Publications

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At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

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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.

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This report is one follow-on to a study of reference geologic disposal design concepts (Hardin et al. 2011a). Based on an analysis of maximum temperatures, that study concluded that certain disposal concepts would require extended decay storage prior to emplacement, or the use of small waste packages, or both. The study used nominal values for thermal properties of host geologic media and engineered materials, demonstrating the need for uncertainty analysis to support the conclusions. This report is a first step that identifies the input parameters of the maximum temperature calculation, surveys published data on measured values, uses an analytical approach to determine which parameters are most important, and performs an example sensitivity analysis. Using results from this first step, temperature calculations planned for FY12 can focus on only the important parameters, and can use the uncertainty ranges reported here. The survey of published information on thermal properties of geologic media and engineered materials, is intended to be sufficient for use in generic calculations to evaluate the feasibility of reference disposal concepts. A full compendium of literature data is beyond the scope of this report. The term “uncertainty” is used here to represent both measurement uncertainty and spatial variability, or variability across host geologic units. For the most important parameters (e.g., buffer thermal conductivity) the extent of literature data surveyed samples these different forms of uncertainty and variability. Finally, this report is intended to be one chapter or section of a larger FY12 deliverable summarizing all the work on design concepts and thermal load management for geologic disposal (M3FT-12SN0804032, due 15Aug2012).

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This memo describes rough-order-of-magnitude (ROM) cost estimates for a set of off-normal (accident) scenarios, as defined for two waste package emplacement method options for deep borehole disposal: drill-string and wireline. It summarizes the different scenarios and the assumptions made for each, with respect to fishing, decontamination, remediation, etc.

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This report presents four concepts for packaging of radioactive waste for disposal in deep boreholes. Two of these are reference-size packages (11 inch outer diameter) and two are smaller (5 inch) for disposal of Cs/Sr capsules. All four have an assumed length of approximately 18.5 feet, which allows the internal length of the waste volume to be 16.4 feet. However, package length and volume can be scaled by changing the length of the middle, tubular section. The materials proposed for use are low-alloy steels, commonly used in the oil-and-gas industry. Threaded connections between packages, and internal threads used to seal the waste cavity, are common oilfield types. Two types of fill ports are proposed: flask-type and internal-flush. All four package design concepts would withstand hydrostatic pressure of 9,600 psi, with factor safety 2.0. The combined loading condition includes axial tension and compression from the weight of a string or stack of packages in the disposal borehole, either during lower and emplacement of a string, or after stacking of multiple packages emplaced singly. Combined loading also includes bending that may occur during emplacement, particularly for a string of packages threaded together. Flask-type packages would be fabricated and heat-treated, if necessary, before loading waste. The fill port would be narrower than the waste cavity inner diameter, so the flask type is suitable for directly loading bulk granular waste, or loading slim waste canisters (e.g., containing Cs/Sr capsules) that fit through the port. The fill port would be sealed with a tapered, threaded plug, with a welded cover plate (welded after loading). Threaded connections between packages and between packages and a drill string, would be standard drill pipe threads. The internal flush packaging concepts would use semi-flush oilfield tubing, which is internally flush but has a slight external upset at the joints. This type of tubing can be obtained with premium, low-profile threaded connections at each end. The internal-flush design would be suitable for loading waste that arrives from the originating site in weld-sealed, cylindrical canisters. Internal, tapered plugs with sealing filet welds would seal the tubing at each end. The taper would be precisely machined onto both the tubing and the plug, producing a metal-metal sealing surface that is compressed as the package is subjected to hydrostatic pressure. The lower plug would be welded in place before loading, while the upper plug would be placed and welded after loading. Conceptual Waste Packaging Options for Deep Borehole Disposal July 30, 2015 iv Threaded connections between packages would allow emplacement singly or in strings screwed together at the disposal site. For emplacement on a drill string the drill pipe would be connected directly into the top package of a string (using an adapter sub to mate with premium semi-flush tubing threads). Alternatively, for wireline emplacement the same package designs could be emplaced singly using a sub with wireline latch, on the upper end. Threaded connections on the bottom of the lowermost package would allow attachment of a crush box, instrumentation, etc.

This document presents design requirements and controlled assumptions intended for use in the engineering development and testing of: 1) prototype packages for radioactive waste disposal in deep boreholes; 2) a waste package surface handling system; and 3) a subsurface system for emplacing and retrieving packages in deep boreholes. Engineering development and testing is being performed as part of the Deep Borehole Field Test (DBFT; SNL 2014a). This document presents parallel sets of requirements for a waste disposal system and for the DBFT, showing the close relationship. In addition to design, it will also inform planning for drilling, construction, and scientific characterization activities for the DBFT. The information presented here follows typical preparations for engineering design. It includes functional and operating requirements for handling and emplacement/retrieval equipment, waste package design and emplacement requirements, borehole construction requirements, sealing requirements, and performance criteria. Assumptions are included where they could impact engineering design. Design solutions are avoided in the requirements discussion. Deep Borehole Field Test Requirements and Controlled Assumptions July 21, 2015 iv ACKNOWLEDGEMENTS This set of requirements and assumptions has benefited greatly from reviews by Gordon Appel, Geoff Freeze, Kris Kuhlman, Bob MacKinnon, Steve Pye, David Sassani, Dave Sevougian, and Jiann Su.

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This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design July 23, 2015 iv ACKNOWLEDGEMENTS This report has benefited greatly from review principally by Steve Pye, and also by Paul Eslinger, Dave Sevougian and Jiann Su.

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This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

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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.

Commercial spent nuclear fuel (SNF) continues to accumulate in dry storage, sealed into welded dual-purpose canisters (DPCs). Direct disposal of DPCs, without cutting them open and re-packaging the fuel, is technically feasible at least for some DPCs and some disposal concepts. Options for DPC direct disposal are taking form, based on an ongoing study by the U.S. Department of Energy. Direct disposal of DPCs should be viewed as one part of a diverse fuel management system that will eventually switch to loading standardized multi-purpose canisters (MPCs). Nearly all DPCs that are loaded before this switch could be directly disposed depending on the disposal environment selected. DPC direct disposal options have been developed for salt, crystalline and sedimentary host media. These options are suited to different populations of DPCs, ranging from those containing older, colder fuel (e.g., in sedimentary media) to all DPCs (salt). The timing of DPC use offers an opportunity to simplify the SNF management system. Commercial SNF will be generated in the U.S. for more than 90 years, whereas facility lifetimes are typically on the order of 50 years. Efficiencies could be realized by implementing disposal in "campaigns. " Additional accumulation of DPCs over the next 10 to 20 years, followed by a transition to MPCs, would define two such campaigns. A repository could first be constructed for MPCs, and disposal of DPCs could be deferred and addressed later using new, dedicated facilities. During the interim storage period DPC thermal output would decay, further expanding disposal options.

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After an exposition of the materials used in DPCs and the factors controlling material corrosion in disposal environments, a survey is given of the corrosion rates, mechanisms, and products for commonly used stainless steels. Research needs are then identified for predicting stability of DPC materials in disposal environments. Stainless steel corrosion rates may be low enough to sustain DPC basket structural integrity for performance periods of as long as 10,000 years, especially in reducing conditions. Uncertainties include basket component design, disposal environment conditions, and the in-package chemical environment including any localized effects from radiolysis. Prospective disposal overpack materials exist for most disposal environments, including both corrosion allowance and corrosion resistant materials. Whereas the behavior of corrosion allowance materials is understood for a wide range of corrosion environments, demonstrating corrosion resistance could be more technically challenging and require environment-specific testing. A preliminary screening of the existing inventory of DPCs and other types of canisters is described, according to the type of closure, whether they can be readily transported, and what types of materials are used in basket construction.

This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to establish corrosion rates and component lifetimes. Finally, it is unlikely that the aluminum-based neutron absorber materials that are commonly used in existing DPCs would survive for 10,000 years in disposal environments, because the aluminum will act as a sacrificial anode for the steel. We recommend additional testing of borated and Gd-bearing stainless steels, to establish general and localized corrosion resistance in repository-relevant environmental conditions.

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