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At least 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Update on Investigations of Viability of Cold Spray and FSW as a Spent Nuclear Fuel Dry Storage Canister Mitigation Tool

The purpose of this work is to investigate the use of cold spray and friction stir welding or processing (FSW/P) for repair and mitigation of chloride-induced stress corrosion cracking (CISCC) in dry cask storage system (DCSS) canisters to ensure integrity far beyond their original license period. This report provides a viability analysis of cold spray and friction stir welding for repair and fabrication of DCSS canisters. This report presents experimental work and a research strategy to optimize and evaluate cold spray and FSW/P for CISCC resistance. The literature reports that the following three conditions must exist for CISCC to occur: tensile stress, corrosive environment, and susceptible material. These conditions must be understood and accounted for while evaluating CISCC susceptibility. Furthermore, non-obvious factors, such as surface conditions and geometric confinement, can dramatically increase the intensity of one or more of the three listed conditions for CISCC. Arc welding should be considered for either temporary or permanent repair when combined with a mitigation technology. It is known that alloys used in the existing inventory of spent nuclear fuel storage canisters become sensitized in areas around fusion welds. Technologies, such as cold spray or FSW, can mitigate the sensitization caused by fusion welds. FSW/P and cold spray, potential alternatives for fabrication and repair, are solid-phase processes. No melting occurs and the energy input can be controlled such that no harm is done to the base metal during these processes. FSW/P should be considered for use in new fabrication of canisters and repair of existing canisters. Low heat input and grain refinement result in material with improved mechanical properties and comparable generalized corrosion performance relative to base metal. FSW/P offers the potential of through-thickness CISCC resistance. Cold spray should be considered for production, mitigation, and repair of canisters. Cold spray has demonstrated the ability to deposit stainless steel, Inconel alloys, and other alloys on stainless steel 304L such that galvanic potential is matched and resistance to pitting is improved. The technical basis for cold spray mitigation and repair within the overpack using remote robotic equipment is established. A non-obvious application to the research community is the sealing of crevices. All reported instances of CISCC in the fielded nuclear applications were associated with crevice corrosion. This can be avoided by cold spraying to seal crevices. The bulk of the work planned for FY 2020 will be generating coupons for CISCC testing by collaborative projects at Pacific Northwest National Laboratory and Sandia National Laboratories. First an exploratory study will be executed that will include screening studies and process optimization. This task will identify the effect of process parameters on microstructure and CISCC resistance. Optimized parameters will be used to produce a smaller number of coupons for full testing and evaluation. Test methods for evaluating CISCC resistance must meet the following criteria: (1) normalized surface geometry and exposed process microstructure, (2) account for residual stresses, (3) be in a controlled and appropriate test environment, and (4) be numerically quantifiable and repeatable. Failure to meet these criteria could lead to results that cannot be used for comparative analysis of processes and materials.

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DOE OSTI2016

Safeguards Implications for Deep Borehole Disposal of Spent Fuel

Deep borehole disposal (DBD) has been suggested as an option for disposing spent nuclear fuel in a number of countries, including several countries that are subject to international safeguards. While potential benefits of deep borehole disposal include increased safety, reduced cost, and greater flexibility, the method could also impact the implementation of international safeguards. DBD presents some unique safeguards challenges compared to a conventional MGR. These challenges include 1) verifying borehole design below the surface; 2) strong reliance on CoK up to and including disposal; 3) limitations on the ability to observe or verify successfully emplaced canisters; and 4) successfully monitoring a closed and sealed DBD facility over the long term. In some cases, such challenges may prove easier for a DBD facility than for a conventional MGR, others more difficult, and still others may require new methodologies (or existing methodologies newly applied to safeguards). Long-term monitoring in particular might be somewhat less onerous.

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DOE OSTI2019

Oxidation of Spent Fuel During Fuel Handling Accidents Test Plan

The transfer of commercial spent nuclear fuel (CSNF) from transport or dry storage cask into final repository cask represents a scenario where fuel elements are potentially at risk to exposure to air and elevated temperatures. In the event of an accident during handling operations, the facility cooling may cease to function, and the temperature of the fuel would begin to rise due to decay heat generation. CSNF is generally stored in an inert environment, but handling operations could be performed in an air environment. If any damaged fuel elements are present during this postulated accident the fuel could begin to oxidize from UO 2 to U 3 O 8 . If the oxidation event is severe, volume expansion from oxidation could result in large-scale cladding failure and large releases of CSNF debris into the transfer facility. Depending on the design of the fuel handling facility, any SNF release could have an impact on off-site dose.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Aluminum Spent Fuel Performance in Dry Storage Task 4 Aluminum Oxide Sampling of ATR Dry Stored Fuel

Milestone report on sampling of long-term dry-stored ATR spent fuel elements. It describes tooling, sample acquisition methods and analysis of the aluminum oxide samples by scanning electron microscopy, thermogravimetric analysis, X-ray diffraction and transmission electron microscopy. The observations provide input to projections on controls necessary for maintaining the integrity of the fuel during extended interim storage.

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DOE OSTI2020

Corrosion-Resistant Coatings for Mitigation and Repair of Spent Nuclear Fuel Dry Storage Canisters

This report summarizes the results of a literature survey on coatings and surface treatments that are used to provide corrosion protection for exposed metal surfaces. The coatings are discussed in the context of being used on stainless steel spent nuclear fuel (SNF) dry storage canisters for potential prevention or repair of corrosion and stress corrosion cracking. The report summarizes the properties of different coating classes, including the mechanisms of protection, their physical properties, and modes of degradation (thermal, chemical, radiological). Also discussed are the current standard technologies for application of the coatings, including necessary surface pretreatments (degreasing, rust removal, grinding) and their effects on coating adhesion and performance. The coatings are also classified according their possible use for in situ repair; ex situ repair, requiring removal from the overpack; and ex situ prevention, or application prior to fuel loading to provide corrosion protection over the lifetime of the canister.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report provides a description of the testing (in progress at the Savannah River National Laboratory) to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The details of the establishment of this plate test, the interim corrosion results (from visual examination), and the general plan for examination of the plate in CY21 to evaluate crack growth behavior (if any) are provided. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage. A “large plate” (approximately 51 x 46 cm with thickness 16 mm and weight about 30 kg) and containing a circumferential weldment from the mockup is the test specimen. The knowledge of the salt, temperature, humidity conditions for chloride-induced stress corrosion cracking (CISCC) including the threshold stress intensity factor, K iscc , were used to setup the large plate test in which a set of part-through-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 marine salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a sustained 71-73% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC

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DOE OSTI2020

Spent Nuclear Fuel and Reprocessing Waste Inventory

This report provides information on the inventory of commercial spent fuel (SNF) and high-level radioactive waste in the United States, as well as non-commercial SNF and reprocessing waste in the U.S. Department of Energy (DOE) complex. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future commercial nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).

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DOE OSTI2020

Effects of increasing the initial nitric acid concentration from a maximum of 7.5 to 8.5 M during the dissolution of aluminum spent nuclear fuel

H-Canyon is blending solutions from the dissolution of High Flux Isotope Reactor (HFIR) and Material Test Reactor (MTR) fuels with Target Residue Material (TRM) to prepare feed solution for the 1st Cycle of solvent extraction. The initial acid concentration for HFIR fuel dissolution is limited to 7.5 M by the flowsheet; however, an increase in the initial concentration is desirable to more easily achieve the target acidity for solvent extraction using the current blending strategy. To provide flexibility in batching the highly enriched uranium (HEU) solutions, the Savannah River National Laboratory (SRNL) was requested to evaluate the potential for increasing the maximum HNO3 concentration for HFIR fuel dissolution from 7.5 to 8.5 M. In response to this request, a literature review was performed to evaluate the impact of a higher starting HNO3 concentration on the dissolution of aluminum spent nuclear fuels (ASNF) including both HFIR and MTR fuels.

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DOE OSTI2021

Spent Crystalline Silicotitanate Storage Study—Post AP-105DF Processing

A Tank-Side Cesium Removal (TSCR) system is under development by Washington River Protection Solutions to support initial production of immobilized low-activity waste (LAW) by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW Facility. Tank waste supernate is filtered to remove suspended solids and then Cs is removed by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The Cs-loaded CST columns will be stored indefinitely, with a goal of eventual CST removal and treatment. Thus, the spent CST needs to be recoverable after undetermined storage time. Previous testing with AP-105 simulant showed that rinsing the CST bed with 3 bed volumes (1.4 apparatus volumes [AVs]) of 0.1 M NaOH resulted in a dried bed that maintained flow characteristics indicative of ease of recovery. This study explored the intermediate conditions (between feed dried in place and the 1.4 AVs of 0.1 M NaOH rinse) to evaluate CST bed properties.

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DOE OSTI2021

Augmented Monitoring and Condition Assessment Program (AMCAP) - Proof-of-Principle (POP) Mockup for Non-Aluminum Spent Nuclear Fuel Container In-Situ Examinations

A disciplined engineering approach is being followed to develop an engineered system of tooling and sensors, characterization techniques, and deployment subsystems for in-situ inspection of the several container types used for the storage of non-aluminum spent nuclear fuel (NASNF) in L Basin under the Augmented Monitoring and Condition Assessment Program (AMCAP). Inspecting the containers to provide information on their structural condition helps ensure the safe handling and storage of the NASNF containers pending final disposition. This report describes the work performed in the initial two phases of this developmental work, namely the bench scale and proof-of-principle (POP) scale, which focus on sensor selection and the tooling design and fabrication for two remote non-destructive examination (NDE) methods. These methods include visual testing (VT) for a visual examination of the container surfaces and ultrasonic testing (UT) for a n examination to characterize container wall material thickness and flaws.

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DOE OSTI2021

Aerosol Deposition Inside a Spent Fuel Transportation and Storage Canister (Interim Report)

This report documents the progress in preparing the ANSYS/Fluent ® and MELCOR models to perform characterization studies of aerosols dispersal and deposition within a select spent fuel cask system. Steady state thermal-hydraulic cask response is modeled with both codes, at present, while the MELCOR source code is being modified to allow imposed thermal-hydraulic conditions with aerosol physics calculations. This will allow the MELCOR model to assume the thermal-hydraulic calculation from the ANSYS/Fluent ® , while only computing the aerosol physics. Detailed results are presented on the thermal-hydraulic analysis of the MAGNASTOR ® cask for the current ANSYS/Fluent ® model, with convergent conditions observed over two fidelities. While the MELCOR computation computes the steady state conditions, they differ sufficiently from that computed with ANSYS/Fluent ® . The MELCOR analysis include a set of UO 2 sources to investigate system response for interim reporting. The airborne concentrations and evolving distributions are presented. Model development is anticipated to continue as additional components impacting flow conditions or aerosol deposition within the MAGNASTOR ® cask and Westinghouse fuel assemblies are identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Spent Crystalline Silicotitanate Storage Study—Post AP-105DF Processing

A Tank-Side Cesium Removal (TSCR) system is under development by Washington River Protection Solutions to support initial production of immobilized low-activity waste (LAW) by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW Facility. Tank waste supernate will be filtered to remove suspended solids and then Cs will be removed by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The Cs-loaded CST columns will be stored indefinitely, with a goal of eventual CST removal and treatment. Thus, the spent CST needs to be recoverable after undetermined storage time. Previous testing with AP-105 simulant showed that rinsing the CST bed with 3 bed volumes (1.4 apparatus volumes [AVs]) of 0.1 M NaOH resulted in a dried bed that maintained flow characteristics indicative of ease of recovery. This study explored the intermediate conditions (between feed dried in place and the 1.4 AVs of 0.1 M NaOH rinse) to evaluate CST bed properties after: 1) stoppage with feed in place; 2) stoppage after draining feed; 3) stoppage after 0.7 AV of 0.1 M NaOH rinse through column; 4) stoppage after 1.5 AVs of 0.1 M NaOH rinse through column. Post processing, each column was heated at 50 °C for 19 days under pseudo-storage conditions to simulate the expected dried and stored CST bed conditions. Testing was conducted at the small scale (12-mL bed volume); actual, Cs-depleted, AP-105 tank waste was used as the feed. Post-dried CST bed physical properties (angle of repose and penetration depth) were measured to evaluate how CST moved and flowed. All process stop-conditions resulted in a solidified CST bed except for the final condition, 1.5 AVs of 0.1 M NaOH rinse. At this small scale, the three CST beds presented an issue for retrievability after the short storage period (19 days at 50 °C). The latter case confirmed the results from simulant testing. The testing was intended to provide a preliminary assessment of issues that may arise from desiccation of CST during storage with the indicated salt solutions in place. Since these were small-scale tests, the processing system did not scale to full scale conditions exactly; however, the tests did provide insight into the impact on the dried and stored CST bed after stopping processing at an earlier step (upset condition) than normal. These results indicate that if an upset condition occurs at TSCR, a dilute hydroxide rinse should be considered before the CST dries from internal heating.

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DOE OSTI2021

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report is the comprehensive description of the completed experiment to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ (SNL) full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The plate was subjected to conditions aggressive to chloride-induced stress corrosion cracking for a total time of 2 years. The details of the establishment of this experiment, and the corrosion results are reported. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage.

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DOE OSTI2021

Update on the Simulation of Commercial Drying of Spent Nuclear Fuel

The purpose of this report is to document improvements in the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes.

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DOE OSTI2021

Evaluation of Nuclear Spent Fuel Disposal in Clay-Bearing Rock - Process Model Development and Experimental Studies (M2SF-21SN010301072)

The DOE R&D program under the Spent Fuel Waste Science Technology (SFWST) campaign has made key progress in modeling and experimental approaches towards the characterization of chemical and physical phenomena that could impact the long-term safety assessment of heatgenerating nuclear waste disposition in deep-seated clay/shale/argillaceous rock. International collaboration activities such as heater tests, continuous field data monitoring, and postmortem analysis of samples recovered from these have elucidated key information regarding changes in the engineered barrier system (EBS) material exposed to years of thermal loads. Chemical and structural analyses of sampled bentonite material from such tests as well as experiments conducted on these are key to the characterization of thermal effects affecting bentonite clay barrier performance and the extent of sacrificial zones in the EBS during the thermal period. Thermal, hydrologic, and chemical data collected from heater tests and laboratory experiments has been used in the development, validation, and calibration of THMC simulators to model near-field coupled processes. This information leads to the development of simulation approaches (e.g., continuum and discrete) to tackle issues related to flow and transport at various scales of the host-rock, its interactions with barrier materials, and EBS design concept.

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DOE OSTI2021

Modeling Environmental Effects on Ventilated Spent Fuel Storage Systems

This report describes newly developed external environment wind effects models of spent nuclear fuel (SNF) dry storage systems. The primary purpose of these wind effects models is to better understand particle deposition on SNF canisters in the context of chloride-induced stress corrosion cracking. The goal of this effort is to further the understanding and improve the wind effects model of the Nuclear Horizontal Modular Storage (NUHOMS ® ) Advanced Horizontal Storage Module–High Seismic (AHSM-HS) storage system in support of the planned Canister Deposition Field Demonstration (CDFD) project (Durbin et al. 2021, Suffield et al. 2021). The steps in this wind effects investigation begin with a validation case of comparing experimental data with a STAR-CCM+ computational fluid dynamics (CFD) model of the Belowground Vertical Dry Cask Simulator (BVDCS) with the external environment explicitly modeled. Next, a test case is performed with the AHSM-HS models comparing solar-loading strategies for a standalone and wind effects model in STAR-CCM+ and a standalone model in ANSYS Mechanical Parametric Design Language (APDL). The final portion of this study compares results of the validation exercise with applications of wind effects models for two horizontal storage systems, a site specific NUHOMS ® horizontal storage module and a NUHOMS ® AHSM-HS.

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DOE OSTI2022

Summary of Chloride-induced Stress Corrosion Cracking Experiments with a Spent Fuel Storage Canister Large Plate

This report provides a comprehensive summary of the crack growth rate testing and the experiments to evaluate the corrosion behavior of welded plate material taken from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The information includes that previously reported, and new work on detailed surface characterization of the corrosion attack from the large plate experiment. The work and the results inform aging management programs for the SNF canisters.

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