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Technical Approach for Determining Key Parameters Needed for Modeling the Performance of Cast Stone for the Integrated Disposal Facility Performance Assessment

Washington River Protection Solutions (WRPS) and its contractors at Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) are conducting a development program to develop / refine the cementitious waste form for the wastes treated at the ETF and to provide the data needed to support the IDF PA. This technical approach document is intended to provide guidance to the cementitious waste form development program with respect to the waste form characterization and testing information needed to support the IDF PA. At the time of the preparation of this technical approach document, the IDF PA effort is just getting started and the approach to analyze the performance of the cementitious waste form has not been determined. Therefore, this document looks at a number of different approaches for evaluating the waste form performance and describes the testing needed to provide data for each approach. Though the approach addresses a cementitious secondary aqueous waste form, it is applicable to other waste forms such as Cast Stone for supplemental immobilization of Hanford LAW. The performance of Cast Stone as a physical and chemical barrier to the release of contaminants of concern (COCs) from solidification of Hanford liquid low activity waste (LAW) and secondary wastes processed through the Effluent Treatment Facility (ETF) is of critical importance to the Hanford Integrated Disposal Facility (IDF) total system performance assessment (TSPA). The effectiveness of cementitious waste forms as a barrier to COC release is expected to evolve with time. PA modeling must therefore anticipate and address processes, properties, and conditions that alter the physical and chemical controls on COC transport in the cementitious waste forms over time. Most organizations responsible for disposal facility operation and their regulators support an iterative hierarchical safety/performance assessment approach with a general philosophy that modeling provides the critical link between the short-term understanding from laboratory and field tests, and the prediction of repository performance over repository time frames and scales. One common recommendation is that experiments be designed to permit the appropriate scaling in the models. There is a large contrast in the physical and chemical properties between the Cast Stone waste package and the IDF backfill and surrounding sediments. Cast Stone exhibits low permeability, high tortuosity, low carbonate, high pH, and low Eh whereas the backfill and native sediments have high permeability, low tortuosity, high carbonate, circumneutral pH, and high Eh. These contrasts have important implications for flow, transport, and reactions across the Cast Stone – backfill interface. Over time with transport across the interface and subsequent reactions, the sharp geochemical contrast will blur and there will be a range of spatially-distributed conditions. In general, COC mobility and transport will be sensitive to these geochemical variations, which also include physical changes in porosity and permeability from mineral reactions. Therefore, PA modeling must address processes, properties, and conditions that alter the physical and chemical controls on COC transport in the cementitious waste forms over time. Section 2 of this document reviews past Hanford PAs and SRS Saltstone PAs, which to date have mostly relied on the lumped parameter COC release conceptual models for TSPA predictions, and provides some details on the chosen values for the lumped parameters. Section 3 provides more details on the hierarchical modeling strategy and processes and mechanisms that control COC release. Section 4 summarizes and lists the key parameters for which numerical values are needed to perform PAs. Section 5 provides brief summaries of the methods used to measure the needed parameters and references to get more details.

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Effects of alteration product precipitation on glass dissolution

Understanding the mechanisms that control the durability of nuclear waste glass is paramount if reliable models are to be constructed so that the glass dissolution rate in a given geological repository can be calculated. Presently, it is agreed that (boro)silicate glasses dissolve in water at a rate dependent on the solution concentration of orthosilicic acid (H 4 SiO 4 ) with higher [H 4 SiO 4 ] leading to lower dissolution rates. Once the reaction has slowed as a result of the buildup of H 4 SiO 4 , another increase in the rate has been observed that corresponds to the precipitation of certain silica-bearing alteration products. However, it has also been observed that the concentration of silica-bearing solution species does not significantly decrease, indicating saturation, while other glass tracer elements concentrations continue to increase, indicating that the glass is still dissolving. In this study, we have used the Geochemist’s Workbench code to investigate the relationship between glass dissolution rates and the precipitation rate of a representative zeolitic silica-bearing alteration product, analcime [Na(AlSi 2 O 6 )∙H 2 O]. To simplify the calculations, we suppressed all alteration products except analcime, gibbsite (Al(OH) 3 ), and amorphous silica. The pseudo-equilibrium-constant matrix for amorphous silica was substituted for the glass pseudo-equilibrium-constant matrix because it has been shown that silicate glasses act as a silica-only solid with respect to kinetic considerations. In this article, we present the results of our calculations of the glass dissolution rate at different values for the analcime precipitation rate constant and the effects of varying the glass dissolution rate constant at a constant analcime precipitation rate constant. From the simulations we conclude, firstly, that the rate of glass dissolution is dependent on the kinetics of formation of the zeolitic phase. Therefore, the kinetics of secondary phase formation is an important parameter that should be taken into account in future glass dissolution modeling efforts. Secondly, the results indicate that, in the absence of a gel layer, the glass dissolution rate controls the rate of analcime precipitation in the long term. Finally, the meaning of these results pertinent to long-term glass durability is discussed.

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M4SF-20LL010301042: Updated Thermodynamic Database for Use in Generic Disposal System Assessment

This progress report (Level 4 Milestone Number M4SF-20LL010301042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Activity Number SF-20LL010301041. LLNL is leading efforts in the development of thermodynamic databases in support of the Spent Fuel and Waste Science Technology (SFWST) program. Thermodynamic models provide the basis for understanding the stability of solid phases and speciation of aqueous species and modeling the evolution of repository conditions. The LLNL effort is being performed in coordination with other US database development efforts. The effort includes a review of available thermochemical databases and a path forward for database integration. International coordination with the NEA-TDB is supported through crystalline international work package.

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SUPCRTNE Thermodynamic Database Engine

This progress report (Level 4 Milestone Number M4SF-21LL010301052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Activity Number SF-21LL01030105. LLNL is leading efforts in the development of thermodynamic databases in support of the Spent Fuel and Waste Science Technology (SFWST) program. Thermodynamic models provide the basis for understanding the stability of solid phases and speciation of aqueous species and modeling the evolution of repository conditions. The LLNL effort is being performed in coordination with other US database development efforts. The effort includes a review of available thermochemical databases and a path forward for database integration. International coordination with the NEA-TDB is supported through crystalline international work package.

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A Generic Study of High-Level Radioactive Waste Isolation in Crystalline Rock - 19633

This paper is based on a numerical study aimed at predicting long-term (1 Ma period) placement room performance of a generic site for a Deep Geologic Repository (DGR) of used nuclear fuel in a crystalline rock setting. As part of the study, coupled Thermo-Hydro-Mechanical (THM) processes were analyzed in a sequence of bounding numerical simulations of rare, strong earthquakes, advancing and retreating glacial ice loading, extreme repository-induced effects including used-fuel canister heat generation and repository gas pressure generation, and considering long-term rock strength degradation. The details of the THM modeling approach, and the selection and evaluation of the poro-mechanical parameters used in the analyses are described. The results of these analyses are presented and compared with results of analyses where evolution of pore pressure due to temperature change and pore pressure dissipation were not considered. The sensitivity of the response to a variation of poro-elastic parameters used in the analyses is discussed. (authors)

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Gas-phase stability of large lanthanide:diglycolamide clusters evaluated using collision-induced dissociation

Efficient utilization of long-term deep geologic storage repositories is critical for the large-scale deployment of nuclear energy. As the minor actinides (americium and curium) are the largest contributor to the decay heat of used nuclear fuel after a few hundred years, their removal from used fuel prior to disposal can significantly increase the amount of waste that can be stored in a given volume. The development of ligands that can effectively separate the minor actinides from other components of used nuclear fuel are crucial to efficient utilization of geological storage repositories. N,N,N’,N’-tetraoctyl diglycolamide (TODGA) is a promising extractant for the separation of lanthanides and the minor actinides from other components of used nuclear fuel. While the use of TODGA in f-element separations has been investigated in process-based formulations, gas-phase metal ion cluster experiments enable the study of covalent interactions in reprocessing systems absent from solvent effects. Exploring fundamental differences in lanthanide-ligand covalent interactions can impact the development and implementation of actinide-lanthanide separation systems in nuclear fuels reprocessing. In this study, lanthanide-TODGA clusters were synthesized in the gas-phase and identified using mass spectrometry fragmentation experiments. Large europium and samarium clusters were identified that contained up to 8 and 10 bound TODGA ligands, respectively; this was surprising due to the size and multidentate binding that is normal for TODGA. In addition, while smaller clusters showed evidence of sequential ligand fragmentation, larger clusters displayed the loss of neutral TODGA with applied collision voltage. Interestingly, the voltage required for this removal decreased as more TODGA ligands were bound to the metal, suggesting that the metal coordination sphere was becoming more saturated and TODGA ligands were more weakly bound as the clusters got larger.

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Impact of Corrections to the Spallings Volume Calculation on Waste Isolation Pilot Plant Performance Assessment

The numerical code DRSPALL (from direct release spallings) is written to calculate the volume of Waste Isolation Pilot Plant solid waste subject to material failure and transport to the surface (i.e., spallings) as a result of a hypothetical future inadvertent drilling intrusion into the repository. An error in the implementation of the DRSPALL finite difference equations was discovered and documented in a software problem report in accordance with the quality assurance procedure for software requirements. This paper describes the corrections to DRSPALL and documents the impact of the new spallings data from the modified DRSPALL on previous performance assessment calculations. Updated performance assessments result in more simulations with spallings, which generally translates to an increase in spallings releases to the accessible environment. Total normalized radionuclide releases using the modified DRSPALL data were determined by forming the summation of releases across each potential release pathway, namely borehole cuttings and cavings releases, spallings releases, direct brine releases, and transport releases. Because spallings releases are not a major contributor to the total releases, the updated performance assessment calculations of overall mean complementary cumulative distribution functions for total releases are virtually unchanged. Therefore, the corrections to the spallings volume calculation did not impact Waste Isolation Pilot Plant performance assessment calculation results.

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Impact of Corrections to the Spallings Volume Calculation on Waste Isolation Pilot Plant Performance Assessment [Poster]

The numerical code DRSPALL (from direct release spallings) is written to calculate the volume of Waste Isolation Pilot Plant solid waste subject to material failure and transport to the surface (i.e., spallings) as a result of a hypothetical future inadvertent drilling intrusion into the repository. An error in the implementation of the DRSPALL finite difference equations was discovered and documented in a software problem report in accordance with the quality assurance procedure for software requirements. This paper describes the corrections to DRSPALL and documents the impact of the new spallings data from the modified DRSPALL on previous performance assessment calculations. Updated performance assessments result in more simulations with spallings, which generally translates to an increase in spallings releases to the accessible environment. Total normalized radionuclide releases using the modified DRSPALL data were determined by forming the summation of releases across each potential release pathway, namely borehole cuttings and cavings releases, spallings releases, direct brine releases, and transport releases. Because spallings releases are not a major contributor to the total releases, the updated performance assessment calculations of overall mean complementary cumulative distribution functions for total releases are virtually unchanged. Therefore, the corrections to the spallings volume calculation did not impact Waste Isolation Pilot Plant performance assessment calculation results.

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Impact of Corrections to the Spallings Volume Calculation on Waste Isolation Pilot Plant Performance Assessment - 16135

The numerical code DRSPALL (from direct release spallings) is written to calculate the volume of Waste Isolation Pilot Plant solid waste subject to material failure and transport to the surface (i.e., spallings) as a result of a hypothetical future inadvertent drilling intrusion into the repository. An error in the implementation of the DRSPALL finite difference equations was discovered and documented in a software problem report in accordance with the quality assurance procedure for software requirements. This paper describes the corrections to DRSPALL and documents the impact of the new spallings data from the modified DRSPALL on previous performance assessment calculations. Updated performance assessments result in more simulations with spallings, which generally translates to an increase in spallings releases to the accessible environment. Total normalized radionuclide releases using the modified DRSPALL data were determined by forming the summation of releases across each potential release pathway, namely borehole cuttings and cavings releases, spallings releases, direct brine releases, and transport releases. Because spallings releases are not a major contributor to the total releases, the updated performance assessment calculations of overall mean complementary cumulative distribution functions for total releases are virtually unchanged. Therefore, the corrections to the spallings volume calculation did not impact Waste Isolation Pilot Plant performance assessment calculation results. (authors)

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Regulatory Compliance Strategy For Recovery-Driven Underground Ventilation System Changes - 16355

The Waste Isolation Pilot Plant (WIPP) facility is managed and operated by Nuclear Waste Partnership LLC (NWP) for the U.S Department of Energy (DOE). On February 14, 2014, a radiological event occurred in the WIPP facility underground repository, which resulted in the release of transuranic isotopes contaminating a portion of the underground primarily along the ventilation path from Panel 7, the location of the incident. The release was initially detected by an underground continuous air monitor (CAM) which caused an automatic switch of the ventilation system to filtration mode and ventilation was directed through high-efficiency particulate air (HEPA) filtration banks located in the surface exhaust building. When the HEPA filtration was engaged, the underground normal ventilation was reduced from 425,000 standard cubic feet per minute (scfm) to 60,000 scfm. Since the event, all exhaust air is filtered through HEPA filters at 60,000 scfm to minimize the possibility of exposure to workers, the public and the environment. Ventilation of fresh air to the WIPP underground is a primary limitation to performing underground recovery activities; therefore, recovery-driven ventilation changes are needed. The safe recovery of the WIPP facility and resumption of normal waste emplacement activities require an increase in ventilation capacity to be accomplished in three phases: 1) Interim Ventilation System, 2) Supplemental Ventilation System and 3) Permanent Ventilation System. Each phase adds increased air flow to the underground and allows for underground activities to be performed concurrently. The regulatory strategy for ventilation recovery focuses on ensuring that needed changes to the WIPP facility are protective of human health and the environment. Recovery-driven ventilation changes are implemented using the procedures, processes, and regulations that are required by the two primary regulatory agencies for the WIPP facility: the U.S. Environmental Protection Agency (EPA) and the New Mexico Environment Department (NMED) as well as DOE requirements. Regulatory compliance is being achieved for the three ventilation phases to restore underground ventilation by providing the regulators and stakeholders opportunities for involvement during the design and development processes. A strategy for a tiered NMED permitting process and a plan for developing required EPA documentation are under consideration. The Project's objective is to resume normal waste emplacement activities with safety, protection of the public, the worker and the environment as the highest priorities. Important elements of ensuring that WIPP facility recovery and future waste emplacement activities occur as planned are transparency and communication with the regulatory community. (authors)

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Report on the Status of the SFWST Campaign International Activities in Disposal Research at SNL

The following summaries are provided as fulfillment of milestone M4SF-17SN080305022 and represent international coordination activities in disposal research funded by the US DOE Spent Fuel and Waste Storage and Technologies (SFWST) Campaign during Fiscal Year 2017: SFWST funded bi-lateral interactions with Taiwan, OECD-NEA Repository Metadata (RepMet) project, SFWST funded bi-lateral interactions with the Republic of Korea.

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Standard Analysis Report INV-SAR-81, Revision 0 Chemical and Cement Components 2023 Inventory Estimates for the Interim State Performance Assessment of the Waste Isolation Pilot Plant

This standard analysis report provides the estimates for the chemical (oxyanions and complexing agents) and cement components with a data collection cut-off date of December 31, 2023. These estimates will be included in a Performance Assessment Inventory Report developed for the U.S. Department of Energy (DOE) interim state repository configuration performance assessment (PA) for the 2026 Compliance Recertification Application (CRA).

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Swedish Solution for Reconditioning, Transport, Interim Storage and Final Disposal of Failed Fuel - 17024

Despite today's increasingly more reliable operation of nuclear power plants, failed fuel rods exist in all Swedish plants. Managing intermediate storage and sub-sequential final disposal of such rods has been a growing problem. The Swedish system for spent fuel management uses the principle of intermediate storage of the fuel in a central intermediate storage facility (CLAB) before disposal in a final repository according to the KBS-3 method. However, the intermediate storage facility does not allow storage of failed fuel and the final repository has limitations on moisture in the deposited fuel and hence failed or broken fuel rods cannot be accepted as-is. The nuclear power companies in Sweden have jointly established the Swedish Nuclear Fuel and Waste Management Co, SKB, to manage the spent fuel and the operational waste. To find a solution to transport the failed fuel to storage, SKB has established a project with the goal to have all currently failed fuel rods from all the Swedish nuclear power plants reconditioned and transported to the intermediate storage facility before the year of 2020. The project has established the acceptance criteria for reconditioned failed fuel in the final repository and evaluated a number of options for treatment, encapsulation and transportation of the failed fuel rods. Furthermore, two suppliers have developed reconditioning methods fulfilling the criteria and contracts were signed with both suppliers in May 2016. This paper outlines the details of the Swedish solution for treatment and storage of failed fuel including the technical requirements for the reconditioning method and transportability. (authors)

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Immediate Dismantling of a Large Fleet of LWR NPPs: Consequences for Spent Fuel and Waste Management - 16027

The experience feedback shows that the complete dismantling of one Light Water Reactor (LWR) unit is an activity now globally well under control (reactor pressure vessel and its internals removal and cutting included). However some specific industrial difficulties may arise in view of the dismantling of numerous LWR units simultaneously. Indeed more than 300 LWR units are currently operated worldwide, which have been commissioned these last 50 years. So in some countries, many units may be permanently shut down over a period of few years in the next decades. Such a situation addresses the issue of the overall management of large quantities of spent nuclear fuel (SNF) and decommissioning waste for the concerned units. For example, in France, the legislative and regulatory framework for the nuclear facilities favors their dismantling 'as soon as possible' after their permanent shutdown which implies as well to limit the duration of the transition period from operation to decommissioning. Furthermore 58 LWR units have been commissioned between 1977 and 1999 in France - on average more than 2 units per year. In this context, the operator (Electricite de France - EDF) plans to remove quickly the SNF then to perform dismantling actions immediately after the permanent shutdown of the LWR units. One issue is to remove the SNF from all the relevant units, even if this removal is simultaneous in many units (permanently shut down or still under operation). Similar issue has to be taken into account regarding the management of the radioactive waste (RW) produced by the dismantling and clean-up actions, notably the RW that cannot be disposed of in a near surface repository. One method followed by the French technical support organization (Institut de Radioprotection et de Surete Nucleaire - IRSN) to analyze these issues is the use of estimates of flows of SNF and RW. These estimates are based notably on a phasing-scenario and a planning template defined for the decommissioning of the units of one nuclear power plant (NPP) and coupled to an overall schedule for phase out all the units of the fleet. The estimates relative to the next decades can be compared to the current experience feedback of flows of SNF and RW for units under operation, in order to identify risks when facing decommissioning. The risks highlighted are driven by key parameters (as duration of the main dismantling actions) of the estimates which can be adapted to minimize their impact. On this basis, it is possible to identify the key-factors to dismantle each unit of NPPs and phase out the fleet regarding SNF and RW management. It is noteworthy that this work needs to be done in any case upstream the studies and the implementation of dismantling actions. Finally it can be underlined that another issue is the human resources (staff, skills and knowledge) necessary to perform all the decommissioning actions, but this aspect is not addressed in the present paper. (authors)

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Sensitivity analysis of generic deep geologic repository with focus on spatial heterogeneity induced by stochastic fracture network generation

Geologic Disposal Safety Assessment Framework is a state-of-the-art simulation software toolkit for probabilistic post-closure performance assessment of systems for deep geologic disposal of nuclear waste developed by the United States Department of Energy. This paper presents a generic reference case and shows how it is being used to develop and demonstrate performance assessment methods within the Geologic Disposal Safety Assessment Framework that mitigate some of the challenges posed by high uncertainty and limited computational resources. Variance-based global sensitivity analysis is applied to assess the effects of spatial heterogeneity using graph-based summary measures for scalar and time-varying quantities of interest. Behavior of the system with respect to spatial heterogeneity is further investigated using ratios of water fluxes. This analysis shows that spatial heterogeneity is a dominant uncertainty in predictions of repository performance which can be identified in global sensitivity analysis using proxy variables derived from graph descriptions of discrete fracture networks. New quantities of interest defined using water fluxes proved useful for better understanding overall system behavior.

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Multilateral Spent Fuel Management Efforts: Nonproliferation and Energy Implications in the Asia Pacific

Management of spent nuclear fuel in the Asia Pacific will be a growing challenge in coming decades as the region continues its rapid economic and nuclear energy expansion. Accompanying this growth has been skyrocketing energy demand. The capacity of nuclear energy in the Asia Pacific is expected to increase by approximately 50% by 2035. In fact, of the 66 reactors under construction worldwide, 47 are being built in the Asia Pacific region alone. In a sense, the area is experiencing the so-called ‘Nuclear Renaissance’, which could simultaneously increase proliferation risks while fueling economic growth and political stability. And because ‘Asia Pacific’ includes North America, Oceania, East Asia, Southeast Asia and, by some definitions, South and Central Asia as well as the Middle East, the demand for nuclear energy in particular is expected to continue. Consequently, the Asia Pacific region will soon face important decisions about how to support and implement the necessary infrastructure for nuclear energy expansion and deal with the waste, while also meeting all nonproliferation obligations under the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). One option for dealing with spent nuclear fuel that should be considered is multilateral spent fuel management. Such an effort could help reduce risks associated with safety, proliferation and security of nuclear energy. While the original idea of multilateral spent nuclear fuel management is a decades-old concept, no official plan has been developed because of challenges related to cost, liability, transportation, export controls, international commerce, and nonproliferation, to name a few. Nonetheless, as more and more newcomer countries decide to develop nuclear energy, but do not yet have a strategic plan or the financial means to deal with the waste, the international community may need to take a more proactive approach in developing a multilateral effort, particularly with respect to geological repositories and interim dry storage. Risks and benefits will need to be analyzed, while the role of such an effort in the revival of the Nuclear Renaissance will need to be evaluated. This paper provides an overview of spent nuclear fuel management efforts in the Asia Pacific region and attempts to analyze the possible benefits and feasibility of a multilateral approach.

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Aluminum Clad SNF: Technical Considerations and Challenges for Extended (>50 years) Dry Storage - 1900y

The U.S. Department of Energy (DOE) manages a significant inventory (>13 MTHM) of aluminum clad spent nuclear fuel (ASNF) using a combination of wet and dry storage at three primary sites; Hanford, Idaho National Laboratory (INL), and Savannah River Site (SRS). The majority of this ASNF is generated from U.S. and international research test reactors (RTR), examples include Advanced Test Reactor (ATR) at INL and High-Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Obama Administration suspension of the Yucca Mountain repository, issuance of the January 2013 DOE 'Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste,' delays in implementing the 'regionalization' SNF approach recommended in the 1995 'Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement,' and limited ASNF processing capability at SRS H-Canyon means that interim storage has become the de-facto strategy for managing ASNF. The 1995 Idaho Settlement Agreement requires the DOE to transfer all SNF stored at INL to interim dry storage prior to December 31, 2023 and wet storage capacity at the SRS is nearing capacity. This situation was recognized by the DOE Spent Nuclear Fuel Working Group (SNFWG) and at its June 2016 meeting formed a ASNF sub-working group comprised subject matter experts to evaluate this area and provide a report of findings. The ASNF sub-working group met many times over several month period to review the engineering and technical basis of existing and planned DOE ASNF dry storage facilities and additional studies on the behavior of ASNF under different conditions, including during reactor service. The ASNF sub-working group focused its review on RTR SNF. Based on these reviews, the ASNF sub-working group identified five knowledge gaps and technical data needs that should be addressed to ensure viability and safety of extended, greater than 50 years, sealed or vented interim dry storage of ASNF. These five areas include; behavior/chemistry of aluminum oxyhydroxide layer in dry storage, radiolytic gas generation properties of aluminum oxyhydroxide layers, effect of episodic breathing and radiolytic generation of corrosive gases during storage, performance of ASNF in existing dry storage systems, and effects of high temperature (>1000 C) drying on chemistry and behavior of aluminum oxyhydroxide layers. The ASNF sub-working group also recommended development of an action plan to identify technical and engineering activities needed to address the knowledge gaps and technical data needs. Results of the ASNF sub-working group are documented in 2017 report 'Aluminum Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 years) Dry Storage,' DOE-ID/RPT-1575. The INL assembled a team comprised of twenty INL and Savannah River National Laboratory SNF management and subject matter experts to develop this action plan. Over the course of 2 months the team worked to develop the action plan. The plan identified and recommended six technical and engineering tasks needed to address the previously identified knowledge gaps and technical data needs. DOE Environmental Management Office of Technology Development is providing funding to the INL to execute these tasks. This paper will present and discuss the results of the ASNF sub-working group, tasks identified in the action plan, and experimental activities underway at INL and SRNL to address the action plan. (authors)

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Radiation Induced Corrosion of Copper in Deep Geological Repositories - 19493

This paper presents an overview of the Cu corrosion research being conducted by the Radiation Assisted Materials Performance Science (RAMPS) group at Western University, in collaboration with the Canadian Nuclear Waste Management Organization (NWMO) and Canada Natural Science and Engineering Research Council (NSERC). This paper presents an overview of our research in three areas to develop a high-fidelity copper corrosion model. Copper is an important candidate material for corrosion protection of used nuclear fuel containers (UFC). The UFC and its environment will be exposed to a continuous flux of γ-radiation. The three areas are chemical kinetics of γ-radiolysis of humid air and groundwater, the effects of solution parameters on Cu corrosion dynamics without γ-radiation and the effects of γ-radiation on Cu corrosion dynamics. These studies show that because of the strong feedback loops between processes occurring during copper corrosion, the overall corrosion mechanism may not follow linear dynamics. Presented here are our key findings, their implications for corrosion science in general, and the potential implications of these results for predictions of the adequacy of a UFC copper corrosion allowance. (authors)

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