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At least 1,351 records · Page 75

Fuel Matrix Degradation Model Development Update: Alloy Corrosion Rates and Hydrogen Generation

This work is being performed as part of the DOE NE Spent Fuel and Waste Science and Technology Campaign, Argillite and Crystalline Rock R&D work packages: SF-19AN01030101 and SF-19AN01030201. This document meets the milestone M4SF-19AN010301013 for Argillite R&D and the milestone M4SF-19AN010302013 for Crystalline R&D. The fuel matrix degradation (FMD) model calculates the degradation rate of spent UO2 fuel based on fundamental electrochemical theory and thermodynamics. The model provides the quantitative basis for radionuclide source term predictions in the generic disposal system analysis (GDSA) performance assessment (PA) model. The FMD model has been implemented in a manner that facilitates its integration with the GDSA-PA code and a preliminary, integrated FMD-GDSA model has been successfully tested (Jerden et al., 2017). The specific focus of on-going work is to accurately and quantitatively represent the generation of H2 in a breached waste package and its effect on the spent fuel degradation rate so that this key process can be accurately represented in PA models. It has been shown experimentally that millimolar concentrations of dissolved H2 in contact with spent fuel will inhibit oxidative dissolution and decrease the fuel degradation rates by as much as three orders of magnitude (e.g., Röllin et al., 2001, Ollila, 2008). This will lead to a significant decrease in the radionuclide source term values used in repository PA models. Sensitivity studies using the FMD model have shown that the dissolved H2 concentration is the dominant environmental variable affecting the UO2 spent fuel dissolution rate as long as a source of H2 persists in the breached waste package (Jerden et al., 2015). Therefore, on-going experimental and modeling efforts focus on H2 production, accumulation and interfacial reactions within a breached waste package. The anoxic corrosion of metallic engineering materials within a breached waste package will be the main source of H2 in crystalline and argillite rock repository systems. The most important H2-producing alloys present within a waste package include stainless steel, carbon steel, and aluminum alloys and the Zircaloy fuel cladding. A steel corrosion module was added to the FMD model to account for these H2 sources and couple the H2 generation rate with fuel degradation processes (Jerden et al., 2017). As discussed below, there is a need for experimental data from electrochemical corrosion tests to quantify the H2 generation rates due to the corrosion of these waste package alloys as a function of key environmental variables (Eh, pH, groundwater chemistry). This report presents results from series of electrochemical corrosion tests with materials present in waste packages that provide information needed to parameterize and validate the FMD model. The alloys tested include 316L stainless steel (316SS), AISI 4320 carbon steel (C-steel) and Zircaloy-4 fuel cladding (unirradiated). Potentiodynamic polarization and potentiostatic polarization tests were performed in buffer solutions at pH 4, 7 and 10 with NaCl concentrations of 0, 4 and 100 millimolar (mM). The potentiostatic tests were performed at voltages varying from 0.5 volts down to -0.66 volts, vs the standard hydrogen electrode (SHE), and all tests were performed at laboratory ambient temperatures (~22 °C). The corrosion rates measured during the electrochemical tests were used as direct input to a new FMD/in-package chemistry model that is being developed to provide accurate long-term spent fuel degradation rates as chemical conditions (Eh, pH, speciation) evolve within a breached waste package. This new model was developed by coupling the FMD model with the reactive transport code X1t, which is a module within the Geochemist’s Workbench (GWB) reaction path modeling code (Jerden et al., 2019). The reactive transport model consists of a 1D domain discretized with 21 reaction diffusion cells and includes a single waste packag

Jerden, James↗

Hybrid User Manual

This document is the user manual for the Hybrid Repository that is a part of the Integrated Energy Systems program. The Hybrid Repository is a collection of physical modelica models used for the creation of transient integrated energy systems.

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

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

Jove-Colon, Carlos↗

Standardized Architecture for a Mega-Biobank Phenomic Library: The Million Veteran Program (MVP)

Electronic health records (EHRs) provide a wealth of data for phenotype development in population health studies, and researchers invest considerable time to curate data elements and validate disease definitions. The ability to reproduce well-defined phenotypes increases data quality, comparability of results and expedites research. In this paper, we present a standardized approach to organize and capture phenotype definitions, resulting in the creation of an open, online repository of phenotypes. This resource captures phenotype development, provenance and process from the Million Veteran Program, a national mega-biobank embedded in the Veterans Health Administration (VHA). To ensure that the repository is searchable, extendable, and sustainable, it is necessary to develop both a proper digital catalog architecture and underlying metadata infrastructure to enable effective management of the data fields required to define each phenotype. Our methods provide a resource for VHA investigators and a roadmap for researchers interested in standardizing their phenotype definitions to increase portability.

Knight, Kathryn↗

Validation of MCNP Critical Benchmarks Models of Highly Enriched Uranium Cylinders

A new centralized repository of high-quality MCNP models of critical benchmark experiments is currently under development at Los Alamos National Laboratory (LANL). The benchmark experiments are described in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and the initial set of benchmark models are derived from the Whisper Suite provided with MCNP6.2. This effort is a collaboration among the Nuclear Criticality Safety, Nuclear Data, and Monte Carlo code development/application organization at LANL. The objective is to create a current single LANL benchmark collection that includes the latest ICSBEP revision that has a formal review and revision process, is contained in an open- source repository, and utilizes new Python tools for improved input and output file review.

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Industry Facing PV Degradation Prediction Tool and Database to Enable a 50-Year Life Module

The Photovoltaic (PV) industry constantly aims for lower costs, higher-efficiency cells, and improved module designs. These trends lead to using new materials, designs, and manufacturing processes, resulting in a continually changing technological landscape. These changes can potentially introduce new, unknown degradation mechanisms and failure modes that are difficult to diagnose, analyze, test, and model. This introduces uncertainty into the expected lifetime of PV modules of 25 years. Furthermore, research efforts aim for up to 50 years of service life while keeping performance degradation at a minimum for decades of outdoor weathering - putting additional pressure on improving the accuracy of long-term durability and reliability assessments. Here there is a need to organize the existing degradation data into an accessible format and to provide industry relevant tools for extrapolation from laboratory to field conditions. Because the core of this type of analysis involves calculations that are complicated but ubiquitous for many degradation processes, an enhanced predictive modeling framework will facilitate the analysis to help researchers keep up with the rapid pace of technological changes. In this work, we present an online tool that can be used to search for and analyze degradation information and extrapolate PV module performance and durability to field exposure. A graphical user interface will aid in the understanding of the results. The prediction tool will be built modular and published as open source, enabling users to expand on the existing framework. We use an integration pipeline approach that allows us to leverage weather data from the National Solar Radiation Database to perform geospatial degradation analysis in the US and worldwide. Our repository will contain various degradation models and material parameters suitable for the reliability and durability assessment of materials and components deployed outdoors. We hope to become a repository that can be used for weathering and degradation analysis for various applications beyond the PV industry.

degradation↗

The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generally applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)

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A Proposal for Geologic Radioactive Waste Disposal Environmental Zero-State and Subsequent Monitoring Definition - First Lessons Learned from the French Environment Observatory - 13188

Obtaining a reference state of the environment before the beginning of construction work for a geological repository is essential as it will be useful for further monitoring during operations and beyond, thus keeping a memory of the original environmental state. The area and the compartments of the biosphere to be observed and monitored as well as the choice of the markers (e.g. bio-markers, biodiversity, quality of the environment, etc.) to be followed must be carefully selected. In parallel, the choice and selection of the environmental monitoring systems (i.e. scientific and technical criteria, social requirements) will be of paramount importance for the evaluation of the perturbations that could be induced during the operational phase of the repository exploitation. This paper presents learning points of the French environment observatory located in the Meuse/Haute-Marne that has been selected for studying the feasibility of the underground disposal of high level wastes in France. (authors)

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Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

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The Creation of a French Basic Nuclear Installation - Description of the Regulatory Process - 13293

CEA is a French government-funded technological research organization. It has to build a medium-level waste interim storage facility because the geological repository will not be available until 2025. This interim storage facility, called DIADEM, has to be available in 2017. These wastes are coming from the research facilities for spent fuel reprocessing and the dismantling of the most radioactive parts of nuclear facilities. The CEA handles the waste management by inventorying the needs and updating them regularly. The conception of the facility is mainly based on this inventory. It provides quantity and characteristics of wastes and it gives the production schedule until 2035. Beyond mass and volume, main characteristics of these radioactive wastes are chemical nature, radioisotopes, radioactivity, radiation dose, the heat emitted, corrosive or explosive gas production, etc. These characteristics provide information to study the repository safety. DIADEM mainly consists of a concrete cell, isolated from the outside, wherein stainless steel welded containers are stored, stacked in a vertical position in the racks. DIADEM is scheduled to store three types of 8 mm-thick, stainless steel cylindrical containers with an outside diameter 498 mm and height from 620 to 2120 mm. DIADEM will be a basic nuclear installation (INB in French) because of overall activity of radioactive substances stored. The creation of a French basic nuclear installation is subject to authorization according to the French law No. 2006-686 of 13 June 2006 on Transparency and Security in the Nuclear Field. The authorization takes into account the technical and financial capacities of the licensee which must allow him to conduct his project in compliance with these interests, especially to cover the costs of decommissioning the installation and conduct remediation work, and to monitor and maintain its location site or, for radioactive waste disposal installations, to cover the definitive shut-down, maintenance and surveillance expenditure. The authorization is issued by a decree adopted upon advice of the French Nuclear Safety Authority and after a public enquiry. In accordance with Decree No. 2007-1557 of November 2, 2007, the application is filed with the ministries responsible for nuclear safety and the Nuclear Safety Authority. It consists of twelve files and four records information. The favorable opinion of the Nuclear Safety Authority on the folder is required to start the public inquiry. Once the public inquiry is completed, the building permit is issued by the prefect. (authors)

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Preliminary Safety Analysis of the Gorleben Site: Overview - 13298

The project preliminary safety analysis of the Gorleben site started in 2010 and is based on the safety requirements for heat generating radioactive waste released from the German Federal Ministry for Environment, natural conservation and nuclear safety. The project consists of several tasks: the database defining the geology of Gorleben and the composition of the waste to be disposed of, the safety and demonstration concept, the repository concepts, the scenario analysis, the system analysis with long-term safety assessment and the synthesis. The overall synthesis indicates presently the compatibility of a repository in Gorleben with the safety requirements. The application of the method for a site selection process is still under evaluation. (authors)

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Preliminary Safety Analysis of the Gorleben Site: Geological Database - 13300

The Gorleben salt dome is 4 km wide and nearly 15 km long. It is composed of different salt rock types of the Zechstein (Upper Permian) series and extends to the Zechstein basis in a depth of more than 3 km. In the course of the salt dome formation the salt was moved several kilometers. During the uplift of the salt the initially plane-bedded strata of the Zechstein series were extensively folded. In this process anhydrite as a competent layer was broken to isolated blocks. In the core of the salt dome the Hauptsalz, which is characterized by a particularly high creeping capacity, forms a homogeneous halite body with a volume of several cubic kilometres. The Hauptsalz contains gaseous and liquid hydrocarbons in separated zones of decimeter to meter dimensions. The overall hydrocarbon content is far below 0.01 %. At the flanks the salt dome consists of salt rocks with lower creeping capacities. Brine reservoirs with fluid volumes in the range of liters to hundreds of cubic meters exist in certain regions of this part of the salt dome. The water content of the Hauptsalz is below 0.02 %. Interconnected pores do not exist in the salt rock outside of fluid bearing or fractured areas, i.e. the salt rock is impermeable. The exploration of the Gorleben site as a potential site for a HLW-repository started in 1979 and is still in progress. To date no scientific findings contest the suitability of the site for a safe HLW-repository. (authors)

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Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

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Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518

As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

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Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for hazardous materials transport. The paper concludes that while the HM 164 regime is sufficient for certain applications, it does not provide an adequate basis for a national plan to ship spent nuclear fuel and high-level radioactive waste to centralized storage and disposal facilities over a period of 30 to 50 years. (authors)

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The Polymers for Liquid Radioactive Waste Solidification: a Lost Chapter in the History of Engineering or a Step Forward? - 13529

Ideas on the application of polymers for the liquid radioactive waste immobilization go a way back, and the first studies in the area were published 30-40 years ago. One should admit that regardless of the fairly large number of publications appeared in the past years currently the interest in this work came down greatly. It was the successful assimilation and worldwide implementation of the LRW cementation technology caused a slump in the interest in polymers. But today it's safe to say that the situation slowly changes, particularly due to the market appearance of the high-tech polymers manufactured by Nochar Company, and unique properties of these polymers gradually raise the demand in various countries. The results of multiple experiments performed with the simulated solutions have passed the comprehensive tests with actual waste. The economic effect from the implementation of the new technology is defined by the volume reduction of waste coming onto the repository, by the decline in the cost of transportation and of the repository construction on account of cutting down the construction volume. Interesting results have been obtained during the search for the technical decisions that would allow using the polymer materials in the processing technology of the industrial toxic waste. One more promising area of the possible application of polymers should be pointed out. It is the application of polymer materials as the assets for the emergency damage control when the advantages of the polymers become obvious. (authors)

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Are We Serious in the US about the Disposal of HLW? - 13561

Since all efforts to date to dispose of HLW in the US have been unsuccessful, the following specific actions need to be taken if we are serious about such disposal: - The requirement in the EPA environmental radiation protection standards to predict the behavior of these unwanted residuals for one million years is meaningless. The Standards must be revisited. - Characterize two sites. There are myriad ways a site can be found to be unacceptable. Additionally, the existing HLW inventory requires a second repository. - Congress should specify incentives to states under consideration for a site. Perhaps 5% of total cost would be appropriate. - An independent technical review group should be established in such states to evaluate a proposed repository similar to the New Mexico Environmental Evaluation Group (EEG) for the WIPP Project because the state's interests are not necessarily the same as DOE's. - Acceptance or rejection of a proposed site should be based on technical issues, not social ones. Professionals in this field should present papers identifying the merits of HLW disposal in their own state. The scarcity of such research suggests Not In My Back Yard (NIMBY) syndrome. - Medical diagnostic ionizing radiation exposure to the US public is now 8,000 times greater than radiation exposure from nuclear energy. People accept this believing the benefits outweigh any risks. A major effort needs to focus on both benefits as well as risks of radioactive waste disposal. - DOE needs to announce preferences of host rock formations, incentives for states, and potential consequences should we fail to act. (author)

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Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗