Spent Fuel Nondestructive Assay
Development of safeguards technology to increase knowledge about spent nuclear fuel assemblies using new integrated nondestructive assay (NDA) instruments.
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A Historical Review of the Safe Transport of Spent Nuclear Fuel, Rev. 1
This report is a revision to M3 milestone M3FT-16OR090402028 for the former Nuclear Fuels Storage and Transportation Planning Project (NFST), “Safety Record of SNF Shipments.” The US Department of Energy (DOE) has since established the Office of Integrated Waste Management (IWM), which builds on the work begun by NFST, to develop an integrated waste management system for spent nuclear fuel (SNF), including the developm
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International Collaboration on Spent Fuel Disposition in Crystalline Media: FY17 Progress Report
Active participation in international R&D is crucial for achieving the Spent Fuel Waste Science & Technology (SFWST) long-term goals of conducting “experiments to fill data needs and confirm advanced modeling approaches” and of having a “robust modeling and experimental basis for evaluation of multiple disposal system options” (by 2020). DOE’s Office of Nuclear Energy (NE) has developed a strategic plan to advance cooperation with international partners. The international collaboration on the evaluation of crystalline disposal media at Sandia National Laboratories (SNL) in FY17 focused on the collaboration through the Development of Coupled Models and their Validation against Experiments (DECOVALEX-2019) project. The DECOVALEX project is an international research and model comparison collaboration, initiated in 1992, for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. SNL has been participating in three tasks of the DECOVALEX project: Task A. Modeling gas injection experiments (ENGINEER), Task C. Modeling groundwater recovery experiment in tunnel (GREET), and Task F. Fluid inclusion and movement in the tight rock (FINITO).
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CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study
The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).
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High-resolution Gamma Spectroscopy Measurements of Pressurized Water Reactor Spent Nuclear Fuel Rods
International nuclear safeguards use nondestructive assay techniques to verify declared characteristics of spent nuclear fuel (SNF) assemblies. The declared characteristics usually include the initial enrichment, assembly average burnup, and irradiation history. The primary nondestructive assay tool used by the International Atomic Energy Agency has been the fork detector irradiated fuel measurement system, which uses gross gamma and neutron correlations; however, there have been several research publications within the last decade on the use of high-resolution gamma spectroscopy for the determination of SNF characteristics. References [1]–[6] are published journal articles on this subject from 2006–2016. This report describes measurements of 25 individual SNF rods using high-resolution gamma spectroscopy. Details of the measurement geometry, measurement settings, and rod characteristics are documented. These spectra are intended for the validation of simulation codes that calculate SNF inventory and source term characteristics. The spectrum archive is available to evaluate future spectrum analysis methods. The archive contains full axial scan spectra of each SNF rod. This allows more detailed comparison between simulation and nondestructive analysis than is possible on assemblies.
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Life Prediction of Spent Fuel Storage Canister Material
The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result of this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.
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In-cell Re-fabrication and Loss-of-coolant Accident (LOCA) Testing of High Burnup Commercial Spent Fuel
A high burnup commercial spent fuel segment was re-fabricated into a pressurized test train for fuel safety testing in the Irradiated Fuels and Examination Laboratory (IFEL) hot cell facility at Oak Ridge National Laboratory. Descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident (LOCA) test train are provided. An in-cell integral LOCA test was performed in the Severe Accident Test Station (SATS) by subjecting the fuel segment to a steam environment, internal pressure, temperatures up to 1200 ºC, and water quench.
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Spent Nuclear Fuel and High-Level Radioactive Waste Inventory Report
This report provides information on the inventory of commercial spent nuclear fuel (SNF), as well as Government-managed SNF and high-level radioactive waste (HLW). Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future commercial nuclear power generation scenarios.
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International Collaboration on Spent Fuel Disposition in Crystalline Media: FY18 Progress Report
Active participation in international R&D is crucial for achieving the Spent Fuel Waste Science & Technology (SFWST) long-term goals of conducting "experiments to fill data needs and confirm advanced modeling approaches" and of having a "robust modeling and experimental basis for evaluation of multiple disposal system options" (by 2020). DOE's Office of Nuclear Energy (NE) has developed a strategic plan to advance cooperation with international partners. The international collaboration on the evaluation of crystalline disposal media at Sandia National Laboratories (SNL) in FY18 focused on the collaboration through the Development of Coupled Models and their Validation against Experiments (DECOVALEX- 2019) project. The DECOVALEX project is an international research and model comparison collaboration, initiated in 1992, for advancing the understanding and modeling of coupled thermo-hydro-mechanicalchemical (THMC) processes in geological systems. SNL has been participating in three tasks of the DECOVALEX project: Task A. Modeling gas injection experiments (ENGINEER), Task C. Modeling groundwater recovery experiment in tunnel (GREET), and Task F. Fluid inclusion and movement in the tight rock (FINITO). FY18 work focused on Task C and preparing the interim reports for the three tasks SNL has been imvolved. The major accomplishments are summarized
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Evaluation of Spent Nuclear Fuel Disposition in Salt (FY18)
This report summarizes the 2018 fiscal year (FY18) field, laboratory, and modeling work funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign as part of the Sandia National Laboratories Salt Research and Development (R&D) and Salt International work packages. This report satisfies level-two milestone M2SF-18SNO10303031and comprises three related but stand-alone sections. The first section summarizes the programmatic progress made to date in the DOE-NE salt program and its goals going forward. The second section presents brine composition modeling and laboratory activities related to salt evaporation experiments, which will be used to interpret data collected during the heater test. The third section presents theoretical and numerical modeling work done to investigate the effects brine composition have on dihedral angle and the permeability of salt.
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Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository
This report discusses an initial criticality consequence analysis approach of commercial spent fuel canisters in a repository that will be developed to support a performance assessment (PA) of a repository with criticality events. A repository PA with criticality events may be needed to support direct disposal of currently loaded dual-purpose canisters. The ultimate goal is to develop a criticality consequence analysis model that can interface with any repository PA model. The criticality consequence analysis approach describes multi-physics coupling between neutronics, thermal hydraulics, and mechanics analysis codes. This report also discusses the sensitivity and uncertainty analysis approach for quantifying conservatism in the criticality consequence outcomes, as well as a defensible multi-physics validation approach to support any future licensing effort. This criticality consequence analysis approach will be refined in the future to support specific regulatory frameworks for a specific repository PA if needed.
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Ruthenium Volatiliazation During Dissolution of Spent Nuclear Fuels in H-Canyon Dissolvers
During nuclear material processing, many radioactive elements form volatile species under certain process conditions which may allow release to the environment unless they are removed from offgas streams. Ruthenium, a product from the fission of U, will form a volatile compound (RuO 4 ) under highly oxidizing conditions which may be released during spent nuclear fuel (SNF) dissolution. The principal Ru isotopes of concern are 103 Ru and 106 Ru with half-lives of 39.27 and 372.6 days, respectively. Ruthenium-103 will decay to less than 1% of its initial activity within one year; therefore, its presence is not a concern for fuel cooled more than one year after discharge. However, the half-life of 106 Ru is sufficiently long that it will persist for many years at levels of concern. The isotope does not decay to less than 1% of its initial activity until approximately 6.8 years. Therefore, the dissolution of SNF cooled for less than approximately 7 years is defined as processing short-cooled fuel as it relates to Ru volatilization.
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Spent Fuel Reprocessing
The energy and material requirements for operating a facility that reprocesses spent nuclear fuel. To access the model referenced in the report, please visit https://www.netl.doe.gov/energy-analysis/details?id=78436374-0F3D-41D4-8660-D01E04C52ABF
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Truck transport of spent uranium fuel
Transport of spent uranium fuel from a nuclear power plant to long-term waste disposition. Assumes backhaul and front haul have different energy intensities. Includes diesel consumption. To access the model referenced in the report, please visit https://www.netl.doe.gov/energy-analysis/details?id=539A6261-14AD-4F85-AC24-DE343D8FF331
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Oxyhydroxides on Aluminum Spent Nuclear Fuel: Formation Studies and Removal Practices to Prevent Radiolytic Gas Production
Aluminum-clad Spent Nuclear Fuel (ASNF) forms hydrated oxides or (oxy)hydroxides on the cladding due to water exposure during irradiation and storage. This poses challenges to sealed (road-ready) dry storage due to risks of release of bound water and decomposition releasing hydrogen and or oxygens. To assess the risks, it is necessary to understand the (oxy)hydroxides’ growth/formation and morphology as the precursor state for drying and their gas generation from radiolysis. This will help determine safe (oxy)hydroxide loadings on dry-stored fuel as well as methods of removal. In this work, (oxy)hydroxides were grown on aluminum alloy substrates and the resulting layers studied for morphology and thickness; radiolysis studies were performed on powders to correlate morphology and particle size to gas generation, and drying studies were performed on powders and adherent oxide films. These experiments will be used to guide development of drying strategies for safe dry storage.
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Crack Growth Rate and Large Plate Demonstration of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters
This report describes the body of work involving testing and analyses performed in FY18 and FY19 at the Savannah River National Laboratory and at Korea University to determine chloride-induced stress corrosion crack (CISCC) growth rates in weldments of austenitic stainless steel plate material. The primary work involved: several separate crack growth rate test campaigns with new testing setups using laboratory-scale corrosion specimens; analyses to estimate the stress intensity factors for flaws postulated to be present in the residual stress fields of axial and circumferential weldments in a prototypic dry storage canister; and establishment of a “large plate” (approximately 51×46 cm with thickness 16 mm and weight about 30 kg) CISCC test that uses a plate with a circumferential weldment harvested from the Sandia National Laboratories’ full-size mockup dry storage canister. The tests for crack growth rate were conducted on ASTM E1861 bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse the brine to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The crack growth rate test results matched previous experimental data in the open literature. These literature data had been compiled by SNL and EPRI to develop a temperature-dependent (K-insensitive) crack growth rate model to support flaw disposition in the ASME Boiler and Pressure Vessel Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.” The knowledge of the salt, temperature, humidity conditions for CISCC including the threshold stress intensity factor, $K_{ISCC}$, were used to setup the large plate test in which a set of partthrough- wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a constant 75% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC. The visual results at the surface of the plate show corrosion product staining (brown rust appearance) but with no CISCC emanating from the machined notches after approximately 5 months exposure to date. The plate test will continue in FY20 and phased-array ultrasonic testing and other NDE techniques will be applied to identify and characterize CISCC that may emanate from the machined defects. The notches and any CISCC from them will be sectioned at the conclusion of the test to verify NDE characterization, and to develop effective through-wall crack growth rates from cracks that would likely be branched.
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MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel
The work presented in this report applies the MELCOR code to evaluate potential accidents in non-reactor nuclear facilities, focusing on Design Basis Accidents. Ten accident scenarios were modeled using NRC's best-estimate severe accident analysis code, MELCOR 2.2. The accident scenarios simulated a range of explosions and/or fires related to a nuclear fuel reprocessing facility. The objective was to evaluate the radionuclide source term to the environment following initiating explosion and/or fire events. The simulations were performed using a MELCOR model of the Barnwell Nuclear Fuel Plant, which was decommissioned before beginning reprocessing operations. Five of the accident scenarios were based on the Class 5 Design Basis Accidents from the Final Safety Analysis Report. Three of the remaining accident scenarios include sensitivity studies on smaller solvent fires. The final two accidents included an induced fire from an initial explosion. The radionuclide inventory was developed from ORIGEN calculations of spent PWR fuel with an initial enrichment of 4.5% U-235 by weight. The fuel aged for five years after a final 500-day irradiation cycle. The burn-up was conservatively increased to 60 GWd/MTU to bound current US operations. The results are characterized in terms of activity release to the environment and the building decontamination factor, which is related the leak path factor used in Department of Energy safety analyses. The MELCOR 2.2 results consider adverse consequences to the filters, ventilation system, and structures as a result of the explosions and fires. The calculations also include best-estimate models for aerosol transport, agglomeration, and deposition. The new calculations illustrate best-estimate approaches for predicting the source term from a reprocessing facility accident.
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Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design
The purpose of this document is to provide the technical specifications for the Engineering Scale Design for the drying tests of aluminum-clad spent nuclear fuel (ASNF). The engineering scale design is to accommodate both forced helium drying (FHD) as well as vacuum drying. The experiments will employ mock ASNF assemblies with a limited number of chemically treated aluminum plates to simulate the aluminum hydroxides found on ASNF. The goal here is to understand the transformation that may occur in plates including the loss of chemisorbed water and changes in morphology. Physiosorbed and bulk water found in experiments will be dried in the process and the effect on drying operations noted. Decay heat will be simulated in one assembly using a custom designed resistance heating element. A typical radial arrangement of assemblies will be employed using a modified Type 1A Basket design. For the chamber/canister, a single assembly layer with full radial dimension will be employed. Modifications to the standard canister design and configuration include the addition of a siphon tube to enable recirculation of hot gas for forced helium drying. A ring of ports/flanges are added near the top of the chamber to accommodate the feedthroughs for instrumentation and power needed for the experiment and data collection. Four viewports are added to the main section of the chamber with two in line with one of the fuel assembly and siphon tube. The other two viewports are on the opposite side of the chamber. The purpose here is to provide the option of thermal imaging in addition to the nearly four dozen thermocouples available inside the chamber. Modeling of the chamber, basket, and mock ASNF assemblies was conducted to estimate temperature distributions throughout the chamber during forced helium drying and over the anticipated drying period. This modeling was used as input to decisions on the Engineering Scale Design. Work on modeling continues in the period during setup and testing to develop preliminary drying models. Analysis of experimental results will later be used to validate these and subsequent models.
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