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DOE OSTI2015

Robust disposal concept, uniform regulations, and trust in the messenger; three fundamental building blocks for consent-based HLW-disposal solutions in the USA - 15104

Pending enabling legislation, the USA's HLW-disposal program had been held in political abeyance since 2009 at the crossroads to one or both of the Yucca Mountain candidate (since 1987) HLW-repository in Nevada and/or the consent-based siting of the new HLW-disposal and centralized HLW-storage facilities proposed in 2012; a road not previously travelled. The consent-based siting approach would require the following gap between the reality (A) and the requirement (B) for public trust, acceptance and sustained support (PASS) to be bridged: A. Less than 1% of the general public in any given country fully comprehends the scientific-underpinnings and the projected health risks imposed by a HLW repository. B. More than 50% of the general public in a given geographical area would have to accept to host it from start to end. The following PASS-building blocks, successfully used at the WIPP-site in New Mexico and abroad to get from A to B during the past 30+ years, are thus highly recommended: 1. Defining the consent footprint upon host-county/state boundaries and/or radiation/health-risks and/or to identify/designate the directly affected parties (DAPs). 2. Using databases and lessons-learned by other repository programs to advantage. 3. Presenting the information in terms and/or by examples more readily understood by laypeople, because both their acceptance and opposition/rejection are influenced by: a. The type, number, problems, and status of 'similar' disposal concepts; and b. The level of i. Radiation protection provided by applicable regulations; ii. Involvement and decision-making authority vested in the DAPs; and iii. Their trust in the 'spokesperson/messenger. (authors)

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DOE OSTI2015

The role of independent monitoring in maintaining community support through a radiological incident - a WIPP case study - 15107

The Waste Isolation Pilot Plant (WIPP) is a U.S. Department of Energy (DOE) waste repository for the permanent disposal of defense transuranic (TRU) wastes. The repository is mined in the Salado Formation, a bedded salt formation of the Permian era, approximately 655 m (2150 ft.) below land surface. Located near Carlsbad, New Mexico, the WIPP facility is the world's first deep geological disposal facility licensed to accept TRU waste. Environmental monitoring is a key component in the development and operation of any nuclear facility. Well after the WIPP facility had been sited and constructed, but before repository operations began, DOE and local community leaders recognized the value of having an independent environmental monitoring program in addition to the self-monitoring program required by regulation. An independent monitoring organization, the Carlsbad Environmental Monitoring and Research Center (CEMRC) was created by New Mexico State University at its Carlsbad campus. After almost fifteen years of successful waste disposal operations, the first airborne radiation was accidentally released from the WIPP and detected beyond the site boundary on February 14, 2014. The accident released moderate levels of radioactivity into the underground air. A small but measurable amount of radioactivity also escaped to the surface and was detected beyond the site's inner boundary. The results were reported to the public as they were obtained and verified. Releases were low and localized, and no radiation-related health effects among local workers or the public would be expected. CEMRC's independence and its extensive monitoring program played a crucial role in timely informing the local populations on the levels and types of radiation released from the WIPP. CEMRC also helped eliminate unnecessary concern surrounding fear of radiation by engaging local populations through news releases and presentations at 'town hall' meetings sponsored by the Mayor's Office and the Department of Energy. (authors)

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DOE OSTI2015

Risk considerations for WIPP passive institutional controls - 15173

The regulatory assurance measures for the Waste Isolation Pilot Plant (WIPP) include such things as active and passive institutional controls, monitoring, waste retrieval/removal capabilities and engineered barriers. Passive Institutional Controls (PICs) are an assurance measure intended to reduce the likelihood of humans unintentionally intruding the WIPP geologic radioactive waste disposal system for thousands of years after it is closed. The waste repository is designed such that it should not pose a significant risk to future populations if it is left undisturbed. The design has also been demonstrated through a performance assessment (PA) analysis to not pose a significant risk to future generations even if it is intruded upon by an inadvertent drilling intrusion [1, Appendix PA]. Through PA, it has been determined that the overall risk of the disposal system to people and the environment is small. The intent of the regulatory Assurance Requirements and PICs is to provide additional confidence that the repository will perform as expected after closure. However, the actual implementation of the proposed PICs design may pose a greater risk to the current population than they would potentially reduce for future generations. Specifically, the risks due to industrial accidents, construction and transportation risks incurred during the fabrication and construction of the permanent markers system may be greater than the overall reduction in risk to future generations due to the existence of the permanent markers. If this were true, implementation of the current PICs design may be counter to the intention of the regulatory requirement to include such assurance measures. As such, provisions for assurance measures used in geologic radioactive waste disposal facilities, such as PICs, should consider risk impacts. A provision of any disposal program's Regulatory Assurance Requirements should address the impacts of the known risks from all the components of the assurance-related systems. The initial WIPP PICs design was developed as part of the regulatory certification process in 1996 with the intent of being revised when the WIPP closes several decades from now [2, Appendix PIC]. The revised design that will be developed at closure should address the design's potential reduction to long-term risks from the repository and the risks due to the implementation of the final design. (authors)

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DOE OSTI2015

Knowledge management information tool Web analytics - 15182

The Knowledge Management Information Tool (KM-IT) is a web-based [01] system developed to maintain and preserve the Department of Energy's (DOE's) D and D knowledge base. The system was developed by Florida International University's Applied Research Center (FIU-ARC) with the support of the D and D community including the DOE Office of Environmental Management, and with the active collaboration and support of the DOE's Energy Facility Contractors Group (EFCOG). The KM-IT system is a D and D community driven system tailored to serve the technical issues faced by the D and D workforce across the DOE Complex. The D and D KM-IT is an interactive web-based system operational and available for free access at www.dndkm.org. [02] The D and D KM-IT is currently composed of the following tailored D and D modules: Web Crawler, Hotline, Technology, Document Library (including Hanford's and Savannah River Sites' ALARA Center Reports), Specialist Directory, Lessons Learned, Best Practices, Video/Picture Library, Vendors, Collaboration Tools, Training, and Mobile Systems. The Web Crawler dynamically searches through the D and D KM-IT repository as well as the web and displays search results based on the search criteria. The Hotline allows registered users to post questions/problems related to D and D and to receive solutions from a subject-matter specialist. The Technology module serves as repository of D and D technologies and contains technology/ demonstration fact sheets and vendor information. D and D community members can add Lessons Learned, Best Practices, Documents, and Pictures/Videos to the D and D KM-IT repository, after a formal approval process. The Vendor module provides a directory of commercial vendors who provide D and D related technologies, supplies, and services. D and D KM-IT makes excellent use of the knowledge that exists within the D and D community by allowing D and D project managers around the DOE complex to collaborate by sharing innovative ideas, past experiences, and practices and by maintaining a directory of subject matter specialists. The mobile version of the KM-IT can be accessed from m.dndkm.org [03] and currently has six active modules: Vendor, Technology, Picture/Video Library, Hotline, Lessons Learned and Specialist Directory. Web analytics is a key tool being implemented by the D and D KM-IT project team in an effort to enhance the system to better meet users' needs. The D and D KM-IT project team incorporates web analytics as part of its operational elements to analyze information and key data points on how the system is being used. Web analytics is the measurement, collection, analysis and reporting of internet data for the purposes of understanding and optimizing web usage. The use of web analytics retains or attracts D and D community members to access and share knowledge through virtual collaboration on the KM-IT platform. D and D KM-IT web analytics measures details on the overall usage of the system without tracking personal information or specific user behavior: how many users visited the application, how many of those users were unique visitors, how visitors come to the system (i.e., by following a link to the system or by using the web site URL via bookmark or directly typing the address), what keywords were used to perform searches with the search engine, how long the average user stayed on a given screen or on the entire system and what links were clicked on. This information is used to identify which areas of the system are more popular than the others and provides information that can be used to streamline the application for a better user experience. (authors)

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DOE OSTI2015

Influences of temperature on permeability changes of flow-paths altered by highly alkaline Ca-rich groundwater - 15192

Fractures (or cracks) in host rock around the geological repository are considered as main flow-paths for the migration of radionuclides. Such fractures would be altered by the alkaline components (Ca, Na, K) leaching from cementitious materials for the construction of the repository. While silicate minerals in fractures are dissolved by highly alkaline groundwater, secondary minerals would deposit on the surface of fractures because of the reaction between the alkaline components and silicate species. The authors' previous study showed the generation of CSH-gel and the clogging of micro flow-path under the condition of Ca-rich and high pH by flow experiments. The present study examined the permeability changes with temperature in micro flow-path flowing highly alkaline Ca-rich solution, considering temperature at the depth of geological disposal system. For the simulation of micro flow-path such as fractures, 'micro flow-cell', which nipped a thin Teflon sheet with a slit between granite chip and Teflon basement, were used for the experiments. Highly alkaline Ca-rich solution (pH>12 and 8.5 mM Ca(OH){sub 2}) were injected into micro flow-cell with 2.0 ml/h of flow rate under the given temperature conditions (278 K, 296 K or 313 K). As a result, the permeability became lower with the increase in temperature. In addition, the deposition rates of CSH-gel became larger with the increase in temperature, and the activation energy was estimated to be 32.0 kJ/mol. These results suggest that flow-paths around the geological repository would undergo the clogging with CSH-gel in a shorter time-period under a condition of higher temperature. (authors)

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DOE OSTI2015

Effects of pH on deposition rate of supersaturated silicic acid around geological disposal system - 15245

For the migration assessment of radionuclides around geological repository, it is important to estimate the dynamic behavior of silicic acid under a high alkalinity condition due to the cementitious materials for the construction of the repository. This study examined the pH dependencies of the deposition behavior of supersaturated silicic acid in the co-presence of solid phase. The parameters for the deposition experiments were as follow: the pH was set to 8, 9 and 10, the initial saturated concentration was set to 2-8 mM, the amount of amorphous silica (350 m{sup 2}/g) as solid phase was set to 0.1-1.0 g. The experimental results showed that the initial supersaturated concentrations strongly affected the formation of colloidal silicic acid, increasing the deposition amount of silicic acid and decreasing the soluble silicic acid with time. In other words, the polymerization and the deposition of silicic acid simultaneously arose in these experiments. Furthermore, this study evaluated the apparent deposition rates of silicic acid by assuming the first-order reaction equation considering the specific surface area. As a result, the deposition rate constants, k, were in a narrow range of 10{sup -11} to 10{sup -10} m/s under all experimental conditions in this study. This means that the deposition rate constants do not depend both on pH-value (in the range from 8 to 10) and the initial concentration of supersaturated silicic acid. That is, the estimated deposition rate constants are useful in estimation of the deposition behavior of supersaturated silicic acid, such as the clogging in fractures (flow-paths) around the repository, even if the pH-value of groundwater gradually or dramatically changes. (authors)

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DOE OSTI2015

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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DOE OSTI2015

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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DOE OSTI2016

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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DOE OSTI2016

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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DOE OSTI2016

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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DOE OSTI2018

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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DOE OSTI2018

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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DOE OSTI2018

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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DOE OSTI2018

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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DOE OSTI2018

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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DOE OSTI2019

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

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DOE OSTI2019

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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