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Research Overview: Canister Stress Corrosion Cracking
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Technology Development of a High-Capacity High-Assay Low Enriched Uranium Transportation Concept
This paper discusses the technology development (TD) efforts that led to the development of a High-Capacity High-Assay Low Enriched Uranium Transportation (HALEU) transportation concept. In 2018, the Department of Energy (DOE) Office of Nuclear Technology and Research Development tasked Idaho National Laboratory (INL) to investigate strategies to transport large quantities of HALEU. To complete this task, INL collaborated with Pacific Northwest National Laboratory and Oak Ridge National Laboratory. The project was completed in 2020, and one of the project outcomes was a transportation concept that consisting of five individual Type B packages transported on a single legal-weight truck (LWT). The total payload capacity of this concept is 1,881 kg (4,149 lb) of HALEU in the form of uranium dioxide (UO2) powder. The concept utilizes an existing Type B packaging design carrying a novel fuel basket design with an incorporated flux trap. The basket can be loaded with 18 individual fuel canisters. The research collaboration investigated the U.S. certification potential of this concept. This part of the project included evaluations of criticality safety, radiological safety, thermal safety, structural integrity, and confinement under hypothetical accident scenarios of transport. The results of these evaluations demonstrated a promising potential for U.S. certification of this concept. Eventually, the described efforts led to the pursuance and issuance of a U.S. patent, thus, protecting the associated intellectual property (IP). Current short-term goals include making this IP available to private industry partners for licensing, directly supporting DOE’s objectives of accelerating commercialization of national laboratory-generated IP. If additional funding becomes available, long-term research goals could include exploring the feasibility of transporting other uranium chemical forms (e.g., UF4) with this concept, or refining operational procedures to load or unload the packagings.
Milestone 1.2.12: The Fate of Oxygen Radical Species in Corroded Aluminum Alloys under Irradiation
The U.S. Department of Energy tasked Idaho National Laboratory to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in heliumbackfilled canisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen (H 2 ) generation from corrosion layers that would be present on the fuel assembly surfaces. However, limited attention has been directed to the oxygen radical species that are concurrently generated in these H 2 producing radiation environments. This report collates and summarizes the available experimental and computational studies on the fate of these reactive oxygen radicals during the irradiation of aluminum oxyhydroxide and hydroxide polymorphs typically formed on ASNF. Based on this review, it is determined that radiation-induced oxygen-centered radicals do not react to give molecular oxygen gas (O 2 ), but rather remain trapped in the aluminum corrosion layers in their original crystal lattice positions. Overall, O 2 generation will not be a concern for extended storage of ASNF in helium-backfilled canisters. That is, no O 2 contribution to pressurization of sealed dry storage systems or creation of a flammable atmosphere is anticipated.
Non-destructive evaluation and machine learning methods for inspection of spent nuclear fuel canisters: A state-of-the-art review
Nuclear energy is among the cleanest and most efficient energy sources currently available. The operation of nuclear power plants (NPPs) produces large amounts of high-level radioactive waste known as spent nuclear fuel (SNF). Currently, large amounts of SNF is stored in dry cask storage systems (DCSSs) for extended interim storage until a permanent disposal solution becomes available. During the extended interim storage, the DCSS, particularly the SNF canisters, may degrade and abnormal conditions may occur. Therefore, non-destructive evaluation (NDE) and machine learning (ML) approaches are necessary for inspection of SNF canisters. This paper presents a state-of-the-art review of literature by summarizing recent progress made on the applications of NDE and ML for inspection of SNF canisters. Sixteen NDE methods are examined and compared: visual inspection, ultrasonic guided waves (UGWs), laser-based approaches, acoustic emission (AE), eddy current testing (ECT), non-invasive acoustic sensing, dynamic modal testing, cosmic ray muons tomography, neutron imaging, gamma rays detection, fiber optical sensors, through-wall communications, X-ray computed tomography (CT), vibrothermography, monoenergetic photon sources, and surface acoustic wave (SAW) sensors. The technology readiness level (TRL) for each method is assessed and compared. Recent publications on ML-enhanced visual inspection, AE, non-invasive acoustic sensing, dynamic modal testing, and neutron imaging for SNF canisters are summarized and future research needs are identified. In conclusion, this review article provides a convenient reference on the state-of-the-art applications of NDE and ML methods for inspection of SNF canisters.
ADDITIVE MANUFACTURING AND REGULATORY TESTING OF CANISTERS FOR SPENT NUCLEAR FUEL MANAGEMENT
The back end of the fuel cycle focuses on the interim storage, transportation, and final disposition of the spent nuclear fuel from nuclear reactors. Commercial light water nuclear power stations across the United States operate with fuel assemblies being irradiated for up to 6 years (planned) in the reactor pressure vessel. After their planned irradiation, the fuel assemblies are moved to a spent fuel pool within the facility complex. After the spent nuclear fuel is removed from the fuel pool, it is typically inserted into a welded metal canister that can be transferred between overpacks for storage, transportation and possibly disposal. Most dry storage systems being used by industry today use dual purpose canisters (DPCs), designed for use in storage and transportation overpacks, but not specifically designed for disposal. Triple purpose-canisters, designed for disposal in addition to storage and transportation, have also been researched. Traditional manufacturing methods for spent fuel canisters involve fusion welding along the length or circumference of the canister which results in high tensile residual stresses in the joint weld zone (WZ) and heat affected zone (HAZ). In this paper, spent fuel canister designs were printed by wire arc additive manufacturing (AM) using the 316L SS welding wire to demonstrate: 1. Feasibility of spent nuclear fuel canister fabrication using this advanced manufacturing method, 2. Dynamic response of the additive manufacture canister design when subjected to the federally mandated Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) physical tests for Type B packages. This paper will focus on the canister printing design and structural tests. The AM 3D printed design, regulatory testing, and post-test evaluation of the canister tested to the 10 CFR 71.71 and 10 CFR 71.73 requirements will be presented. One AM canister design was subjected to the penetration, free drop, and puncture test. Before and after the dynamic structural tests, the AM canister design was scanned with a handheld scanner to capture a 3D CAD geometry to compare to the 3D printed canister design in the deformed shape. The scanned geometry was sectioned in areas with deformation and the cross-section profile was measured to determine accurate and repeated results of the deformed shape of the AM canister design.
The Hanford Lead Canister Collaboration
The Hanford Site, Management of the Cesium and Strontium Capsules (MCSC) Project is preparing to place cesium and strontium material into dry storage using stainless steel canisters that are housed inside vertical concrete casks. Part of the site’s aging management program is to use a spare canister system as a leading indicator canister to provide advance warning for signs of chloride induced stress corrosion cracking (CISCC) or any other kind of canister degradation that might take place over the system’s 300-year design life. The Hanford Lead Canister (HLC) will use electric heaters to simulate the decay heat of cesium and strontium to ensure the HLC simulates the environment and conditions the capsule-loaded systems will experience. In addition to being part of the site’s aging management program, the HLC is being made available for research and development that advances the industry regarding canister mitigation and repair (M&R). The HLC system closely resembles current spent nuclear fuel (SNF) dry storage systems, making the HLC useful for demonstrating technologies or performing research that is relevant to SNF canisters . Several organizations are collaborating to prepare the HLC, along with technical strategy, to perform the needed research and development activities, to include the Pacific Northwest National Laboratory (PNNL), Central Plateau Cleanup Company (CPCCo), and the Electric Power Research Institute (EPRI). In one example of this collaborative research, EPRI is planning to use the HLC to demonstrate in-situ M&R technologies under controlled conditions before the HLC is deployed at Hanford. There is a period of several years when the HLC will be available for canister M&R activity before it is deployed. This paper describes the development of the HLC, its role in the Hanford site’s MCSC Project aging management program, and some of the research activities that are planned for the HLC.
Implementation of the Glycolate Destruction Process with Sodium Permanganate in the Defense Waste Processing Facility (DWPF) - 20141
The Savannah River Site (SRS) seeks to replace formic acid with glycolic acid as a chemical reductant in the Defense Waste Processing Facility (DWPF), where borosilicate glass is mixed with high-level radioactive waste, melted at high temperatures, then poured into stainless steel canisters for safe storage. Prior research and development have demonstrated the feasibility and advantages of the new nitric-glycolic acid flowsheet over the current nitric-formic acid flowsheet to chemically adjust the radioactive sludge slurry waste in the Sludge Receipt and Adjustment Tank (SRAT) prior to vitrification. Use of glycolic acid reduces hydrogen and ammonia generation during the reduction process, thus reducing flammability hazards and purge requirements in the DWPF processing vessels. In addition, glycolic acid will support the new Salt Waste Processing Facility (SWPF) by allowing for receipt of higher volumes of waste due to reduction of flammable gases, increased boil-up rates in the vessels, and easing the processing of future sludge batches containing high levels of mercury. However, during DWPF operations, trace amounts of glycolate are anticipated to be entrained in the recycle stream and sent to the Concentration, Storage, and Transfer Facilities (CSTF) via the Recycle Collection Tank (RCT). The RCT receives condensate from the Slurry Mix Evaporator Condensate Tank (SMECT) and Off-gas Condensate Tank (OGCT) that may contain unreacted glycolate. A literature review found the potential for hydrogen generation due to thermolysis of glycolate in caustic CSTF conditions. Savannah River National Laboratory (SRNL) conducted several tests on simulant and radioactive waste from various tanks at SRS and confirmed that glycolate added to waste at concentrations of approximately 1000 mg/L and higher generates hydrogen through thermolysis. There are potential operating scenarios in the future where glycolate concentration could exceed 1000 mg/L in the 2H Evaporator Drop Tank, the highest accumulation point for DWPF organics received via recycle. These findings presented a need to develop another process to mitigate the impact of glycolic acid to the CSTF. Several DWPF compatible options were evaluated for their ability to destroy glycolate in the DWPF recycle stream under various processing conditions, thereby mitigating its introduction to the CSTF. Downselection testing identified sodium permanganate as the most promising and simplest option to oxidize glycolate. It has been demonstrated to be effective in various RCT simulant compositions, operating temperatures, oxidant addition rates, solution pH, and initial glycolate concentrations. A feasibility study identified existing tanks within the facility that could be used for receiving and storing the commercially available sodium permanganate, as well as feeding it to the RCT for processing. The use of existing equipment as well as the ability to purchase the pre-mixed chemical reduces implementation and operational complexity. The downstream impacts of the sodium permanganate reaction were also evaluated and demonstrated to have minimal impact on other chemical species present in the RCT and CSTF, as well as on the materials of construction of the RCT and CSTF vessels and associated piping and instrumentation. Future work is in progress to validate the glycolate destruction process with testing in actual radioactive RCT waste before implementation in DWPF. (authors)
THERMAL STRESS ANALYSIS OF A SPENT NUCLEAR FUEL CANISTER
The potential for chloride induced stress corrosion cracking (CISCC) in spent nuclear fuel dry storage canisters is a current topic of research by the US Department of Energy Spent Fuel and Waste Science and Technology program. One of the important prerequisites for CISCC is a tensile residual stress state. This study utilizes computational models to provide an initial analysis of thermal stresses in a generic spent nuclear fuel canister. A STAR-CCM+ thermal fluid model provides a temperature profile analysis of the canister based on four different ambient temperature conditions. An ANSYS APDL finite element model incorporates the temperature profiles to analyze the thermal stresses in the canister. The calculated thermal stress magnitudes are in the range of 10 MPa to 80 MPa, which could be significant for crack propagation through the canister wall. The maximum thermal stress is not a concern for the structural failure of the canister, but it is high enough that it is a significant contribution to the total stress state of the canister wall, which also includes residual stress from fabrication, internal pressure from helium cover gas, and potential transient mechanical loads, such as earthquakes. This paper describes the initial finite element thermal stress analysis that was completed in 2020, reports the initial findings, and identifies areas where model refinement is still needed to complete this work in the future.
DOE?s Disposal Research Activities: Geologic Disposal Safety Assessment (GDSA) and Direct Disposal of Dual-Purpose Canisters (DPCs).
Abstract not provided.
Magnetostrictive EMAT for Nuclear Spent Fuel Canisters
One approach for mitigation of stainless steel (SS) spent fuel canisters (SFCs) is to coat the welds or even the entire SFC with a cold-sprayed nickel coating. The coating can serve as a magnetostrictive interface between an electromagnetic acoustic transducer (EMAT) and the SS plate material. The EMAT can send ultrasound into the canister and “listen” for reflections caused by chloride-initiated stress corrosion cracking (CISCC) or other damage mechanisms that may lead to leaks if undetected and unmitigated. Here, this project outlines an approach for monitoring the integrity of SFCs using magnetostrictive EMAT sensors, which emit and detect discontinuity-initiated reflections indicative of pits or cracks likely arising from CISCC. Flat 13 mm SS plates with welds similar to actual canister welds were prepared, with flat-bottom holes 3 mm and 6 mm deep and 6 mm in diameter placed within the weld heat-affected zone (HAZ). The most effective sensor configurations were able to detect critical discontinuities more than 2 m away. Although a single sensor could be robotically deployed, the large volume covered by each sensor could support a small number of permanently mounted sensors, equipped with an external connector for periodic interrogation. The HAZ of a hypothetical SFC cylindrical surface could be effectively monitored with just 16 sensors.
Review and Summary of Oxide Thickness Data for Aluminum-Clad Spent Nuclear Fuel
The U.S. Department of Energy (DOE) owns a large inventory of aluminum-clad spent nuclear fuel (ASNF) in interim storage pending ultimate disposition, with more being generated by currently operating research reactors. Dry storage in sealed DOE Standard Canisters is being investigated as an approach for long-term interim storage and/or disposition in a repository for ASNF. A primary challenge for ASNF storage is the presence of aluminum (oxy)hydroxide layers formed on the cladding surfaces during water exposure in the reactor and in wet storage, which forms a reservoir of chemisorbed water not readily removed at low (<100°C) drying temperatures. Free, physisorbed, and chemisorbed water are all susceptible to radiolytic breakdown under irradiation and can release hydrogen gas. Identifying the likely range of (oxy)hydroxide loadings on ASNF that may be placed in dry storage will help to ensure that the impact of the (oxy)hydroxide is adequately accounted for while avoiding over conservatism and enable mitigation strategies to be implemented where needed. The current report summarizes information on (oxy)hydroxide thicknesses and characteristics from the literature and from recent measurements taken in the present research campaign. In general, corrosion studies have indicated that the corrosion and oxide buildup on aluminum are affected by numerous conditions, including pH, temperature and heat flux, coolant flowrates, irradiation (in-reactor vs. unirradiated tests), duration of water exposure, and the amount of oxide already on the surface. Some of these factors are interrelated (e.g., the local temperature, heat flux, and coolant flowrate). Observations of the impact of irradiation (in-reactor versus ex-reactor measurements) on corrosion and oxide thickness are mixed, but multiple studies indicate that there is a significant difference between in-reactor and ex-reactor corrosion kinetics, even when other operating conditions such as pH, flowrate, heat flux, etc., simulate those in a reactor.
Features Events and Processes Relevant to DPC Disposal Criticality Analysis
The Department of Energy is evaluating the technical feasibility of disposal of spent nuclear fuel in dual-purpose canisters in various geologies. As part of ongoing research and development, the effect of potential post-closure criticality events on repository performance is being studied. Many different features, events, and processes (FEPs) could affect the potential for criticality or the extent of a criticality event. Additionally, a criticality event could affect other FEPs. This report uses existing lists of FEPs as a starting point to evaluate the FEPs that could affect or be affected by an in- package criticality event. The evaluation indicates that most of the FEPs associated with the waste form, the waste, or the EBS have some effect on post-closure criticality and/or are affected by the consequences of post-closure criticality. In addition, FEPs not previously considered are identified for further development.
SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY21 Status Report)
This progress report describes work performed during FY21 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of canister materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In FY21, modeling and experimental work was performed that further defined our understanding of the potential chemical and physical environment present on canister surfaces at both marine and inland sites. Research also evaluated the relationship between the environment and the rate, extent, and morphology of corrosion, as well as the corrosion processes that occur. Finally, crack growth rate testing under relevant environmental conditions was initiated.
Assessing the release, transport, and retention of radioactive aerosols from hypothetical breaches in spent fuel storage canisters
Interim dry storage of spent nuclear fuel involves storing the fuel in welded stainless-steel canisters. Under certain conditions, the canisters could be subjected to environments that may promote stress corrosion cracking leading to a risk of breach and release of aerosol-sized particulate from the interior of the canister to the external environment through the crack. Research is currently under way by several laboratories to better understand the formation and propagation of stress corrosion cracks, however little work has been done to quantitatively assess the potential aerosol release. The purpose of the present work is to introduce a reliable generic numerical model for prediction of aerosol transport, deposition, and plugging in leak paths similar to stress corrosion cracks, while accounting for potential plugging from particle deposition. The model is dynamic (changing leak path geometry due to plugging) and it relies on the numerical solution of the aerosol transport equation in one dimension using finite differences. The model’s capabilities were also incorporated into a Graphical User Interface (GUI) that was developed to enhance user accessibility. Model validation efforts presented in this paper compare the model’s predictions with recent experimental data from Sandia National Laboratories (SNL) and results available in literature. We expect this model to improve the accuracy of consequence assessments and reduce the uncertainty of radiological consequence estimations in the remote event of a through-wall breach in dry cask storage systems.