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At least 703 records · Page 39

The Use of Underground Research Laboratories to Support Repository Development Programs. A Roadmap for the Underground Research Facilities Network.

Under the auspices of the International Atomic Energy Agency (IAEA), nationally developed underground research laboratories (URLs) and associated research institutions are being offered for use by other nations. These facilities form an Underground Research Facilities (URF) Network for training in and demonstration of waste disposal technologies and the sharing of knowledge and experience related to geologic repository development, research, and engineering. In order to achieve its objectives, the URF Network regularly sponsors workshops and training events related to the knowledge base that is transferable between existing URL programs and to nations with an interest in developing a new URL. This report describes the role of URLs in the context of a general timeline for repository development. This description includes identification of key phases and activities that contribute to repository development as a repository program evolves from an early research and development phase to later phases such as construction, operations, and closure. This information is cast in the form of a matrix with the entries in this matrix forming the basis of the URF Network roadmap that will be used to identify and plan future workshops and training events.

MacKinnon, Robert J.↗

RANGERS: State of the Art and Science on Engineered Barrier Systems in Salt Formations

The construction of deep geological repositories (DGR) in salt formations requires penetrating through naturally sealing geosphere layers. While the emplaced nuclear waste is primarily protected by the containment-providing rock zone (CRZ), technical barriers are required, for example during handling. For closure geotechnical barriers seal the repository along the accesses against water or solutions from outside and the possible emission paths for radionuclides contained inside. As these barriers must ensure maintenance-free function on a long-term basis, they typically comprise a set of specialized elements with diversified functions that may be used redundantly. The effects of the individual elements are coordinated so that they are collectively referred to as the Engineered Barrier System (EBS).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Next Generation System Analysis Model Recently Added Features and Future Plans - Abstract

The Nuclear Waste Policy Act of 1982, as amended (NWPA 1982), established the federal government’s responsibility to accept spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from waste owners and generators for ultimate disposition. SNF generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWM system are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWM system. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. Recent major enhancements that have been implemented into NGSAM since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort and buffer car acquisition. • Addition of heavy haul and barge routes for some sites, as well as support for user-defined inter-modal routes. • Updates to the logic that checks the thermal maps prior to package transport. • Addition of an allocation method that predicts when reactor sites will pack assemblies from their pools for dry storage and allocates packages to those reactor sites in the preceding periods, favoring direct transport packages and reducing the number of packages that reactor sites pack for dry storage at their ISFSIs. • Addition of reactor site family operational limits, which are used to limit the number of loads from the pool and from dry storage at a given reactor site per year. • Support has been added for multiple canister loading maps and packages having multiple compatible transportation overpacks. • Updates in the handling of non-commercial fuel, including a new database containing data to support the updates. • Support for repackaging at reactor sites. • Implementing additional output reports or modifying existing reports. • User edits can now be created and edited via the NGSAM website. • Ability to load packages for dry storage at ISF pools. • Same-type package blending at DOE sites. • Support for multi-mode transloading at reactor sites. These new features have improved NGSAM capabilities and/or improve the user experience with the model and will be discussed in more detail. The initial NGSAM requirements for advanced reactor fuels, reprocessing, treatment, and conditioning are preliminary and are described at a high level in this paper: analysts will provide more specific requirements to the NGSAM team in the future. Additionally, there are many data needs associated with modeling advanced reactors in NGSAM, but many of the data or plans are still in progress and/or yet to be fully defined. However, this document describes an initial exploration of the data relevant to this program. Advanced reactor data will likely require revision as concepts evolve and new considerations are made. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by DOE. No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Experiments and Modeling in Support of Generic Salt Repository Science

Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Ion-Exchange Interdiffusion Model with Potential Application to Long-Term Nuclear Waste Glass Performance

Ion exchange is an integral mechanism influencing the corrosion of glasses. Due to the formation of alteration layers in aqueous conditions, it is difficult to conclusively deconvolute the process of ion exchange from other processes, principally dissolution of the glass matrix. Therefore, we have developed a method to isolate alkali diffusion that involves contacting glass coupons with a solution of 6 LiCl dissolved in functionally inert dimethyl sulfoxide. Here, we employ the method at temperatures ranging from 25 to 150 °C with various glass formulations. Glass compositions include simulant nuclear waste glasses, such as SON68 and the international simple glass (ISG), glasses in which the nature of the alkali element was varied, and glasses that contained more than one alkali element. An interdiffusion model based on Fick’s second law was developed and applied to all experiments to extract diffusion coefficients. The model expands established models of interdiffusion to the case where multiple types of alkali sites are present in the glass. Activation energies for alkali ion exchange were calculated and the results are in agreement with those obtained in glass strengthening experiments but are nearly five times higher than values reported for diffusion-controlled processes in nuclear waste glass corrosion experiments. A discussion of the root causes for this apparent discrepancy is provided. The interdiffusion model derived from laboratory experiments is expected to be useful for modeling glass corrosion in a geological repository when the silicon concentration is high.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Time-Temperature-Transformation (TTT) Diagram for a Sludge Batch 9 Glass Composition Based on Coupled-Operation with the Salt Waste Processing Facility

The amorphous structure of a glass waste form has the potential to rearrange into crystalline phases at temperatures between the liquidus temperature and the glass transition temperature (Tg). Certain phases that can form will be detrimental to the durability of the glass and it is important to know the conditions that promote devitrification. The canister-centerline-cooling (CCC) profile is used to replicate the area within the center of the Defense Waste Processing Facility (DWPF) canister with the slowest cooling during the initial cool down after pouring, which has the greatest potential for crystallization. Other time-temperature conditions that cause significant changes in either phase structure or phase composition are identified by a time-temperature-transformation (TTT) study. The phase stability of a waste form must be determined as a part of the Waste Acceptance Product Specifications (WAPS) if it is to eventually be stored in a geologic repository. This requires the creation of a TTT diagram and analysis of the Tg, as defined by the Department of Energy (DOE). The previous TTT study for a DWPF glass waste form was completed in 2010 prior to coupled operation with the Salt Waste Processing Facility (SWPF). SWPF transfers two high activity waste streams to DWPF for vitrification: a cesium-containing strip effluent and a stream containing monosodium titanate/sludge solids. These SWPF streams were first transferred to DWPF for vitrification during Sludge Batch 9 (SB9) in 2021. The impact of these SWPF streams on crystallization behavior was not determined in previous studies and the need for data to satisfy WAPS Specification 1.4 for SB9 was identified.

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Corrosion Kinetics of Stainless Steel by Hydrogen Measurement under Deep Geological Repository Condition - 16047

Spent fuel assemblies consisting of stainless steels are expected to be disposed of in a deep underground repository. The radionuclides release and gas generation from activated metallic wastes in a disposal system will be controlled by the corrosion reaction. Most studies on stainless steel corrosion have been investigated under corrosion rates around several μm.a{sup -1}. A recent study has revealed that the corrosion rate of stainless steel is extremely slow-less than nm.a{sup -1} at 303 K. In the present study, corrosion kinetics at different temperatures up to 353 K were evaluated from the oxidation reaction by hydrogen gas measurement using 18Cr-8Ni austenitic stainless steel in a deoxygenated glass ampoule filled with a dilute NaOH solution of pH 12.5 for 24 months. Uniform corrosion and a reaction of 3Fe + 4H{sub 2}O → Fe{sub 3}O{sub 4} + 4H{sub 2} were assumed. After 24 months, the corrosion rates, which decreased with time but increased with rising temperature, were on average 7.80 x 10{sup -4}, 2.65 x 10{sup -3}, and 7.62 x 10{sup -3} μm.a{sup -1} at 303 K, 323 K, and 353 K, respectively. The corrosion kinetics were found to follow a parabolic rate law, which suggests the corrosion process is under diffusion control. The activation energy evaluated by the Arrhenius relationship of the parabolic rate constants has been determined to be 82.5 ± 6.2 kJ/mol, which corresponds to the values at high temperatures down to 423 K. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Rapid methods for radionuclide contaminant transport in nuclear fuel cycle simulation

Here, nuclear fuel cycle and nuclear waste disposal decisions are technologically coupled. However, current nuclear fuel cycle simulators lack dynamic repository performance analysis due to the computational burden of high-fidelity hydrolgic contaminant transport models. The Cyder disposal environment and repository module was developed to fill this gap. It implements medium-fidelity hydrologic radionuclide transport models to support assessment appropriate for fuel cycle simulation in the Cyclus fuel cycle simulator. Rapid modeling of hundreds of discrete waste packages in a geologic environment is enabled within this module by a suite of four closed form models for advective, dispersive, coupled, and idealized con- taminant transport: a Degradation Rate model, a Mixed Cell model, a Lumped Parameter model, and a 1-D Permeable Porous Medium model. A summary of the Cyder module, its timestepping algorithm, and the mathematical models implemented within it are presented. Additionally, parametric demonstrations simulations performed with Cyder are presented and shown to demonstrate functional agreement with parametric simulations conducted in a standalone hydrologic transport model, the Clay Generic Disposal System Model developed by the Used Fuel Disposition Campaign Department of Energy Office of Nuclear Energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Monte Carlo characterization of the cosmic ray muon flux in shallow subsurface geological repositories intended for disposal of radioactive materials

Recent challenges in monitoring subsurface geological repositories intended for disposal of radioactive materials such as spent nuclear fuel call for new, innovative concepts that are facility independent, cost-effective, passive, and reliable. Once nuclear material is in place at these facilities, reverifying the inventory may no longer be feasible if continuity of knowledge is lost or unavailable to the inspectors. Using cosmic ray muons may present several potential advantages over conventional photon/neutron signatures, and their use in safeguards applications have only received attention in the past decade. However, there have been limited efforts to explore the integration of cosmic ray muons into repository safeguards and study potential gains, risks, and costs. This paper presents a Monte Carlo-based methodology to characterize the cosmic ray muon flux, including muon angular and energy differential distributions at depths representative of subsurface geological repositories. Since there have been limited measurements at these sites and a measurement made in one site is not always transferable to another site, the objective is to develop an efficient simulation method and useful parametrizations to provide a convenient tool for enabling muon simulations at any geological repository site. It is expected these results will provide a better understanding of how muons can be integrated into an existing geological repository safeguards framework.

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Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project team is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.

Cumberland, Riley M↗

Simulations of Water Flow in Relation to a Steady Criticality in an Unsaturated Alluvial Repository

Oak Ridge National Laboratory (ORNL) is examining the conditions under which a criticality event could be ruled out for large packages placed in an unsaturated alluvial repository. This work specifically examines the water level in a stylized waste package that is assumed (1) to fail on the top, and (2) to empty as water is driven away—not by boiling—but by fluid flow phenomena driven by thermal gradients. This analysis uses the PFLOTRAN code package. Specifically, this year’s work includes examination of the heat rate at which water is driven away for various percolation rates using the same assumptions listed in the 2022 annual report. The STANDARDS (formerly UNF-ST&DARDS) package has predicted dual-purpose canister (DPC) neutron multiplication factors (k-effectives) ranging above 1.1 (unitless) with failed components. The DPC designs were not created with the intent to use for disposal purposes. Emplacing such packages may be worth consideration if it can be shown that criticality is either extremely unlikely or of very low consequence to the repository’s safety case. This work aims to help establish hydrologic boundary conditions for a highly hypothetical, highly stylized subsurface criticality event. The approach to critical is assumed to be slow, producing steady heating rates under a few kilowatts. For a DPC containing capillary media, the simulated power required for dry-out was proportional to the percolation rate but was unaffected by the size of a large breech on top of the DPC. Results for single cases with no capillary media in the DPC suggest that higher infiltration rates may be required in those cases to cause flooding. Bounding mechanisms and assumptions must be identified and characterized further by performing additional perturbations to various input parameters to demonstrate robustness of the results. This might be accomplished by examining responses to power pulses to determine the system transfer function required to enable a reduced order feedback model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Molten Salt Oxidation of Ion Exchange and Oils in Carbonate Salts and Evaporator Residue - 16531

Molten Salt Oxidation is a well-studied and robust technology. Recently, renewed interest in the technology has been seen due to limitations on the use of incineration for hazardous and radioactive waste streams. The nuclear research center at Rez, near Prague, has been investigating a bench scale molten salt oxidation system to provide insight for the design of a pilot scale system to condition spent ion exchange resin and scintillator oils. The use of both an eutectic K/Na/Li carbonate salt and a simulated residue salt for liquid waste obtained from evaporated primary coolant of nuclear reactors was examined. In these studies the usable capacity of the salts were examined. Materials doped with surrogate radioisotopes were processed and the capture efficiency was measured. Corrosion of the reactor vessel is of significant engineering concern. Corrosion coupon tests were performed for Inconel HX and stainless steel. These tests provided corrosion rates to assist in design of pilot scale systems. While the bench scale reactor's detention time was too short to provide complete combustion, the values of emissions were tracked and showed an increase in the combustion efficiency at temperatures greater than 750 deg. C. A two part reactor has been designed and work is planned to continue to optimize the system. Additionally, further work has examined the final conditioning of salt which has been saturated with radionuclides after processing organic waste streams A conceptual pilot plant has been designed to solidify radionuclide saturated salt into a form acceptable for repositories according to EU directives. Saturated salt is milled to a powder and immobilized in a matrix with a high percentage of waste loading. Both geo-polymer and polysiloxane matrices have been methods for low temperature immobilization. Conditioning in this manner decreases operating and capital costs when compared to current conditioning technologies and procedures utilized for high volume reduction factors of final waste forms. (authors)

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Review of Inputs Provided to Jason Associates Corporation in Support of RWEV-REP-001, the Analysis of Postclosure Groundwater Impacts Report

Report RWEV-REP-001, Analysis of Postclosure Groundwater Impacts for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High Level Radioactive Waste at Yucca Mountain, Nye County, Nevada was issued by the DOE in 2009 and is currently being updated. Sandia National Laboratories (SNL) provided support for the original document, performing calculations and extracting data from the Yucca Mountain Performance Assessment Model that were used as inputs to the contaminant transport and dose calculations by Jason Associates Corporation, the primary developers of the DOE report. The inputs from SNL were documented in LSA-AR-037, Inputs to Jason Associates Corporation in Support of the Postclosure Repository Supplemental Environmental Impact Statement. To support the updating of the original Groundwater Impacts document, SNL has reviewed the inputs provided in LSA-AR-037 to verify that they are current and appropriate for use. The results of that assessment are documented here.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Effect of solution chemistry on the iodine release from iodoapatite in aqueous environments

To ensure the safe disposal of nuclear waste, understanding the release process of radionuclides retained in the nuclear waste forms is of vital importance. Iodoapatite Pb 9.85 (VO 4 ) 6 I 1.7 , a potential waste form for iodine-129, was selected as a model system for ceramic waste forms in this study to understand the effect of aqueous species on iodine release. Semi-dynamic leaching tests were conducted on bulk samples in cap-sealed Teflon vessels with 0.1 mol/L NaCl, Na 2 CO 3 , Na 3 PO 4 , and Na 2 SO 4 solutions under 90 °C, fixed sample surface area to solution volume ratio of 5/m, and periodic replacement of leaching solutions. The reacted solutions were then analyzed by Inductively Coupled Plasma-Mass Spectrometry and Inductively Coupled Plasma-Optical Emission Spectrometry; the leached surfaces were characterized by X-ray diffraction, scanning electron microscopy, and infrared spectroscopy. The finding shows that, compared to deionized water, the ion-rich solutions enhanced the iodine release as a result of the increased ionic strength, reduced activity coefficient of dissolved species, and increased solution pH. Surface reactions can lead to the formations of secondary phases by ion-exchange and precipitation. These findings indicate that an ion-rich environment in the geological repository can be detrimental to the disposal safety of the nuclear waste form.

36 MATERIALS SCIENCE↗

Engineered barrier material interactions at elevated temperatures: Bentonite-metal interactions under elevated temperature conditions

The development of deep repository concepts in the USA is evaluating generic options for disposal of heat-generating nuclear spent nuclear fuel (SNF) waste for a suite of host rock media (clay rock, granite, salt). Large waste canister designs (e.g., dual purpose canisters or DPC’s) are currently being considered to accommodate many SNF assemblies. Large amounts of SNF in the waste packages will produce high thermal loads generating temperatures in excess of 200 °C for long periods of time. For a bentonite backfilled repository concept, prolonged exposure to high temperatures will induce chemical reactions in the engineered barrier system (EBS), particularly at barrier interfaces. Our focus is on experimental investigations and the application of thermodynamic modeling to evaluate clay-zeolite phase equilibria as a function of temperature and fluid chemistry. Experimental work on barrier material interactions under hydrothermal conditions (150 – 300 °C, 15-16 MPa) has elucidated mineral phase changes in Wyoming (Colony mine) bentonite in the presence of steel phases. Glassy material in bentonite is replaced by analcime-wairakite phases, and through apparent clinoptilolite recrystallization. The initial increase in dissolved silica leads to authigenic quartz formation. Such mineral assemblage suggests an initially silica-rich environment (analcime, clinoptilolite) moving towards Si-depleted conditions. Analcime-wairakite compositions suggest a well-defined solid-solution between these Na and Ca end-members. Smectite clay in these experiments is stable with Fe-saponite and chlorite growth co-existing with binary/ternary sulfides at steel interfaces. Little or no illite was observed in the reaction products which could be tied to silica oversaturation and low K in the system. The thermodynamic analysis is used to evaluate thermodynamic data and develop phase diagrams to describe stability field relations of secondary mineral phase occurrences. This analysis allows for delineation of potential reaction pathways in bentonite clay degradation and interactions with metallic phases. One example of this is the important role of dissolved silica plus other phase components to the formation of alteration mineralogy observed in these interactions. All these investigations are key to the assessment of thermally-induced degradation zones in the EBS during the thermal period and their effect on barrier performance in the safety assessment.

Jove-Colon, Carlos↗

Used Nuclear Fuel Management in Asia

Used fuel discharged from nuclear reactors are presently stored in wet pools or in dry cask at-reactor and/or away-from- reactor sites. Only a few countries reprocess their civilian used fuel, notably France, Russia, and UK. No country has yet operated a geologic repository to dispose of civilian used fuel or high-level radioactive wastes (HLW). As a result, the management and disposal of used fuel and HLW are viewed as the Achilles' heel of nuclear power. Figure 1 shows the used fuel inventory in the United States (US) and East Asian nuclear power programs. The inventory is estimated based on that in 2010 and the projected growth of nuclear power in these programs to 2020 and 2030. Figure 1 indicates that in 2010, the US had about 63,900 tons of used fuel from light-water reactors (LWR), more than the statutory limit of what was supposed to go into the now-abandoned Yucca Mountain Repository, and accounting for more than 33% of global used LWR fuel total. Currently, about 2,000 tons of used fuel are generated in the existing US LWRs annually. Russia had 25,100 tons of used fuel in 2010. Based on a modest growth of nuclear power, its used fuel inventory is expected to increase to about 31,800 and 38,000 tons by 2020 and 2030, respectively. The nuclear energy programs in East Asia (China, Japan, ROK, and Taiwan) account for 19% of global used LWR fuel inventory in 2010. China's used fuel inventory in 2010 was relatively small at 1,900 tons, but is expected to grow rapidly to {approx}12,400 and {approx}31,000 tons in 2020 and 2030, respectively, based on its expanding nuclear power program (from the current installed capacity of 21.8 GW to {approx}60 and {approx}200 GW in 2020 and 2030, respectively). Besides PWRs, China also operates 2 Candu reactors which each produces roughly 8 times more used fuel in tons than a PWR of the same power. The used Candu fuel inventory would increase from 950 tons in 2010 to 2,850 and 4,750 tons in 2020 and 2030, respectively. Used fuel in Japan, ROK, and Taiwan are US-obligated. Japan's future trends of used fuel inventory are uncertain due to the unknown numbers and timing of the currently-shutdown nuclear power plants that would be allowed for restart, and the uncertain future of Japan's reprocessing at Rokkasho. Under two scenario cases (a no-nuke/no-reprocessing and a 80%-nuke with reprocessing), the Japanese used fuel inventory could be capped from 2020 to 2030 to {approx}21,500 tons for the first case, and to {approx}22,500 tons for the second. The ROK's used PWR fuel inventory would increase from 4,900 tons in 2010 to about 8,800 and 14,000 tons in 2020 and 2030, respectively. Its used Candu fuel inventory would also increase from 5,900 tons in 2010 to about 9,700 and 13,500 tons in 2020 and 2030, respectively. The Taiwan used fuel inventory should be kept at current level as nuclear power is being phase out. Its growth should be modest from 3,000 tons in 2010 to about 4,300 and 5,000 tons in 2020 and 2030, respectively even if the 4. nuclear power station would start and the operation of existing LWRs is extended. Used fuel inventory in other East Asian countries (DPRK, Vietnam, etc.) should be small, if any before 2030.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Codisposal Waste Package Loading Options for DOE SNF and HLW PPT

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Co-Disposal Waste Package Loading Options for DOE SNF and HLW - 20331

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗