution to source documents is provided in the text, tables, and figures below. Additional R&D on direct disposal of existing DPCs was planned and funded in mid-FY, and the associated reporting is separate from this milestone. Additional discussion of that new scope and how it implements findings from an independent expert review of the fillers R&D program (Section 10) is provided in the Summary (Section 11).

This memo describes the engineering technical and costing analysis support needed for identifying and evaluating technical and programmatic solutions for spent nuclear fuel (SNF) in dual-purpose canisters (DPCs), and the resources planned to provide that support. The Technical and Programmatic Solutions (T&PS) work scope is intended to identify and evaluate the range of feasible options available for DPC direct disposal, considering the range of DPC designs in the existing fleet and a range of generic geologic disposal concepts. It will also identify changes to the way DPCs are loaded, and/or additional hardware that could be installed in DPCs as they are loaded, to improve disposability (chiefly, post closure criticality control). These two thrusts are the focus of engineering support to the work package.

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This is a work planning document that describes technical and programmatic goals for disposition of spent nuclear fuel (SNF) that is currently in dry storage in dual-purpose canisters (DPCs), or will be in the foreseeable future. It then describes how those goals can be promoted by a research and development (R&D) program. The needed R&D is compared to the ongoing work supported by the U.S. Department of Energy in FY18, and planned for FY19 and beyond. Some additional R&D activities are recommended, and plans are presented for technical integration activities that address the efficacy of the Direct Disposal of DPCs program (WBS 1.08.01.03.05), and integration with the overall Disposal Research program (WBS 1.08.01.03). The planned deliverable for this work package in FY19 (M2SF-1951\1010305051-Analysis of Solutions for DPC Disposal; 6/19/19) will be the product of this workplan. The deliverable will evaluate technical options for DPC direct disposal, taking into account the range ofpast and current DPC designs in the existing fleet. It will describe a set of goals for successful disposition of spent fuel in DPCs. It will analyze the scope and timing of needed R&D activities (R&D Plan), and discuss the uses of generic and site-specific analyses. Where appropriate, it will use alternative management cases to represent how DPC direct disposal could be incorporated in the overall geologic disposal program, given uncertainties in program direction and funding.

Objectives for Multi-Physics Simulations include: Provide a systematic framework for multi-process modeling — Conduct parallel model development efforts that cover the technical areas needed to support criticality consequence screening in performance assessment (PA) and that will be more closely integrated as development proceeds; Investigate separate effects — Allow partitioning of the overall waste package (WP) internal criticality multi-physics modeling effort during development activities, for study of specific processes that can later be coupled if warranted from interpretation of results; Study scaling and bounding approaches — Where possible, represent criticality consequences in PA using simplification of uncertain criticality event frequency and magnitude, bounding of consequences for screening purposes, and scaling of consequences to multiple WPs; and, Integration among participants — Multiple modeling teams (mainly SNL and ORNL, and their collaborators) will work on different parts of the in-package criticality phenomenology. Insights generated this way will be combined for more realistic coupled modeling, and for validation.

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There are currently 2,462 dual-purpose canisters (DPCs) containing spent nuclear fuel (SNF) across the United States. Repackaging DPCs into specialized disposal canisters could be financially and operationally costly with additional radiological, operational safety, and management risks. There are several approaches to facilitate direct disposal of DPCs and demonstrate acceptable repository performance. A promising approach is to fill the void space within the DPCs with a material that would significantly limit the potential for criticality through limiting moderation and/or the addition of neutron absorbers in the interstitial spaces within the fuel assemblies and baskets. An acceptable filler would demonstrably show that the probability of criticality in DPCs during the disposal period of interest to be below the probability threshold for inclusion in repository performance assessment. Based on previous work conducted by domestic and international organizations, two approaches were identified as potentially viable for introduction of fillers into DPCs as liquids that would eventually solidify: (1) molten metal fillers introduced at higher temperatures, and (2) resins or cement slurries that solidify at lower temperatures.

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