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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Deep Isolation - Innovation for the Storage and Disposal of High-level Waste and Spent Nuclear Fuel - 19367

Deep Isolation, Inc. is a private, Berkeley based company that has brought innovation to nuclear waste disposal. Deep Isolation has developed a disposal technique that adapts mature drilling technology and customizes it for the purpose of providing a safe and secure solution. Though the stalled nuclear waste program in the US has created unwanted burdens and costs, it has also provided object lessons which Deep Isolation has learned from and designed a novel approach. The solution proposed by Deep Isolation combines the best of the geosciences and social sciences to present the US and other countries with an optimized and tailored approach to disposing of a variety of nuclear waste. This system can complement existing potential storage and disposal sites, or in the case of countries with smaller amounts of waste--the national solution. Deep Isolation innovates where others have stalled and offers a means of deep geological disposal that costs less, takes less time to implement, and cuts down and in some cases eliminates the need for transportation of the waste. The technical solution optimized by Deep Isolation uses established directional drilling technology from the oil and gas industry to drill a vertical drill hole thousands of meters deep and then transition to a horizontal drill hole that is thousands of meters in length. The target media is in or below shale formations, which are present at various depths throughout much of the United States and throughout the world. The availability of suitable geology allows for permanent disposal at multiple locations - including existing nuclear sites, without a need to transport the waste to a central repository. This innovative solution not only offers options for commercial used nuclear fuel and defense spent nuclear fuel and high-level waste, but also other nuclear wastes such as those considered by DoE's past deep bore-hole program, i.e. cesium and strontium capsules. Greater-Than-Class-C (GTCC) waste is also an opportunity for Deep Isolation. DoE's preferred alternative for GTCC includes 'land disposal at generic commercial facilities.' Deep Isolation could provide safe, secure, and permanent disposal at a commercial facility. Deep Isolation complements the technical innovation with best practices from the social sciences. In acknowledgement that the past challenges to implement nuclear waste disposal facilities have been primarily due to socio-political failings, we have designed a means to deeply engage with public and stakeholders. Our approach is committed to partnering with communities, local jurisdictions, states, and Tribes to determine a path forward that benefits all involved, that is fully transparent, collaborative and accountable. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The Role of Indian Tribes in the Safe, Secure, and Routine Transportation of Spent Nuclear Fuel and High-Level Radioactive Waste - 19463

Far too often federally recognized Indian Tribes are the last stakeholders invited to the table to discuss a particular issue with potentially wide-ranging impacts. Indian tribes were the first governments on this continent. With respect to the future transportation of spent nuclear fuel (SNF) or high-level radioactive waste (HLW), Indian tribes are potentially impacted in a myriad of ways: economics, cultural, environmental, health and safety. Yet few regulators fully understand and appreciate the legal and political standing of Indian tribes. This paper will examine the ways in which Indian tribes must be involved with the planning for safe, secure and routine SNF and HLW shipments and the ways in which the federal government is obligated to involve tribes. Indian tribes are similar to states in that they are responsible for the health and safety of its members (which they themselves define). Tribes, like states, make, implement, interpret, and enforce laws, and provide public services - healthcare, public education, social services, drug treatment, public safety and emergency management, transportation, managing natural resources, public housing, and environmental and cultural preservation. Many tribes also operate economic enterprises - manufacturing, farming, forestry, and gaming. The most basic principle of Indian law is the principle of 'sovereignty,' that is, an Indian tribe possesses all the powers of any sovereign state. Tribal governments are not subject to state or local jurisdiction but are subject to federal jurisdiction. The tribes enjoy a government-to-government relationship (and expectation) with the federal government (i.e., agencies, like the Department of Energy, the Nuclear Regulatory Commission). Of paramount concern to Indian tribes is the doctrine of Trust Responsibility, the general principle that the federal government's conduct 'in dealing with the Indians should therefore be judged by the most exacting fiduciary standards.' With respect to SNF/HLW this means that the federal government has an obligation (or a Trust Responsibility) to ensure that tribal people, lands and resources are protected. Equally important is the expectation that federal agencies will work with the Indian tribes on a government-to-government basis before undertaking any action with potential negative consequences on the tribe's people, land or resources. The concept of government-to-government consultation between Indian tribes and the United States is long standing and well recognized. As expressed in Executive Order 13175, 'Consultation and Coordination with Indian Tribal Governments': 'Consultation - Each federal agency must have an accountable process to ensure meaningful and timely input by tribal officials in the development of regulatory policies that have tribal implications.' The purpose of consultation is to ensure that Indian tribes are consulted before government officials make decisions that may have an impact on their lives and territories. The three agencies involved with SNF and HLW transportation - the US Department of Transportation (DOT), the US Nuclear Regulatory Commission (NRC), and the US Department of Energy (DOE) - have had various degrees of success with respect to direct consultation with Indian tribes. Given that there is plenty of lead-time before the first shipments commence, there should be ample time to develop a strategy for ensuring that tribes are equal partners when planning and preparing for shipments of SNF and HLW to the first interim storage facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Spent Nuclear Fuel Exploratory Roadmaps - 20054

The U.S. Department of Energy manages nearly 2,500 metric tons of heavy metal (MTHM) of spent nuclear fuel (SNF) resulting from several decades of research, testing, and production reactors [DOE 1995]. This SNF is managed at the DOE Hanford Site in Washington State (Hanford), Idaho National Laboratory (INL), and Savannah River Site in South Carolina (SRS). From 1995 to 2004, the Department of Energy (DOE) made several key programmatic decisions, supported by appropriate documentation in accordance with the National Environmental Policy Act (NEPA) for the management of SNF [DOE 1995]. These decisions have provided an overarching framework for SNF management, as well as site-specific and SNF-related management decisions for the past two decades. In the years since these decisions were made, with the notable exception of the successful drying and packaging of the production reactor fuel at Hanford, a majority of the decisions have been largely unimplemented. Also, since these decisions were made, a number of changes that bear on considerations relative to the path forward have occurred. A SNF Exploratory Road map activity, identified reasonable alternate pathways for DoE's inventory of SNF. Three fundamental pathways were identified in this activity for the long-term management and disposition of DOE SNF: Direct Disposal Pathway, Existing Processing Pathway and Alternate Processing Pathway. Further actions would be required (e.g., evaluation of NEPA analysis, technology maturity evaluations, refinement of cost and risks, benefits and advantages relative to other alternate paths) to define and inform many key decisions that will result in the selection of the disposition pathways that are most beneficial to the US Government in dispositioning DoE's SNF inventory. DOE is currently managing all SNF safely; however, the age of DOE-owned SNF and facilities for storing and processing SNF, coupled with the uncertainty of the storage duration, necessitates decisions and actions to ensure that the infrastructure will be in place to ensure continued safe and effective long-term management and eventual disposal. The identification of reasonable alternate pathways took these considerations into account in an effort to proactively manage conditions that could challenge the safety of storing and managing DOE SNF over the time periods now contemplated and also to ensure that flexibility is preserved to ensure that the DOE SNF remains compatible with final disposition pathways when they become available. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Criticality Consideration for Geologic Disposal of DOE-Owned Spent Nuclear Fuels - 20255

Decisions must be made relative to packaging spent nuclear fuel (SNF) for storage, transportation, and eventual disposal. This paper evaluates potential criticality-related loading and packaging considerations for Department of Energy (DOE)-owned SNF that could be impacted favorably or unfavorably by disposal uncertainties and proposes approaches to minimize such impacts. Using Al-Based SNF as an example and drawing on analyses performed in support of the Yucca Mountain Repository License Application (LA), this paper explores the role of natural barriers, engineered barriers, and waste form characteristics relative to reducing the potential for postclosure criticality. Several parameters important to postclosure criticality are considered in order to determine which are useful discriminators to support packaging decisions. Using these parameters, a qualitative assessment is performed relative to three options for managing DOE-owned SNFs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Overview of Salt Repository Research and Development for Spent Nuclear Fuel and High-level Nuclear Waste in the United States - 20307

This paper summarizes the current US Department of Energy Office of Nuclear Energy's (DOE-NE) work towards developing and executing a research and development program that addresses both scientific and technical issues related to long-term disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a hypothetical bedded salt based geological repository. A primary goal of the program is to create a generic Geologic Disposal Safety Assessment (GDSA) Framework that can be used to help guide decisions on siting a possible future bedded salt repository. The generic GDSA work includes analysis of the impacts of heat generation caused by decay of short-lived radionuclides. We report progress in four primary areas. First, we discuss the development and recent modifications of a research and development road-map. Second, we briefly describe an experimental approach to better understand thermal processes in salt. Third, we highlight collaborations with the international research community that leverage salt-based repository science around the world. Finally, we discuss how our findings are being used to aid in the development of a generic safety assessment for a bedded salt repository containing SNF and HLW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Technology Development for Dry Storage of Aluminum-Clad Spent Nuclear Fuel - 20490

A candidate disposition pathway for the > 13 MTHM of aluminum-clad spent nuclear fuel (ASNF), owned and managed by the U.S. Department of Energy, is the drying and placement of the SNF into sealed-canister dry storage, with the ASNF-in- canisters 'road-ready' for transportation to and final direct disposal in a repository waste package. Technical information gaps in fuel drying, and fuel dry storage behavior, have previously challenged the declaration of technology readiness for drying and placement of this fuel into the DOE Standard Canister design for > 50 years of safe dry storage. The principal technical information gaps included: i) characterization and thermal dehydration behavior of aluminum (oxy)hydroxide films attendant on the cladding due to film formation during reactor operation and during post-discharge up to long-term wet storage histories; and ii) G-value data to enable estimation of the radiolytic gas generation from the cladding with its (oxy)hydroxide films. Thus, the oxide films on the ASNF challenged the safety of a sealed storage canister with thermal and radiolytic decomposition of the waters on the films that can lead to corrosion, pressurization, and flammability issues. These gaps in the technical information base have largely been closed. This paper discusses the investigations at the Idaho National Laboratory (INL) and the Savannah River National laboratory (SRNL), and outlines the pending technology development work for input to an engineering design to enable a road-ready dry storage system for ASNF. The ASNF inventory considered for road-ready dry storage is stored at the Savannah River Site (SRS) and at the Idaho Nuclear Technology and Engineering Center (INTEC) at the INL. The ASNF inventory in the SRS L Basin is from foreign and domestic research reactors (FRR and DRR), and is diverse in terms of design, irradiation, and post-reactor-discharge storage conditions; these factors yield a range of characteristics of cladding oxide films on ASNF. Mixed aluminum (oxy)hydroxide (boehmite and bayerite/gibbsite) films, non-uniform in thicknesses up to a maximum local thickness bounded by 25 μm, were observed on ASNF materials removed from wet storage in the L Basin and in non-sealed dry storage at the SRS following reactor service and a long-term (up to 40+ years) interim storage history. The ASNF inventory at INTEC, in both wet and vented dry storage, is predominantly from the Advanced Test Reactor (ATR), but it also includes DRR and FRR fuel. To address a profound behavior of these films, radiolysis testing of aluminum specimens with mixed type boehmite/bayerite oxide films was performed to develop basic data on G-values for production of radiolytic hydrogen under dry storage conditions with nominal relative humidity, temperature, and cover gas. Modeling and simulation of canister internal environments with postulated inventories of oxide films provides estimation of the evolution of the conditions of the canister loaded with ASNF. Simultaneous Thermal Analysis (TGA/DSC) of hydrated oxide powders, and laboratory-scale aluminum specimens with a bayerite film (∼10 μm) using TGA methods, inform drying time/temperature conditions to be used for the ASNF. A demonstration project is recommended for Verification and Validation of the drying and storage of the ASNF. Remaining major tasks leading up to the hot demonstration include scale-up radiolysis testing and scale-up drying testing. Engineering design with the information from the technology program will establish the safety basis and enable long-term (> 50 years) dry storage compatibility with ASNF in the DOE Standard Canister pending its transportation to and disposal in a repository. This full capability would show the ASNF-in-canister storage system to be road-ready. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Spent Nuclear Fuel Package Consolidation using Metallic Void Fillers

This study introduces the idea of a consolidated spent nuclear fuel (SNF) package, formed by casting a metal into a SNF canister to fill any accessible void spaces. It is well known that castings experience shrinkage during cooling resulting in air gaps between the filler and its boundaries. For a consolidated SNF package, the air gap formed between the metallic filler and package components would have a negative impact on the thermal effusivity benefit offered by the filler. Here we present one solution to this challenge based on thermodynamic/solidification and metallic bonding principles. The key idea here is to pre-coat the package components with a low melting point metal, and to oxidize the coating instead of the components during the pre-heating step of the casting process. This allows for wetting of the boundary with the filler metal without the use of a flux that wouldn't survive the preheat temperature and could also contaminate the boundary with corrosive compounds. Potential benefits of successfully adopting this idea are twofold. First, improved heat diffusion inside the package due to the absence of the gap, which is significant for conceptual SNF with higher decay heat and high peak to average ratio. Second, improved proliferation resistance by making the fissile material recovery process more complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Accelerated Corrosion Tests to Evaluate Long-Term Performance of BORAL{sup R} in Spent Fuel Pools

Neutron absorber materials are used to increase storage capacity in spent fuel pools (SFPs) while maintaining criticality safety margins. BORAL{sup R} - a neutron absorber material commonly used in SFPs - is a metal matrix composite containing boron carbide (B{sub 4}C) and Al-1100 mixture. The aluminum (Al) cladding, which is on both sides of the core, serves as a protective barrier. Al cladding is susceptible to corrosion, which can lead to blisters between the core material and the cladding over time. Based on operational experience to date, pits also have been observed via visual examination and microscopy. EPRI has maintained a database containing coupon data spanning more than 25 years, and to date no loss of BORAL{sup R} efficacy has been observed. In the United States, about 50% of the plants use BORAL{sup R} as neutron absorber material in SFPs. BORAL{sup R} is also used in nuclear plants in Mexico, Korea, Taiwan, and other countries. Given that the plant lifetime is increasing-and, hence, the SFP lifetime-EPRI initiated a five-year accelerated corrosion test in 2013 to evaluate and determine the long-term performance of BORAL{sup R} in SFPs. This paper presents an overview of the accelerated corrosion test project. Results from the Year 1 and Year 2 test results are also presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Initial Shipment of MNSR Reactor HEU Spent Nuclear Fuel from Ghana to China

The repatriation of Russian and US origin fresh and spent, high enriched uranium (HEU) research reactor fuel has been conducted since 1996. In 2006, efforts were initiated to convert Chinese designed Miniature Neutron Source Reactors (MNSRs) from HEU to low enriched uranium (LEU) fuel. The Ghana Research Reactor-1 (GHARR-1) is the first of five such MNSR reactors outside of China eligible for conversion and fuel return to China. Under a project involving China and Ghana as operating country, as well as the U.S.A and the IAEA the HEU core is to be repatriated to China and a new LEU core is to be installed. The present paper discusses the preparation of the HEU core for removal, describing the logistics of the shipment of the HEU core and the lessons learned which can be applied to the remaining MNSR shipments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Use of tsunami in the validation of scale 6.1 for the Swedish spent fuel repository - 14381

In March 2011 SKB applied to the Swedish Radiation Safety Authority (SSM) for a permit to build a final repository for spent Fuel and an encapsulation plant (Clink). The application contained a criticality safety analysis for the canisters to be used in Clink and the copper canister to be used in the final repository. In 2012 SSM requested several updates to the application. One of SSM's concerns was the validation of the codes. SSM emphasized that SKB must improve the method to select critical experiments to the validation suite. To satisfy this demand, SKB used the tool Tsunami in the Scale 6.1 code package to select experiments from the handbook of criticality experiment (IHECSBE). With help of sensitivity files SKB evaluated almost 700 experiments from the LEU-COMP-THERM library. SKB then picked the experiments that most resembled our safety cases. The sensitivity coefficient, ck, from Tsunami has been used to evaluate the similarity between the experiments and the applications. It has been considered that a ck over 0.8 provides a 'good enough' similarity between the experiment and our evaluated safety cases. For the different safety cases between 50-600 experiments satisfied the 0.8 criteria giving a good base of data for statistical analysis. It is worth noticing that PWR-fuel generally receives a higher ck than BWR fuel. Initial analysis shows that the differences in results between BWR and PWR can be explained by a larger sensitivity for the U{sup 238} n,n' part of the cross section in the PWR cases. To make sure the validation covers all materials and required physical parameters a gap analysis was made using engineering judgement by criticality safety experts. From this exercise additional experiments were added to the suite. This paper will discuss the Tsunami results for different safety cases and the reasons for differences between them. It will also present the final results of the validation and discuss experiments that were excluded from the validation suite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Validation study for crediting chlorine in critically analyses for spent nuclear fuel disposition - 14428

Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 k{sub eff}). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Analysis of an existing spent fuel pool storage rack using current practices and updated depletion bias and uncertainty values - 14458

In accordance with NRC Information Notice 2011-03: Nonconservative Criticality Safety Analysis for Fuel Storage, a US nuclear power plant reviewed their analysis of record (AOR) for possible nonconservative technical specifications related to the criticality safety of fuel storage. During this review, it was determined that some legacy fuel assemblies did not meet all the assumptions of the AOR, but did meet the plant Technical Specifications. The plant entered into an Operability Evaluation in order to evaluate the moderator temperature and boron concentration assumptions used in the AOR, as well as evaluate the impact of including additional biases and depletion uncertainties for the non-borated case. As this study was part of a response to the NRC Information Notice, a full assessment was not conducted. A re-analysis of the spent fuel rack in accordance with current practices was performed. This analysis used the bounding core conditions, and included all biases and depletion uncertainties. The results challenged the 10CFR50.68(b) requirement that 'the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with un-borated water'. Credits for additional reactivity offsets were then analyzed. Previous assumptions for the measurement uncertainty were evaluated and subsequently reduced. New values for the depletion bias and uncertainty were applied. More detailed, realistic modeling was used to remove conservatisms. The results showed that the storage racks met the 10CFR50.68(b) criteria for the un-borated case. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Technical basis for peak reactivity burnup credit for BWR spent nuclear fuel in storage and transportation systems - 14601

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k{sub eff}) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15-20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Overview of zion comparative analysis project for assessment of Boral{sup R} neutron absorber material performance and monitoring in spent fuel pools - 14618

Neutron absorber materials are used in spent fuel pool (SFP) storage racks to control reactivity. In the United States, about half of the operating plants currently use BORAL{sup R} neutron absorber material in SFP racks. Although operational experience spanning more than 25 years has not shown any loss of the BORAL{sup R} core material, corrosion effects, including blisters and pits, have been observed and reported. To improve understanding for the behavior of BORAL{sup R} in SFPs and evaluate the performance of monitoring approaches, EPRI initiated the Zion comparative analysis project in 2014. The Zion project involves three phases. During the first phase of the project, in situ measurements were performed on selected SFP rack panels in December 2014. Panels for in situ measurements were selected after an extensive analysis of each panel in the pool to ensure that the selected panels represented the pool in terms of variation in neutron and gamma dose values. As part of the second phase of the project, the remaining 10 coupons from the Zion SFP were shipped to a Penn State University (PSU) laboratory for analysis in January 2015. During the final phase of the project, a subset of the panels for which in situ measurements were performed will be harvested from the Zion SFP, cut into samples, and analyzed at PSU. The overall scope of this project will allow comparison of in situ measurements to real panel data as well as data obtained from coupon analysis. In this paper, an overview of the Zion comparative analysis project, the criteria and procedure for panel selection for in situ measurements and harvesting, preliminary coupon analysis results and in situ measurement results are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Overview of neutron absorber materials used in spent fuel pools - 14622

Spent fuel pools (SFPs) were originally designed with low density storage racks that maintained subcriticality based on geometric separation. With the increasing need for storage, due to elimination of reprocessing as a viable alternative, SFPs with high-density racks required neutron absorber materials to control reactivity and maintain criticality safety margins. The Nuclear Energy Institute (NEI) conducted an industry wide survey in 2013 and the survey was updated recently based on additional utility data provided by Electric Power Research Institute (EPRI). The survey results show that about one-quarter of the pools in the United States either do not have neutron absorber material in the SFP or do not credit it even when it is present. Also, about half of the plants in the U.S. use BORAL{sup R} as the neutron absorber material followed by Metamic, Carborundum, Alcan, and borated stainless steel. For neutron absorber monitoring, the majority of the plants rely on a coupon testing program. However, for those plants that never installed coupons or have exhausted their supply of coupons from coupon sampling, in situ measurements at periodic intervals can be performed to determine the status of the neutron absorber. In this paper, an overview of the types of neutron absorber materials used in SFPs, the distribution of absorber material use by material type, types of monitoring programs as part of aging management programs, operational experience to date with different absorber materials based on monitoring programs, and industry initiatives toward addressing long-term performance issues related to neutron absorber material degradation and monitoring are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Equity Considerations in Siting Consolidated Interim Storage Facilities for Commercial Spent Nuclear Fuel

Questions of equity and fairness are often raised by representatives from State, Tribal and local governments and members of the public when resolving conflicts over siting of nuclear facilities. However, the two terms are not synonymous: equity can be quantified while fairness is much more subjective. The Department of Energy's (DOE) December 2021 Request for Information sought input on how to site federal facilities for the temporary, consolidated storage of spent nuclear fuel (SNF) using a consent-based approach. It is anticipated that just such questions about equity and fairness associated with the potential emergence of more than one host site may be raised by the twelve groups of university, nonprofit and private sector partners (Consortia) working to support community engagement, inclusively involve stakeholders, build relationships, and develop innovative forms of mutual learning and public capacity to participate in the consent-based siting process for a federal consolidated interim storage facility during the current Planning & Capacity Building stage of the consent-based siting process. In terms of DOE's Integrated Waste Management System (IWMS), sociopolitical equity concepts are a factor in all aspects of IWMS development and operations and will be particularly applicable should two or more host sites become a possibility, either by the emergence of multiple volunteer hosts or by pursuit as a IWMS program strategy. This paper summarizes recent analyses that explore how four equity metrics - number of states, division of the projected SNF total by the year 2083, current population and land area - might be perceived by host communities, states and regions if two, three or four sites for hosting federal consolidated interim storage facilities were to be contemporaneously realized. Existing institutional arrangements, including Nuclear Regulatory Commission regions, Federal Energy Regulatory Commission regions and Low-Level Radioactive Waste Disposal Compacts, were selected to create hypothetically merged two, three and four-region scenarios to calculate how the four metrics balance out. Projected SNF burden was prioritized with regional contiguity of the consolidated regions a requirement for a functional scenario. The intent was not to promote or suggest any construct as a program objective; the configurations are geopolitical abstractions only to explore potential perceptions of equity. Understanding the possible issues of equity that could arise may benefit program efforts toward achieving a cooperative federalism wherein both the federal and multiple State governments share the goal of manifesting more than one federal consolidated interim storage facility. The main finding is that it is not possible to optimize for all four metrics simultaneously. The analyses demonstrate that trying to create a sense of equity by backfitting a solution to a random population and land distribution will always contain a degree of artificiality.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Long time scale multiphysics simulation of spent nuclear fuel canister in MOOSE

Pebble-bed reactors are an important class of advanced reactors under consideration for various applications where their fuel would give a significant advantage in siting and high-quality heat production. However, the disposal of their fuel is not as thoroughly studied as other fuel forms. In this study, pebble fuel is analysed in a well known spent fuel canister design to characterize the behavior of this fuel form over a long time scale. The results indicate that after approximately 100 years, decay heat is significantly reduced and the maximum temperature in the canister equalizes with the external temperature. The simulation goes on to an end time of a million years, demonstrating the capability of dealing with long time scales efficiently. We conclude that the canister temperatures seem manageable even with very aggressive loading times and while there are several improvements to be implemented in the future, MOOSE is technically capable of simulating the required scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Spent Fuel Management from Advanced Reactors

Presentation to the IAEA Workshop on the Role of Technological Innovation in Reducing Costs and Improving the Economics of Nuclear Power Generation on spent nuclear fuel management from advanced reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗