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At least 1,117 records · Page 62

A study of transfer of UNF from non-disposable canisters - 15388

A study was performed by AREVA and URS for the Department of Energy (DOE) [1] to establish optimal methods for the opening of dry storage canisters (DSC) and the subsequent repackaging of used nuclear fuel (UNF) from the DSCs into generic standard transportation, aging, and disposal canisters (STADs) in wet or dry environments at reactor, consolidated storage, or repository sites. Eight dry and wet cutting technologies were evaluated: (1) skiving; (2) milling; (3) arc gouging/plasma torch; (4) cutting wheel; (5) diamond cable cutting; (6) laser cutting; (7) sawing; and (8) electrical discharge machining. The evaluated UNF repackaging technologies options included: (1) wet transfer in a pool; (2) dry transfer in a dry transfer system (DTS); (3) dry transfer in a fixed hot cell (FHC); (4) dry transfer in a mobile hot cell (MHC), and (5) a hybrid approach utilizing wet transfer in a pool and dry transfer in a FHC. A systems engineering evaluation was utilized for ranking the eight viable cutting options and the five different repackaging approaches for each of the sites. The systems engineering evaluation took into account six criteria (i.e., life cycle costs, worker dose, operational efficiency, technical maturity, ease of licensing, and waste generation) and performed a qualitative pairwise ranking of their relative importance to one another. Then quantitative or qualitative factors associated with each criterion (e.g., cost in millions of dollars, relative ease of licensing, etc.) for each option at each location (reactor, consolidated storage facility [CSF], repository) were developed based on historical experience, existing design studies, current 'state-of-the art', and/or engineering judgment. Using this input, the systems engineering evaluation identified dry skiving as the preferred cutting technology, with the wheel cutting technology in either a dry or wet environment identified as a promising technology that, with some RD and D, could mature to a better option (depending also on the disposition of the used DSC). For a reactor and a CSF/repository site, the systems engineering evaluation ranked the repackaging option using the skiving dry cutting technique and wet transfer of UNF in a pool as the highest option. However for a CSF and a repository site, the team recommended the hybrid option using the skiving dry cutting technique and both dry and wet transfer of UNF in a pool and FHC as a potentially better option (providing specific benefits such as higher throughput and more flexibility in operation). (authors)

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An underground laboratory in the context of salt disposal research - 15549

Evaluating the basis of need for an underground research laboratory for salt science and engineering is imperative because of the significant commitment of time and money involved. Decades of salt repository studies, numerous experiments, and sophisticated modeling capabilities underpin the scientific database that supports safe disposal of nuclear waste in salt. The safety case for disposal of non-heat-generating waste such as transuranic waste interred at the Waste Isolation Pilot Plant is robust, with the only long-term releases to the environment projected to be by way of human intrusion. The scientific evidence also favors safe disposal of heat-generating waste. Technical evaluations for disposal of heat-generating waste in salt experienced a rather long hiatus in the United States subsequent to certification of the Waste Isolation Pilot Plant and issuance of the Nuclear Waste Policy Act Amendment that ended salt disposal research for the civilian program. Similar salt repository research in Germany was delayed by a ten-year moratorium that ended in 2010. In collaboration with German peers, the United States Department of Energy has reviewed and evaluated thermally driven processes in salt disposal and identified key technical areas in which to prioritize resources. The goal for disposal research in salt is to provide sufficient technical information to license a repository successfully. The necessity or utility of a salt underground laboratory is to be evaluated in the context of an overall research agenda that supports a license application. (authors)

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Retrieval of High Level Waste for the Cigeo Project - A Practical Retrieval Test Story - 16085

Since the passing of the June 28, 2006 Act, Andra has incorporated in its step-wise design development of the French high level waste (HLW) and intermediate level long lived waste (IL-LLW) deep geological repository (DGR, aka Cigeo) various requirements linked to 'reversibility' at different stages of the repository life cycle (from its construction phase to its final closure). Reversibility is based on a number of practical dispositions including the progressivity of the development and deployment of the facility, the flexibility of the nature of operations being carried out, the adaptability of the facility to future modifications and the possible retrieval of waste. The capacity to retrieve waste from Cigeo, also called 'retrievability', bears specific requirements. On a shorter term, one must demonstrate that the waste container retrieval operations are technically safe and feasible over at least 100 years. The current paper starts with two acceptations: retrievability and reversibility. It details how those two notions are defined and inter-linked. The 'Retrievability Scale' as proposed by OECD/NEA (and adopted by Andra) is also reminded. The needs for demonstration of retrieval operations are also exposed for the two categories of waste containers disposed of in Cigeo (IL-LLW and HLW) within the context of their respective emplacement configurations (in disposal horizontal vaults and micro-tunnels) at different stages of the repository life cycle. The technical part of the paper is then focused on a practical case story: a technological test campaign (still running) of HLW container retrieval out of a steel cased disposal micro-tunnel. On a test bench built on surface to simulate (create) the environmental conditions that may prevail inside a HLW micro-tunnel at time of retrieval, it was possible to check whether the mechanical means developed for the purpose were adequate and sufficient. The test case story is exposed, c/w the problems and challenges encountered (high temperature impairing the retrieval robot's proper functioning, presence of corrosion products...). The trouble-shooting solutions implemented and the results are commented. This article concludes with a critical analysis of the methods and equipment used, the results obtained and provides some perspectives of improvement. The positive outcomes of this technical test campaign pave the way for planning the in situ retrieval tests to be programmed in the 'industrial pilot phase' of Cigeo (to be implemented over the ten first years of operations, including the active and inactive test period and the first years of disposal activities): the industrial pilot phase is presently scheduled around 2026-2036 (provided the Cigeo construction is authorized in 2021, following the licensing request filing scheduled in 2018). The issues of IL-LLW containers retrieval (for which specific solutions are also studied and technical demonstrations engaged or planned) are also briefly presented. (authors)

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Spent Nuclear Fuel Management and Disposal in Finland - Status and Lessons Learned at the End of 2015 - 16507

In November 2015 the Finnish Government approved and the Parliament ratified Posiva's 2012 License Application for Construction of a mined, deep geological repository for the spent nuclear fuel (SNF) at the Olkiluoto site in the Eurajoki municipality based upon recommendations provided by the Finnish Radiation Safety Authority (STUK), international expert groups and long-standing local support. The preceding siting process began in 1983 and it included disposal-concept selection (KBS-3V), 'consent-based' site selection and large-scale in-situ tests conducted in an underground research laboratory, the ONKALO, situated ∼440 m below the mean Sea level immediately adjacent to the planned repository at the Olkiluoto site. At the end of 2015, the Finnish SNF repository is projected to open in 2024 as the first of its kind. Four related lessons learned in Finland during the past 30 years are: 1. The selection of potentially-suitable sites for geological disposal of SNF proved to be more difficult and more time consuming than initially envisioned. 2. The siting of nuclear facilities has evolved from being mainly a radionuclide containment and isolation challenge to becoming a critical moving-target, socio-political challenge governing progress with majority local acceptance being a fundamental imperative. 3. Public and political acceptance and support are strongly linked to the competency, credibility and transparency of the involved organisations and the maturity/pedigree of the proposed disposal concept. 4. Timely sharing lessons learned and conducting selective collaborations with other countries proved to be successful keys to timely and cost-effectively obtain and maintain public and political acceptance and support. (authors)

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What's in WIPP? Packaging TRU Waste to Enhance WIPP's Capacity - 16515

Beginning in 1999, and continuing for 15 years, the US DOE Waste Isolation Pilot Plant (WIPP) repository successfully disposed of defense Transuranic (TRU) waste. Six of the originally planned ten disposal panels were filled before two unrelated accidents temporarily halted disposal operations in February 2014. Based on counting the gross volume of all containers emplaced in WIPP to date, a little more than half of the legislated volume capacity has been subscribed (91,000 cubic meters). This paper describes the many different packaging configurations that have been used to ship TRU waste to WIPP, and makes observations that may enhance the repository capacity when shipments resume. The contents of shipments to WIPP are limited by many requirements of the repository waste acceptance criteria. These criteria derive from three primary sources; 1) the Certificate of Compliance (CoC), issued by the NRC for Type B shipping packaging used by WIPP; 2) The Hazardous Waste Facility Permit, issued by the New Mexico Environment Department; and 3) the WIPP facility's own Documented Safety Analysis (DSA), developed under DOE's regulation 10CFR830. To meet these criteria, various TRU waste forms have been packaged in different ways, with varying packaging efficiency. Of course, there is an overall upper limit on WIPP's volume capacity that was legislated as part of the WIPP Land Withdrawal Act. The volume counted against this limit has been calculated by simply summing the volumes of individual payload containers shipped to and emplaced at WIPP. This method counts significant non-waste volumes inside the containers due to over-packing inner containers and partially filled containers. Without some change to how disposal volumes are calculated, or how waste is packaged, the amount of TRU waste yet to be shipped to WIPP could exceed the legislated capacity. Concepts to enhance the packaging efficiency of future waste shipments to WIPP are presented. Approximate estimates of the waste versus non-waste volumes that have been emplaced to date are discussed. This paper only considers contact handled TRU waste (i.e., contact dose rates < 2 mSv/hr). (authors)

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Germany's First Steps in the Restarted Site Selection Process - 18015

After the Fukushima event in 2011 Germany has decided to phase out nuclear power. Subsequently, it seemed to be politically attractive also to achieve a broad consensus with respect to final disposal of the high-active heat-generating waste after a long societal conflict lasting for decades. Germany has then decided to start a new site selection program for the determination of a site for high-active and heat-generating waste. In 2013, an overall parliamentary consensus could be achieved to agree on a new Repository Site Selection Act. This law has then been evaluated by a commission and the Repository Site Selection Act became amended. Besides, also organization and responsibilities became adapted as well as the financing system. A fund has been installed, which was financed by the utilities, to secure the availability of money to finance the whole process. In the meanwhile, the organizations for licensing of interim storage and final disposal facilities (BfE), for the operation of interim storage facilities for spent fuel, high-active waste and low and intermediate level waste (BGZ) and for site investigation, construction, operation and decommissioning of repositories for spent fuel and high-active waste as well as for low and intermediate level waste (BGE) had been founded and the site investigation process has just started. (authors)

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Pathways towards Valid Data: Assessment of Optimized Packaging Strategies for Core Components - 18158

Every company that runs nuclear power plants needs a strategy for disposal of the radioactive waste and for decommissioning and dismantling of its nuclear reactors. The dismantling progress of several German NPPs is already in an advanced stage. Hence, particularly highly activated core components have to be disposed of as radioactive waste. Nevertheless, this is also an issue during the operating phase due to necessary replacements of highly activated core components such as PWR control elements or RPV heads. The conditioning of waste packages for final disposal in the German repository Konrad requires the core components to be dismantled, packaged and dried. Therefore, a qualified packaging strategy is needed to ensure a reliable compliance with the protection objectives of the repository and an optimized utilization of prescribed contingents. Exemplified by PWR control elements a pathway is drawn how to perform an optimized packaging strategy in compliance with the protection objectives of the repository based on only few relevant radionuclides derived from activation calculations. For these calculations, particularly the influence of trace impurities in the materials has to be considered and their amounts have to be verified or analyzed by sampling. Furthermore, low nuclide activities from contamination cannot simply be neglected because some of them are obliged to be declared or their radiological relevance has not been sufficiently determined. Due to various influence factors, qualified sampling strategies need carefully considered concepts. Therefore, detailed planning facilitates the qualification and approval process and reduces the costs for analyses and the number of containers. (authors)

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Modeling the Long-Term Behavior of Cementitious Barriers - 18188

Cementitious materials are commonly included as backfill/barrier materials in geological repositories for radioactive waste. An understanding of the long-term behavior of cement-based barriers is therefore required to provide confidence in their safety functions within an environmental safety case. Geochemical modelling has been undertaken on several different radioactive waste repository concepts (and their natural/industrial analogues), to explore the nature of cement degradation over timescales relevant to postclosure safety (many tens of thousands of years to c. 1 Ma). In radionuclide transport calculations undertaken as part of performance assessments, it is typically considered that the properties of cementitious engineered barrier materials change as they degrade, potentially resulting in an increase in porosity hydraulic conductivity, and effective diffusion coefficients. However, reactive transport modelling (supported by some analogue data) suggests that, in some repository systems, a degree of clogging of the cement pore space could occur and that some advective flow paths (cracks in the barriers, fractures in the host rock) could become (to some extent) infilled by cement and rock alteration products. This could result in a reduction in water flow and contaminant transport in the vicinity of the cement-rock interface, but some uncertainties remain (especially process couplings and the nature of water flow in complex fracture networks that are becoming clogged). In addition, the armoring of fracture surfaces could result in decreased potential for diffusion of solutes into the rock matrix. It is envisaged that ongoing work will continue to improve reactive-transport model capabilities, such that the implications cement degradation has for contaminant transport can be predicted with greater confidence. (authors)

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THM-Coupled Modeling of Barrier Integrity for HLW Disposal in Bedded Salt - 18289

Within the framework of the R and D project KOSINA in Germany, numerical studies analyzing the long-term integrity of a bedded salt barrier under the thermal loading of heat generating radioactive waste have been performed using different modeling approaches for various repository concepts and two geological structures. The THM-coupled calculations show an initially steep increase of temperatures in the vicinity of the HLW canisters which reduces an uplift above the repository. This in turn leads to a temporally and spatially limited reduction of horizontal stresses at the top of the salt, allowing for pressure-driven percolation of groundwater into the upper part of the hydraulic barrier. However, at any given time an undisturbed salt barrier of over 300 m remains, shielding the repository from groundwater inflow. (authors)

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Buffer Material Degradation in High Level Radioactive Waste Disposal - 18454

The concept of geological disposal for high-level radioactive waste (HLW) is based on a multi-barrier system comprising the natural geological barrier that repository host rock provides and its surroundings, as well as an engineered barrier system (EBS) that includes waste form, waste canisters, buffer materials, and backfill. The multi-barrier system is expected to perform its desired functions to isolate waste from the biosphere and achieve HLW disposal. Because of the interaction processes of radionuclide decay heat with hydraulic and water-mineral reactions in the radioactive waste repository, the bentonite buffer degradation resulting from the external temperature and pressure warrants investigation. However, seldom addressed in Taiwan is the influence of thermo-hydro-chemical (T-H-C) processes on buffer material degradation in the EBS of HLW disposal repositories as related to smectite clay dehydration. Therefore, we adopted the chemical kinetic model of smectite dehydration to calculate the amount of water expelled from smectite clay minerals caused by higher temperatures of waste decay heat. We discovered that smectite clay could cause volume shrinkage because of a loss of interlayer water, which can lead to bentonite buffer compression. The results of this study may be used in advanced research on the evolution of bentonite degradation for both performance assessments and safety analyses of final HLW disposal in Taiwan. (authors)

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Challenges in Interim and Final Storage of HLW in Germany - 19060

After the Fukushima event in 2011 Germany has decided to phase out nuclear power. Subsequently, it seemed to be politically attractive also to achieve a broad consensus with respect to final disposal of the high-active heat-generating waste after a long societal conflict lasting for decades. Germany has then decided to start a new site selection program for the determination of a site for high-active and heat-generating waste. In 2013, an overall parliamentary consensus could be achieved to agree on a new Repository Site Selection Act. This law has then been evaluated by a commission and the Repository Site Selection Act became amended. Besides, also organization and responsibilities became adapted as well as the financing system. A fund has been installed, which was financed by the utilities, to secure the availability of money to finance the whole process. In the meanwhile, the organizations for licensing of interim storage and final disposal facilities (BfE), for the operation of interim storage facilities for spent fuel, high-active waste and low and intermediate level waste (BGZ) and for site investigation, construction, operation and decommissioning of repositories for spent fuel and high-active waste as well as for low and intermediate level waste (BGE) had been founded and the site investigation process has just started. (authors)

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Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel - 19133

Congress created the U.S. Nuclear Waste Technical Review Board (NWTRB) in the 1987 Nuclear Waste Policy Amendments Act (Public Law 100-203) to evaluate the technical and scientific validity of activities undertaken by the Secretary of Energy to implement the Nuclear Waste Policy Act. In accordance with this mandate, the NWTRB reviewed the U.S. Department of Energy's (DOE) efforts to manage the inventory of spent nuclear fuel (SNF) that is under its control (DOE SNF) at several facilities around the country: the Hanford Site in Washington, the Idaho National Laboratory (INL) in Idaho, the Savannah River Site (SRS) in South Carolina, and the Fort St. Vrain independent spent fuel storage installation (ISFSI) in Colorado. While disposal of SNF and high-level radioactive waste (HLW) in a deep geologic repository remains the ultimate objective of the DOE nuclear waste management program, there is significant uncertainty about when such a repository will be constructed in the United States. Until disposal occurs, it is essential to manage DOE SNF in a way that will facilitate its eventual disposal, and it is also important to improve understanding of processes related to packaging and storage of the SNF that could affect its future transportation and disposal. The NWTRB's review, which was a multi-year effort, was documented in a report to Congress and the Secretary of Energy and is summarized here. The NWTRB reviewed the quantities and characteristics of DOE SNF at each of the four sites. It also examined DOE packaging and storage activities and plans at each of the sites. The NWTRB highlighted three main issue areas in its review. First, the review identified issues related to managing the aging of DOE SNF and the facilities used to store the DOE SNF. Second, the review addressed issues related to packaging of stored non-naval DOE SNF into a standard canister. Third, the NWTRB assessed issues that could affect disposal of DOE SNF if DOE continues to evaluate a range of repository settings. Based on the information and findings developed as part of its review, the NWTRB made six recommendations to Congress and the Secretary of Energy related to aging management, packaging, and disposal of DOE SNF. DOE has begun to address some of the NWTRB recommendations. (author)

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Development and Demonstration of Monitoring Strategies and Technologies -Modern2020 Project - 19317

The Modern2020 project aims at providing the means for developing and implementing an effective and efficient repository operational monitoring program, taking into account the requirements of specific national programs. The work allows advanced national radioactive waste disposal programs to design monitoring systems suitable for deployment when repositories start operating in the next decade and supports less developed programs and other stakeholders by illustrating how the national context can be taken into account in designing dedicated monitoring programs tailored to their national needs. The work has been established to understand what should be monitored within the frame of the wider safety cases and to provide methodology on how monitoring information can be used to support decision-making and to plan for responding to monitoring results. Several developments on innovative repository monitoring techniques (wireless data transmission, alternative power supply sources, new sensors, geophysical methods) has been done from the proof of feasibility stage to the technology development and demonstration phase. Innovative technical solutions facilitate the integration and flexibility of required monitoring components to ease the final implementation and adaptation of the monitoring system. Full-scale in-situ demonstrations of innovative monitoring techniques were done in way to evaluate the performance of the state-of-the-art, the innovative techniques and their comparison with conventional ones. Finally, we will introduce the experiences of the local public stakeholders that have been involved in the Modern2020 project by focusing on the participation in practice of the Belgian, French, Finnish and Swedish local stakeholders, This paper provides an overview of the European Commission Horizon2020 Modern2020 project's main goals and progress to date. (authors)

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Study on Characterization of Colloidal Silica Grout for Rock Excavation Under Saline Groundwater - 19332

A rock excavation grouting technology has been recently studied as significant technology for reducing the ingress of water into the deep repository. However, it has not been studied for applying to the coastal region, where it is discussed to be a more suitable region for the geological disposal in Japan. The latest material called colloidal silica grout (CSG) is good for sealing narrow fractures but is known to be sensitive to the salinity of groundwater because of its gelling property with salt accelerator. Although the gelling of CSG can be controlled by adding an acidic pH adjuster, the methodology for delivering the appropriate grout is not well established for such conditions of saline groundwater. Therefore, this research project was established to enhance the existing rock grouting technology for deep repositories. Initially, a grouting workshop was held with Nordic authorities in order to understand the state of the art and identify the most significant challenges to developing a generic mixing method for injecting CSG under seawater. Based on laboratory tests, we developed a feasible mixing methodology that prevents the rapid reaction with the salty water, which shows that CSG is not negatively impacted by salty mixing water or submerged by saline water. Combining the latest grouting techniques from Japan and the Nordic countries creates an optimal grouting methodology for use under seawater. Using this hybrid approach, the CSG can find wide-scale use as an agent for grouting narrow rock fractures filled with saline or fossil water in order to decrease the ingress of water in the disposal facilities. The reduction of the water ingress will enhance the safety of the repository, reduce water discharge, and diminish the long-term cost of water treatment. (authors)

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A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

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Solidification of Low and Intermediate Level Waste - A Comparison between Different Methods - 19584

The treatment of radioactive materials results in products that have to be suitable for a period of interim storage as well as for disposal in a repository. Considering available options for treatment of radioactive waste, future disposal also has to be taken into account. Up to now, only cementation and encapsulation in geopolymer (only to limited extent) are relevant treatment methods for the solidification of German waste for disposal in the repository Grube Konrad. However, a synopsis of other solidification technologies is considered to be helpful in order to be prepared for future developments in the fields of radioactive waste treatment. Treatment methods addressed in this paper are: - Cementation; - Encapsulation into geopolymer matrix; - Vitrification; - SYNROC; - Geomelting. The criteria for evaluation of the treatment methods are as follows: - Requirements concerning waste properties; - Safety of treatment facilities; - Fulfillment of waste acceptance criteria for repository Grube Konrad; - Fulfillment of requirements for near surface disposal; - Suitability for long-term interim storage. (authors)

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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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Panel Session 3: Prospects for Shared Solutions for Solving the WM Challenges for Small and Emerging Nuclear Countries

This panel focused on sharing WM solutions for small and emerging nuclear countries. As countries with substantial inventories of spent nuclear fuel pursue their national repository programs with varying degrees of progress, countries with only a few reactors and those at the beginning stages of their nuclear programs face a far different and substantial challenge toward developing a national solution. This session brought together a panel of experts from government and industry to explore shared solutions and next steps building off the recent progress made by international organizations and the recent efforts and lessons learned. Panelists with presentations: The Rationale for a Multinational Repository (Sean Tyson); Perspectives on the Future of the Multi-National Repository Concept (Alan Brownstein); Why Would any Country Accept Foreign Spent Nuclear Fuel? (Charles McCombie); Dual Track Strategy for Small Countries (Leon Kegel)

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