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DOE OSTI2025

Slow Strain Rate Testing of A537 Tank Wall Material

At Savannah River Site (SRS), High-Level Waste is stored in below-grade carbon steel tanks. This waste in part consists of sludge, salt cake, and/or supernate. Preparation of this waste for future processing involves dissolution of the salt cake layer. The salt dissolution process can create conditions that leave the carbon steel tanks susceptible to localized corrosion. The salt to be dissolved contains high concentrations of nitrate, that once released, create an environment that may be conducive to pitting corrosion and/or stress corrosion cracking (SCC) of carbon steel. The salt dissolution process also liberates interstitial liquid trapped between the salt crystals. This liquid is initially high in nitrite and hydroxide concentration. High pH and greater ratios of nitrite to nitrate act as inhibitors to minimize corrosion of carbon steel in high nitrate environments. However, as dissolution proceeds, the concentration of nitrate will increase, while the hydroxide and nitrite concentration of the interstitial liquid will deplete and become insufficient to prevent the onset of corrosion attack. Tank blending and the addition of inhibitors are used to ensure adequate concentrations of hydroxide and nitrite. However, this is not desirable during salt dissolution as it can reduce process efficiency and increase the amount of waste that needs processing. This testing program was designed to examine the risk of SCC associated with utilizing the pitting factor (PF) and nitrite/nitrate (NO 2 - /NO 3 - ) ratio limits for handling dissolved salt solutions at an elevated temperature in the carbon steel waste tanks. The previously identified limits are a PF of 1.2 and an NO 2 - /NO 3 - ratio of 0.15. The results indicate that as long as the NO 2 - /NO 3 - ratio exceeds 0.1 and the PF is above approximately 0.8, there is a discernible safety margin between the open circuit potential (OCP) and the critical cracking potential (CCP) observed during applied potential testing. However, this margin, defined by the difference between the OCP and CCP, is relatively narrow, ranging from 0.1 to 0.25 volts. This small margin raises concerns about potential shifts in OCP during waste retrieval operations, which could inadvertently increase the risk of SCC if the OCP approaches or exceeds the CCP. These results confirm that dissolved salt solutions provide a potent chemistry that, under certain conditions, makes carbon steel susceptible to SCC. The next question to consider is the influence these results have on decisions for storage and retrieval of waste from the tanks. For Type III/IIIA waste tanks, the risk of SCC remains very low. First, and most importantly, the post-weld stress relief of the tanks has reduced the residual stress near the welds. Thus, without the stress component, SCC risk is minimized. The material of construction (A537 Carbon steel) for the Type III/IIIA tanks is superior to the steel in its resistance to SCC than the steel that was utilized for the Type I, II, and IV tanks (A285 carbon steel). From a chemistry control standpoint for a Type III/IIIA tank directly involved with handling dissolved salt solutions, the PF and NO 2 - /NO 3 - ratio limits may be utilized wherein chemistry control provides an extra layer of defense against SCC. Chemistry control for a Type III/IIIA tank minimizes the risk for a tank that may receive the dissolved salt solution, particularly if that tank is a Type I, II, or IV waste tank. On the other hand, if the dissolved salt solution is handled by a Type I, II, or IV waste tank the risk of SCC is real. The potent chemistry, absence of stress relief, and inferior material result in a condition that is conducive to cracking. Efforts should be made to either avoid transferring waste that may not meet the PF and NO 2 - /NO 3 - ratio criteria to one of these tanks or if it is unavoidable, take measures to minimize the consequences of a leak. As shown by these tests, even if the PF and NO 2 - /NO 3 - ratio criteria are met, there is a risk that the tank potential may be disturbed in the positive direction and the risk of SCC increase.

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DOE OSTI2022

Multiphysics modeling of a critical dual-purpose canister in a saturated geological repository

Dual purpose dry cask storage canisters for spent nuclear fuel are designed for storage and transportation, but are not licensed for permanent disposal in a geological repository. If dual purpose canisters were to be used to dispose of spent nuclear fuel in a geological repository, they would be expected to eventually breach and be flooded with groundwater, and it is shown that some fraction of these canisters will achieve criticality. In order to evaluate the consequences of canisters going critical in a repository, an initial capability has been developed for estimating the quasi-static power level of a critical canister using loosely coupled multiphysics simulations. The low power level in a critical canister enables coupling through precomputed physics proxies. This calculated power level is then used to compute the change in the critical canister’s isotopic inventory as a function of time. Three as-loaded canisters are evaluated and two were found to have power levels below 4 kW, with a modest effect on the radiological inventory over time. This effort also shows that although some DPCs will have extremely peaked power shapes, the relatively low power and long-time scales result in relatively homogenized thermohydraulic properties in the water within the DPC.

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DOE OSTI2024

Considerations for Defining G-Values for Aluminum-Clad Spent Nuclear Fuel

Sealed-canister dry storage of aluminum-clad spent nuclear fuel (ASNF) generated by research reactors is an alternative to current storage and disposition pathways as directed by the U.S. Department of Energy. The major challenge faced for this storage approach is radiolytic H 2 generation, including from the aluminum (oxy)hydroxide layers on the surface of ASNF. Experimental and modeling activities have been carried out to characterize the radiolytic yield as part of a DOE-sponsored research program to develop the technical basis for ASNF dry storage. The G-value is a commonly way to report results of radiolysis testing and is defined as the radiolytic yield of a species (e.g. molecular hydrogen) per unit radiation energy deposited into the material system. An independent technical review of the ASNF dry storage technical basis performed by Pacific Northwest National Laboratory raised questions about differences in G-value definitions used for experiments on ASNF surrogates consisting of aluminum samples with adherent (oxy)hydroxides compared to G-values reported in prior literature and how the magnitudes compared between different studies. Material systems resembling ASNF pose complications for measuring/defining G-values to predict the evolution of H 2 in a sealed canister, including i) accounting for radiolytic yields potentially arising from multiple sources, i.e., residual free (vapor), physisorbed, and chemisorbed/chemically bound waters; ii) deciding what portions of the multi-material system to include in the absorbed energy (radiation dose) calculation, considering possible energy exchange between materials as well as measurement limitations, and iii) capturing variations in G-value associated with non-linear yield vs. dose curves and/or dependence on the cover gas. This report summarizes previous literature information on radiolytic H 2 generation and associated G-values from mixed-material systems (generally oxides in contact with water or organic compounds) and from (oxy)hydroxides/hydrates to compare with the definitions and values for ASNF surrogate samples containing adherent aluminum (oxy)hydroxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Continued Development and Advanced Testing of DPC Filler Cements (on FY22 R&D and Demonstration Activities) (Progress Report)

Commercial generation of energy by nuclear power plants in the United States (U.S.) has produced thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE). Utilities typically utilize the practice of storing this SNF in dual-purpose canisters (DPCs). DPCs were designed, licensed, and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a criticality event during SNF storage and transport, but were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019). There are several options being investigated that could facilitate the disposal of SNF stored in DPCs in a geologic repository (Hardin et al., 2015; SNL 2020b; SNL 2021b). These include: (1) repackage the SNF into canisters that are designed to prevent criticality during the regulated post-closure period following disposal, but with an increased disposal cost estimated at approximately $\$$20B in United States dollars (USD) (Freeze et al., 2019); (2) analysis of the probability and consequences of criticality from the direct disposal of DPCs during a 1,000,000-year post-closure period in several geologic disposal media (Price, 2019); and (3) filling the void space of a DPC with a material before its disposal that significantly limits the potential for criticality over the post-closure regulatory period. This report further investigates the third option, filling DPC already containing SNF with a material to limit the potential for criticality over the post-closure regulatory period.

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DOE OSTI2021

Tank Side Cesium Removal IXC-150 Cask Detonation Report

A set of experiments were conducted at Los Alamos National Laboratory’s firing point 88 to determine the blast pressure and effects of a stoichiometric gas mixture of Hydrogen (H 2 ) and Nitrous Oxide (N 2 O) on the IXC-150 Storage Pad Vent Stack assemblies. The Lab was asked to design, build, and test the vent stack assemblies using a reaction gas volume of 340 in 3 that mimics the Tank Side Cesium Removal (TSCR) Ion Exchange Column (IXC) assembly. The experimental system with vent stack assembly was designed with input from the Washington River Protection Services (WRPS) to ensure an accurate test setup. The tests were conducted at LANL on the newly built filter/gas assembly at the end of April 2021 timeframe, using WRPS supplied vent stack assemblies and internal HEPA filters. The objective of these tests was to observe that the structural integrity of the vent stack assemblies was maintained and determine the resulting blast overpressures at a distance of 16 and 32 inches from the assembly. All work was performed under the LANL Quality Assurance Program (SD330), using a graded application of ASME NQA-1- 2008/NQA-1a-2009. To ensure that all client quality assurance (QA) expectations were addressed, evaluation of the end data needs was performed and the appropriate controls applied for this work.

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DOE OSTI2020

Workshop to Plan R&D Support of Fuel/Basket Modification for Direct Disposal of Future DPCs

By 2030 about half of all spent nuclear fuel (SNF) arising from the current fleet of commercial power plants will be in dual-purpose canisters (DPCs), which are designed for storage and transportation but not for disposal. As an alternative to complete repackaging of the fuel for disposal, considerable cost savings and lower worker dose could be realized by directly disposing of this SNF in DPCs. The principal technical consideration is criticality control in a geologic repository, because the DPCs are large and depend on neutron absorbing basket components for criticality control. Neutron absorbing materials are generally aluminum-based, and under disposal conditions can degrade after a few hundred years contact with ground water. Simple modifications to the SNF assemblies or the DPC baskets could help to achieve direct disposal, and this is one of the approaches being studied to address the possibility of disposal criticality (SNL 2020a). Five fuel/basket modification concepts have been proposed (SNL 2020b) and a virtual workshop was conducted to solicit review and feedback on these concepts. The proposed solutions are: 1) zone loading of DPCs to limit reactivity, 2) replacing absorber plates with advanced neutron absorbing (ANA) material, 3) adding disposal control rods to pressurized water reactor (PWR) assemblies, 4) rechanneling boiling water reactor (BWR) assemblies with ANA material, and 5) basket insert plates (chevron inserts) made from ANA material. The presentations from the workshop are provided in this report, and the workshop discussions are summarized. This information includes prioritization of the proposed fuel/basket modification solutions, and prioritization of the associated model development, validation testing, and quality assurance activities. Information documented in this report will help to steer research and development efforts at Sandia National Laboratories, Oak Ridge National Laboratory, and Idaho National Laboratory that support the U.S. Department of Energy, Office of Nuclear Energy, Spent Fuel and Waste Science and Technology program

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DOE OSTI2022

Modeling of Radiolytic Hydrogen Generation of Irradiated Surrogate Aluminum Plates

The hydrogen generation from oxyhydroxide layers under irradiation is an important effect to understand for the long-term storage of aluminum-clad spent nuclear fuel. The original chemical model shows steady production of hydrogen from aluminum oxyhydroxide layers, with no roll-over effect in the curves of the hydrogen concentration until significant pressure has built up within the canister. This study uses a variety of kinetic fits to adjust the old chemical model to have roll-over effects that are consistent with new high-dose experiments. The data from aluminum 6061 coupon tests at high doses were used to fit several kinetic models. The 6061 aluminum samples are representative of the aluminum-clad fuel from the ATR. The fitted kinetic models were then used in a model of a mini-canister with much higher packing density to determine which kinetic rates would fit both sets of experimental tests. The rate constants for the kinetic mechanisms that were close were then tweaked such that the chemical models fit both the experimental data from the capsule tests and the data from the mini-canister tests. The kinetic models which fit the best are either inhibition by hydrogen, or a back reaction wherein H radical is reabsorbed onto the oxyhydroxide layer. The H back reaction rolls over quicker and may fit future collected data.

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DOE OSTI2020

Development of a Testbed for Pipeline Flushing - 20425

Pipelines which carry high-level waste within the DOE complex should be properly flushed in the cases where stationary or moving beds of solid sediment occur, or lines are prone to hydrogen gas buildup. Present guidelines establish a minimum for the amount of water (flush volume) and flush velocity values used for post-transfer flushing operations to achieve satisfactory cleaning of pipelines. However, further studies are needed to find optimal operating velocity modes/values in order to minimizing the flush volume and consequent downstream waste. These efforts are significantly helpful to DOE waste remediation sites by preserving tank storage, preventing additional waste processing, and minimizing dilution and changes in waste chemistry. An experimental test loop was recently developed for study of non-Newtonian slurry flushing at the Florida International University (FIU). This loop was designed to create sediment beds of various materials and bed heights and investigate parameters that effect the efficiency of flushing operations. The objective is to find flush velocity values/modes which lead to satisfactory cleaning of transport lines with a minimum amount of water usage. A 165 ft experimental test loop made of 3-inch carbon steel pipes was constructed and used in initial testing campaigns. This pipe loop was equipped with real-time monitoring instruments to record variation of pressure, density, mass flow rate, and temperature during testing. Post-flush in-situ no-flow evaluations were performed using ultrasonic, endoscopy, and visual inspection. Additional post-flush evaluations were conducted using filtration and density monitoring of residuals within a designated circulations loop. These provisions enable comparison of pipeline cleanness before the start and after termination of flushing operations for efficiency evaluation. This paper presents efforts associated with flushing of kaolin-water mixtures at different concentrations and with various flush volume and flush velocity values/modes. In initial testing, efforts focused on creating repeatable sediment beds inside the pipeline in both fully-flooded (no pre-flush drainage) and gravity-drained (pre-flush drained) systems. Bed characterization (i.e., measurement of sediment height and solids concentration) was conducted to ensure about consistency of initial conditions between tests. Preliminary results for flushing study of 10% vol. kaolin-water mixtures starting from a fully-flooded initial condition showed that flushing with 1-line volume (65 gallons) could remove up to 72 percent of solids from the pipeline. (authors)

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DOE OSTI2020

Estimation of Respirable Aerosol Release Fractions through Stress Corrosion Crack-Like Geometries

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the significant backfill pressurization of the canister, up to approximately 800 kPa. This pressure differential offers a relatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the carrier gas flow rates and aerosol transmission properties were evaluated for an engineered microchannel with characteristic dimensions similar to those of an SCC. The microchannel was formed by mating two gage blocks with a slot orifice measuring 28.9 μm (0.0011 in.) tall by 12.7 mm (0.500 in.) wide by 8.86 mm (0.349 in.) long (flow length). Surrogate aerosols of cerium oxide, Ce0 2 , were seeded and mixed inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to demonstrate a new capability to characterize SCCs under well-controlled boundary conditions. Separate modeling efforts are also underway that will be validated using these data. The test apparatus and procedures developed in this study can be easily modified for the evaluation of more complex SCC-like geometries including laboratory-grown SCC samples.

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DOE OSTI2021

Mechanical Loads on Spent Nuclear Fuel in the General 30 cm Package Drop Scenario

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) research program is guided by the high-level goal of closing prioritized knowledge gaps related to spent nuclear fuel (SNF) storage and transportation, which are summarized by Saltzstein et al. (2020). One of the high-priority knowledge gaps is the identification and quantification of mechanical loads that are expected to affect SNF during normal conditions of transportation and storage to inform the range of physical SNF test programs. This report uses modeling and analysis methods to estimate the mechanical loads on spent nuclear fuel (SNF) in the general 30 cm package drop scenario. The drop scenario assumes impact limiters are in place in the transportation configuration and the impact surface is perfectly rigid. The goal of this analysis is to consider the universe of potential mechanical loading conditions that can happen to SNF and present the results in a manner that is useful for materials testing, decision making, and regulatory rule making purposes. This study uses validated finite element models and methods to perform a broad parametric study of key variables that can affect the mechanical loads on SNF during a hypothetical package free drop scenario. Physical drop test data from a cask and fuel assembly drop test campaign is the basis for model validation. Additionally, the results of the parametric study are used to inform a damage model, which uses multiple nonlinear regression to estimate the relationships between input variables and output response. The parametric finite element analyses consider thousands of input variable combinations, while the damage model estimates millions of combinations. The breadth of this study provides confidence that the potential range of mechanical loads that SNF might experience during the general 30 cm package drop scenario are characterized well enough to consider this knowledge gap closed. While this report documents the overall peak values calculated in this study, the 95 th percentile values, the histograms, and the observed trends are equally important. This study covered a large range of SNF temperatures, room temperature to 300°C, and burnups, 10 GWd/MTU to 62 GWd/MTU. Each temperature and burnup combination has a different cladding yield strain, so it is more meaningful to summarize the calculated cladding strain response as its factor of safety, which is defined relative to the yield strain. The factor of safety is calculated as the yield strain divided by the peak cladding strain. A factor of safety greater than unity indicates that the cladding remains below yield, whereas a value less than unity is indicative of plastic deformation. In all cases of this study a safety factor over 1.0 was calculated, although in the most limiting case at 300°C the safety factor was only 1.01, which suggests that yielding could occur when additional loads like rod internal pressure are included. When the temperature is restricted to 200°C the limiting safety factor increases to 1.28, which has significant margin to accommodate internal pressure and potential local cladding defects that could cause a local stress concentration. An important trend in the calculated fuel rod mechanical loads is that the 2 nd highest loaded fuel rod in an assembly tends to be significantly lower than the highest loaded rod. The implication is that even if one rod in an assembly experiences a failure the loads would have to be significantly higher to cause two or more rods to fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Amphiphilic Block Copolymers for Flocculation and Hydrophobization of Legacy Waste Suspensions in Flotation Driven Dewatering Operations

There exists a degrading legacy fuel pond at the Sellafield site (UK), known as the First-Generation Magnox Storage Pond (FGMSP). After being placed into a passive care and maintenance regime resulting in a long storage period, the fuel rods, primarily their cladding, corroded in the pond forming magnesium hydroxide based sludges and suspended material. These ponds are a grave concern to the British government in terms of hazard and risk reduction. Therefore, decommissioning of these ponds is a top priority beginning with the contaminant retrievals process [1]. The pond has accumulated significant quantities of waste materials amongst the skips of fuel, including but not limited to: large inventories of corroded Magnox sludge, fuel rod fragments, metal fragments (from fuel skips), concrete degradation products (from the pond infrastructure), wind-blown sand, and other materials such as bird guano and animal remains. This challenge requires a chemically robust technology to complete sludge retrievals [2]. As nuclear is different, a more stringent process operation criteria is required. From the criteria, flotation was selected as a viable technology. Flotation involves the application of collector molecules which modify the hydrophobicity of suspended particles allowing them to adsorb to rising bubbles. As some particles lack inertia for flotation due to their size, dual flocculant/collector agents can be deployed for greater particle recovery, in this case, amphiphilic diblock copolymers. Two copolymers of Poly(acrylic acid)-b-poly(n-butyl acrylate) (or PAA-b-PnBA), of different hydrophobic chain lengths (PBA) were synthesized for flotation campaigns. Flotation has shown promise to be a valuable rapid dewatering strategy for decommissioning of legacy waste ponds whilst upholding the required operational criteria. This research has shown: PAA-b-PnBA copolymers promote flocculation of Mg(OH){sub 2} particulates. Diblock copolymers retain more fluid than traditional surfactant based collectors wrt. relative particulate recovery performance- where longer PnBA chain length polymers performed best. Performance appears to be hydrodynamically limited due to over-flocculation of Mg(OH){sub 2}. Potential to combine with sedimentation for high particle recovery with low fluid carryover.

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DOE OSTI2023

Characterization of Infrequent Samples from The Concentration, Storage, And Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: February 26, 2023, Sample

Savannah River Mission Completion Engineering (SRMC-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma activities, total alpha activity, free hydroxide, and pH measurements. This report presents characterization results for the leak detection box (LDB) February 26, 2023, drain cell sample. The sample was clear and colorless with no visible particulates. The results are measurements for total gamma, total alpha, total beta, free hydroxide, pH, and density. These analyses were performed in triplicate. A summary of the average analytical results for the LDB sample includes the following. The directly measured pH for the LDB February 26, 2023 "as-received" drain cell sample range was 7.12-7.16, and the free hydroxide concentration was <0.02 M. The density of the "as-received" drain cell sample determined at 25 °C was 1.02±0.00 g/mL. The total alpha activity for the LDB February 26, 2023, sample is reported as a less than value (Upper Limit) because of possible spectral interferences. Thus, the total alpha activity averaged <3.52E+02 dpm/mL. This value is less than 4.83E+03 dpm/mL, which is the procedural limit for non-waste determination. The total beta activity in the LDB February 26, 2023, drain cell sample is above the instrument detection limits and average 9.76E+03±5.94E+02 dpm/mL. The total beta activity in the LDB February 26, 2023, drain cell sample is above he instrument detection limits and averaged 7.76E+03±5.94E+02 dpm/mL. The average measured cesium-137 activity (dominant beta emitter) in the LDB February 26, 2023, drain cell sample is 7.76E+03±2.35E+02 dpm/mL. The corresponding Ba-137m (dominant gamma emitter) activity, calculated as 94.6% of the Cs-137 values, is 7.34E+03±2.23E+02 dpm/mL. The total empirical activity of the beta and gamma emitting (represented by the sum of total beta and Ba-137m activities) averaged 1.71E+04±8.18E+02 dpm/mL. This value is less than 8.69E+05 dpm/mL, which is the procedural limit for non-waste determination.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

MELCOR Input Model for Spent Fuel Transportation and Storage Canister

This report documents the progress and current results of the MELCOR spent fuel cask input model. The MELCOR model is being developed to investigate aerosol transport and deposition given the aerosol physical phenomena models within MELCOR. To perform the analyses, a general portrayal of the MAGNASTOR® cask system has been employed; however, this system was selected based on available information to provide a reasonable representation of a spent fuel cask. The analytical results are not intended to characterize the performance of the MAGNASTOR® cask. Instead, the provided results are intended to enhance our general understanding of the aerosol behavior within casks and the validity of current models. The current model efforts are being performed to investigate hypothetical UO 2 release from failed fuel pins within a spent fuel cask. The existing MELCOR model of the MAGNASTOR® cask system has been adapted to permit future comparative analyses with the GOTHIC representation of the MAGNASTOR® cask. To support this comparison, the PNNL model characteristics that are unrelated to the aerosol modeling were applied to the MELCOR model. These characteristics included improved comparability of the axial fidelity, total spent fuel power, fuel pin axial power profile, and heat losses from cannister. The thermal-hydraulic solutions are improved within the capability of the MELCOR code and will permit better overall agreement with the GOTHIC results. Detailed results are presented on the thermal-hydraulic analysis of the MELCOR cask as well as characterization of UO 2 aerosol dispersion and deposition within the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY2020 Status Report)

This progress report describes work performed during FY20 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask 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. Work in FY20 further defined our understanding of the potential chemical and physical environment present on canister surfaces, evaluated the relationship between the environment and the resultant corrosion that occurs, and initiated crack growth rate testing under relevant environmental conditions. In FY20, work to define dry storage canister surface environments included several tasks. First, collection of dust deposition specimens from independent spent fuel storage installation (ISFSI) site locations helped to establish a more complete understanding of the potential chemical environment formed on the canister. Second, the predicted evolution of canister surface relative humidity RH) values was estimated using ISFSI site weather data and the horizontal canister thermal model used by the SNL probabilistic SCC model. These calculations determined that for typical ISFSI weather conditions, seasalt deliquescence to produce MgCl2-rich brines could occur in less than 20 years at the coolest locations on the canister surface, and, even after nearly 300 years, conditions for NaCl deliquescence (75% RH) are not reached. This work illustrates the importance of understanding the stability of MgCl2-rich brines on the heated canister surface, and the potential impact of brine composition on corrosion processes, including pitting and stress corrosion cracking. In an additional study, the description of the canister surface environment was refined in order to define more realistic corrosion testing environments including diurnal cycles, soluble salt chemistries, and inert mineral particles. The potential impacts of these phenomena on canister corrosion are being evaluated experimentally. Finally, work over the past few years to evaluate the stability of magnesium chloride brines continued in FY20. MgCl2 degassing experiments were carried out, confirming that MgCl2 brines slowly degas HCl on heated surfaces, converting to less deliquescent magnesium hydroxychloride phases and potentially leading to brine dryout.

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DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

DOE manages over 300 types of SNF, most of which are located at the INL site. “Road-ready dry storage” is a management concept where SNF is packaged into dry, sealed canisters, which are then placed in on-site storage in anticipation of later removal. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE SNF for road-ready dry storage. In support of establishing a large-scale road-ready dry storage program at the INL site, the demonstration will package a select amount of DOE-managed SNF into DOE Standard Canisters. The closure process for the DOE Standard Canisters will include fuel and basket loading, welding, inspection, leak testing, and if needed, repair. As a follow-up to previous discussion on the design of the DOE Closure Leak Test Assembly, this report describes recent fabrication and testing efforts performed at INL. DOE Standard Canisters are sealed by two sequential gas tungsten arc welds. Both are performed by remotely operated and semi-autonomous welding systems. The first weld is a circumferential pipe weld that completes assembly of the canister body and lid assembly. The second and final closure weld connects the vent plug to the vent port with an identical butt joint to the circumferential pipe weld. After the second weld is performed on the vent port, these welds are helium leak tested using an inside-out technique. In addition to the commercially available vacuum and leak detector systems, the DOE Standard Canister Closure Leak Test Assembly was designed for both remote and manual operation. This report describes fabrication and performance testing associated with the inside-out technique. INL staff designed and tested systems to accomplish these tasks. Hardware fabrication occurred at INL facilities. Forthcoming work includes design optimization, integration to existing systems, and implementation to packaging demonstration operations.

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DOE OSTI2019

Los Alamos National Laboratory SAVY-4000 Field Surveillance Plan Update for 2020

The Packaging Surveillance Program section of the Department of Energy (DOE) Manual 441.1-1, Nuclear Material Packaging Manual (DOE 2008), requires DOE contractors to “ensure that a surveillance program is established and implemented to ensure the nuclear material storage package continues to meet its design criteria.” The Los Alamos National Laboratory (LANL) SAVY-4000 Field Surveillance Plan was first issued in fiscal year (FY) 2013. The surveillance plan is reviewed annually and updated as necessary based on SAVY-4000 surveillance findings, as well as results of the lifetime extension studies. Five surveillance plan updates have been issued, one in 2014, one in 2016, one in 2017, one in 2018, and one in 2019. This 2020 update documents what was actually done in 2019 and what is planned for 2020. Deviations from the 2019 surveillance plan were necessary because some of the planned surveillance containers were not available for examination. The goal of the surveillance program is to provide data to: evaluate container component performance over time against the initial baseline qualification; validate the current SAVY container design life; provide data for future design life extensions of components as indicated; identify indications of container degradation early and identify similar containers and/or contents that may need remediation; maintain and continuously improve the surveillance information system that documents the complete life cycle of the containers including the initial production baseline, usage history, surveillance, and disposition; identify container issues related to wear and tear that could indicate the need for a maintenance program for in-service containers prior to the end of the design life.

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DOE OSTI2022

Complete Initial Evaluation of Novel Complexants for Tc Holdback for Simplified Single Cycle Separations.

Management of technetium at the back end of nuclear fuel cycle is complicated by its unique physico-chemical character, but must be addressed to due to the environmental, storage and reprocessing challenges of this fission product. In a single-cycle scenario, the recovery and purification of uranium is particularly challenging due to the tendency of a pertechnetate anion to follow uranyl cation. This report summarizes initial studies directed towards finding the solution for technetium management. Two options for controlling technetium’s chemistry in solvent extraction scenarios were studied. Molecular recognition of pertechnetate by aqueous complexants based on guanidinium moieties was compared to redox manipulation of technetium using dihydrazide reagents. Solvent extraction, potentiometric and nuclear magnetic resonance spectroscopy studies were performed. The influence of guanidinium reagents on the liquid-liquid partitioning of technetium is less pronounced, relative to dihydrazides. Strong impact exerted by dihydrazides originates from the technetium-catalyzed decomposition of such compounds which reduces technetium to a tetravalent, non-extractable state. Although this route of technetium management is very effective the destruction of dihydrazides is undesirable. Guanidinium complexants show a more tempered influence on technetium. The effect on the liquid-liquid partitioning of pertechnetate is evident and likely guided by anion recognition due to ion-pair and hydrogen bond formation. Guanidiniums are stable in presence of technetium as evidenced by nuclear magnetic resonance studies. The steady state chemistry of guanidiniums identify this class of aqueous complexants as solid candidates for structure-function pertechnetate recognition studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

FY2022 Status Update: A Probabilistic Model for Stress Corrosion Cracking of SNF Dry Storage Canisters

Understanding the potential risk of stress corrosion cracking of spent nuclear fuel dry storage canisters has been identified as a knowledge gap for determining the safety of long-term interim storage of spent nuclear fuel. To address this, the DOE is funding a multi-lab DOE effort to understand the timing, occurrence, and consequences of potential canister SCC. Sandia National Laboratories has developed a probabilistic model for canister penetration by SCC. This model has been continuously updated at SNL since 2014. Model uncertainties are treated using a nested loop structure, where the outer loop accounts for uncertainties due to lack of data and the inner aleatoric loop accounts for uncertainties due to variation in nature. By separating uncertainties into these categories, it is possible to focus future work on reducing the most influential epistemic uncertainties. Several experimental studies have already been performed to improve the modeling approach through expanded process understanding and improved model parameterization. The resulting code is physics-based and intended to inform future work by identifying (1) important modeling assumptions, (2) experimental data needs, and (3) necessary model developments. In this document, several of the sub-models in the probabilistic SCC model have been exercised, and the intermediate results, as the model progresses from one sub-model to the next, are presented. Evaluating the sub-models in this manner provides a better understanding of sub-model outputs and has identified several unintended consequences of model assumptions or parameterizations, requiring updates to the modeling approach. The following updates have been made, and future updates have been identified.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗