This report is a condensed version of previous reports identifying technical gaps that, if addressed, could be used to ensure the continued safe storage of SNF for extended periods and support licensing activities. This report includes updated gap priority assessments because the previous gap priorities were based on R&D performed through 2017. Much important work has been done since 2017 that requires a change in a few of the priority rankings to better focus the near-term R&D program. Background material, regulatory positions, operational and inventory status, and prioritization schemes are discussed in detail in Hanson et al. (2012) and Hanson and Alsaed (2019) and are not repeated in this report. One exception is an overview of the prioritization criteria for reference. This is meant to give the reader an appreciation of the framework for prioritization of the identified gaps. A complete discussion of the prioritization scheme is provided in Hanson and Alsaed (2019).
The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.
The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.
This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.
The United States Department of Energy (DOE) is conducting research and development (R&D) activities within the Used Fuel Disposition Campaign to support the implementation of the DOE's 2013 Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. R&D activities focus on storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles and are ongoing at nine national laboratories. Additional relevant R&D is conducted at multiple universities through the DOE's Nuclear Energy University Program. Within the storage and transportation areas, R&D continues to focus on technical gaps related to extended storage and subsequent transportation of UNF. Primary emphasis for FY15 is on experimental and analysis activities that support the DOE s dry cask demonstration confirmatory data project initiated at the North Anna Nuclear Power Plant in Virginia by the Electric Power Research Institute in collaboration with AREVA and Dominion Power. Within the disposal research area, current planning calls for a significant increase in R&D associated with evaluating the feasibility of deep borehole disposal of some waste forms, in addition to a continued emphasis on confirming the viability of generic mined disposal concepts in multiple geologic media. International collaborations that allow the U.S. program to benefit from experience and opportunities for research in other nations remain a high priority.
This Update to the Used Fuel Disposition Campaign Implementation Plan provides summary level detail describing how the Used Fuel Disposition Campaign (UFDC) supports achievement of the overarching mission and objectives of the Department of Energy Office of Nuclear Energy Fuel Cycle Technologies Program, building on work completed in this area since 2009. This implementation plan begins with the assumption of target dates that are set out in the January 2013 DOE Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (http://energy.gov/downloads/strategy-management-and-disposal-used-nuclearfuel- and-high-level-radioactive-waste). These target dates and goals are summarized in section III. This implementation plan will be maintained as a living document and will be updated as needed in response to available funding and progress in the Used Fuel Disposition Campaign and the Fuel Cycle Technologies Program.
The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.
This report describes a test of an instrumented surrogate PWR fuel assembly on a truck trailer conducted to simulate normal conditions of truck transport. The purpose of the test was to measure strains and accelerations on a Zircaloy-4 fuel rod during the transport of the assembly on the truck. This test complements tests conducted in FY13 in which the same assembly was placed on a shaker and subjected to vertical vibrations and shocks simulating truck transport. The results of those tests are in the report “FUEL ASSEMBLY SHAKER TEST for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Transport” McConnell, et al., SAND2013-5210P, Rev. 0.1, FCRD-UFD-2013-000190, June 30, 2013 (revised December 1, 2013). This report constitutes the Milestone M2FT-14SN0813041 for the DOE/NE Fuel Cycle Research and Development Used Fuel Disposition Campaign ST Transportation Work Package FT-14SN081304 (Rev. 1). The strains measured on the instrumented Zircaloy-4 rod over a 40.2 mile route in the Albuquerque area over a variety of road conditions – rough dirt to Interstate highway (Figure S.1) – never exceeded 150 µin./in. – a very low level of strain well below the elastic limit/yield strength of Zircaloy-4, Figure S.1. The strains measured in the truck test were slightly lower than those measured in the shaker tests.
11th International Probabilistic Safety Assessment and Management Conference and the Annual European Safety and Reliability Conference 2012, PSAM11 ESREL 2012
Various circumstances around the world have resulted in the potential need to store used nuclear fuel longer than times allowed by the regulations. While current storage of used fuel is safe and is likely to remain safe for extended periods of time, material degradation issues may arise that have not necessarily been considered when used fuel storage was licensed for relatively short periods of time. Material degradation issues associated with the fuel, cask internals, storage overpack, closure seals and bolts, and the storage pad all need to be assessed relative to long term performance. Key functional requirements for this long term performance include subcriticality, containment, shielding, thermal performance, and retrievability. A sufficient degree of understanding of the material degradation issues relative to the functional requirements for storage is necessary to develop the technical basis to ensure material performance over extended periods of time. An important initial step in addressing material degradation issues is to identify technical data gaps relative to existing understanding that are important over long storage periods. An effort has been under way since June 2010 to develop a list and prioritization of technical gaps from an international perspective. This effort is being conducted under the aegis of the U.S. Electric Power Research Institute (EPRI) Extended Storage Collaboration Program (ESCP). As part of this program, an International Subcommittee has been established to solicit the international community's input on storage system material degradation issues associated with long term storage and transportation. The first goal of this subcommittee is to develop a report on the technical data gaps from an international perspective. Since used fuel is stored in various configurations around the world, it is expected that different priorities will be identified relative to importance in maintaining the key performance functions. The second goal of the subcommittee is to identify areas for international collaboration for research on degradation issues in order to leverage existing program and facilities. The current status of the international data gap effort is a draft list of High, Medium, and Low priority issues that should be addressed to demonstrate sufficient understanding of material performance of storage system components over extended operational periods. Although there are differences in the identified gaps and associated priorities due to different regulations and storage and transportation systems, there are also areas of commonalities that are important to recognize. These are the areas that have the greatest potential for collaboration.
The safe management and disposition of used nuclear fuel and/or high level nuclear waste is a fundamental aspect of the nuclear fuel cycle. The United States currently utilizes a once-through fuel cycle where used nuclear fuel is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. However, a decision not to use the proposed Yucca Mountain Repository will result in longer interim storage at reactor sites than previously planned. In addition, alternatives to the once-through fuel cycle are being considered and a variety of options are being explored under the U.S. Department of Energy's Fuel Cycle Technologies Program. These two factors lead to the need to develop a credible strategy for managing radioactive wastes from any future nuclear fuel cycle in order to provide acceptable disposition pathways for all wastes regardless of transmutation system technology, fuel reprocessing scheme(s), and/or the selected fuel cycle. These disposition paths will involve both the storing of radioactive material for some period of time and the ultimate disposal of radioactive waste. To address the challenges associated with waste management, the DOE Office of Nuclear Energy established the Used Fuel Disposition Campaign in the summer of 2009. The mission of the Used Fuel Disposition Campaign is to identify alternatives and conduct scientific research and technology development to enable storage, transportation, and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The near-and long-term objectives of the Fuel Cycle Technologies Program and its ' Used Fuel Disposition Campaign are presented.
Transportation for each step of a closed fuel cycle is analyzed in consideration of the availability of appropriate transportation infrastructure. The United States has both experience and certified casks for transportation that may be required by this cycle, except for the transport of fresh and used MOX fuel and fresh and used Advanced Burner Reactor (ABR) fuel. Packaging that had been used for other fuel with somewhat similar characteristics may be appropriate for these fuels, but would be inefficient. Therefore, the required neutron and gamma shielding, heat dissipation, and criticality were calculated for MOX and ABR fresh and spent fuel. Criticality would not be an issue, but the packaging design would need to balance neutron shielding and regulatory heat dissipation requirements.
This multinational test program is quantifying the aerosol particulates produced when a high energy density device (HEDD) impacts surrogate material and actual spent fuel test rodlets. The experimental work, performed in four consecutive test phases, has been in progress for several years. The overall program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This program also provides significant political benefits in international cooperation for nuclear security related evaluations. The spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC), and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission. This report summarizes the preliminary, Phase 1 work performed in 2001 and 2002 at Sandia National Laboratories and the Fraunhofer Institute, Germany, and documents the experimental results obtained, observations, and preliminary interpretations. Phase 1 testing included: performance quantifications of the HEDD devices; characterization of the HEDD or conical shaped charge (CSC) jet properties with multiple tests; refinement of the aerosol particle collection apparatus being used; and, CSC jet-aerosol tests using leaded glass plates and glass pellets, serving as representative brittle materials. Phase 1 testing was quite important for the design and performance of the following Phase 2 test program and test apparatus.
This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) and supported by both the U.S. Department of Energy and the Nuclear Regulatory Commission.
The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratories has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The authors provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. They focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests, and briefly summarize similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods.
A multinational test program is in progress to quantify the aerosol particulates produced when a high energy density device, HEDD, impacts surrogate material and actual spent fuel test rodlets. This program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments; the program also provides significant political benefits in international cooperation. We are quantifying the spent fuel ratio, SFR, the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are crucial for predicting radiological impacts. This document includes a thorough description of the test program, including the current, detailed test plan, concept and design, plus a description of all test components, and requirements for future components and related nuclear facility needs. It also serves as a program status report as of the end of FY 2003. All available test results, observations, and analyses - primarily for surrogate material Phase 2 tests using cerium oxide sintered ceramic pellets are included. This spent fuel sabotage - aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC, and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission.
A multi-attribute utility analysis is applied to a decision process to select a treatment method for the management of aluminum-based spent nuclear fuel (Al-SNF) owned by the US Department of Energy (DOE). DOE will receive, treat, and temporarily store Al-SNF, most of which is composed of highly enriched uranium, at its Savannah River Site in South Carolina. DOE intends ultimately to send the treated Al-SNF to a geologic repository for permanent disposal. DOE initially considered ten treatment alternatives for the management of Al-SNF, and has narrowed the choice to two of these: the direct disposal and melt and dilute alternatives. The decision analysis presented in this document focuses on a formal decision process used to evaluate these two remaining alternatives.
A multi-attribute utility analysis is applied to a decision process to select a treatment method for the management of aluminum-based spent nuclear fuel (Al-SNF) owned by the US Department of Energy (DOE). DOE will receive, treat, and temporarily store Al-SNF, most of which is composed of highly enriched uranium, at its Savannah River Site in South Carolina. DOE intends ultimately to send the treated Al-SNF to a geologic repository for permanent disposal. DOE initially considered ten treatment alternatives for the management of Al-SNF, and has narrowed the choice to two of these: the direct disposal and melt and dilute alternatives. The decision analysis presented in this document focuses on a formal decision process used to evaluate these two remaining alternatives.
As the United States embarks upon a major effort to cleanup its nuclear defense facilities, a large quantity of low-level waste (LLW) will be generated. This LLW must be managed and ultimately placed into final disposal. Much of this waste is expected to exceed certain limits defined in U.S. regulations (Title 10, U.S. Code of Federal Regulations, part 61) called Class C. The waste which exceeds Class C, called Greater-than-Class-C (GTCC), poses a major challenge to waste managers. Each GTCC waste form must be placed into costly geologic disposal unless separate approval is obtained from the United States regulator to place it into less costly {open_quotes}near-surface{close_quotes} land burial. Management of GTCC will also require, to some extent, storage and transport prior to its final disposal. A further LLW stream exists in the United States also stemming from the prior operations of United States defense facilities, viz., radioactively contaminated and irradiated scrap metal which has been accumulating over the past forty years. Similarly, as cleanup, decontamination, and decommissioning proceeds, this contaminated scrap metal inventory is expected to grow rapidly. This paper explores the notion of the authors that an opportunity for a synergistic solution to two difficult waste management problems may be available in the United States today, and perhaps may similarly be available in other nuclear countries as well. The possibility exists for fabricating packagings from contaminated scrap metal (which would otherwise be part of the waste inventory) and for using these packaging for storage, transport and disposal of GTCC in near-surface burial facilities without reopening or repacking. This approach is appealing and should lead to major safety and cost benefits. An examination of existing regulations with the intent to propose additions, changes, or clarifications that would effectively and beneficially regulate such combined activity is proposed.
Two major initiatives are underway in the US that are creating a significant financial impact on both the US taxpayer and on users of electric power. First, the US Department of Energy (DOE) has been tasked with cleaning-up the defense complex. This task is managed under the direction of the Office of Environmental Restoration and Waste Management (EM) of the DOE. The waste that EM must address includes radioactive, hazardous, and mixed that consists of both radioactive and hazardous constituents. Second, the DOE is required by the Nuclear Waste Policy Act (NWPA) to take title to commercial nuclear spent fuel assemblies starting in 1998. The DOE Office of Civilian Radioactive Waste Management (OCRWM) was established to carry out this charter. Since a final repository is not scheduled for opening until 2010 at the earliest, the DOE is planning on providing a Monitored Retrievable Storage (MRS) facility for centralized storage to bridge the time gap between 1998 and 2010. The NWPA requires that nuclear utilities pay a fee into a specific fund that Congress uses to pay the DOE for the development of the MRS, the transportation system, and the repository. This fund, along with the EM budget, constitutes a multi-billion dollar effort to manage DOE nuclear waste and to store and dispose of commercial spent nuclear fuel. These two seemingly unrelated problems have aspects of commonality that can be considered for the benefit of both programs, the US taxpayer, and the utility rate payer. Both programs are the responsibility of the DOE, and both will require engineered packages for storage, transportation, and disposal of the EM waste and commercial spent fuel. Rather than using specialized systems for each step (storage, transport, and disposal), a concept for a Universal Container System has been developed that could potentially simplify the overall waste management system, reduce expensive handling operations, and reduce total system cost.
The Department of Energy (DOE) is investigating the use of ductile cast iron (DCI) as a candidate material for radioactive material transportation cask construction. The investigation will include materials testing and full-scale cask testing. The major effort will focus on materials qualification and cask evaluation of the 9 meter and puncture drop test events. The test plan shall include a series of drop tests, and several core bars will be removed from the casting wall for material properties testing. Of particular interest is the evaluation of the material microstructure and fracture toughness parameters. Test instrumentation, used to define cask deceleration loads and strain during the drop tests, will be strategically placed in areas of the greatest structural interest. Part of the testing will include placement of an induced flaw. At the conclusion of the cask drop tests, the induced flaw(s) will be sectioned from the cask body for metallurgical examination. All test results will be documented in the safety analysis report for packaging for submission to the US Nuclear Regulatory Commission (NRC). The goal of this program is a certificate of compliance for DCI from the NRC to transport high-level radioactive materials. The acceptance of DCI within the USA cask design community will offer an alternative to present-day materials for cask construction, and its entry has the potential of providing significant cost-savings.
Borated stainless steel tensile testing is being conducted at Sandia National Laboratories (SNL). The goal of the test program is to provide data to support a code case inquiry to the ASME Boiler and Pressure Vessel Code, Section 3. The adoption by ASME facilitates a materials qualification for structural use in transport cask applications. The borated stainless steel being tested conforms to ASTM specification A-887, which specifies 16 grades of material as a function of boron content (0.20% to 2.25%) and fabrication technique. For transport cask basket applications, the potential advantage to using borated stainless steel arises from the fact that the structural and criticality control functions can be combined into one material. The test program at SNL involves procuring material, machining test specimens, and conducting the tensile tests. From test measurements obtained so far, general trends indicate that tensile properties (yield strength and ultimate strength) increase with boron content and are in all cases superior to the minimum required properties established in SA-240, Type 304, a typical grade of austenitic stainless steel. Therefore, in a designed basket, web thickness using borated stainless steel would be comparable to or thinner than an equivalent basket manufactured from a typical stainless steel without boron additions. 3 figs., 5 tabs.
