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DOE OSTI2023

Evaluation Of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

Per a directive from the Department of Energy Savannah River Operations Office (DOE-SR) in 2008, the fissile mass loading concentration must remain below 897 g/m 3 in each high-level waste (HLW) glass canister produced by the Defense Waste Processing Facility (DWPF). To support Sludge Batch 5 (SB5) processing, the Savannah River National Laboratory (SRNL) developed a technical basis that facilitates the evaluation of fissile mass loading of the glass product. The calculation is based on the iron (Fe) concentration in the glass as determined by measurements from the Slurry Mix Evaporator acceptability analysis as well as the glass density. In April 2022, a subsequent DOE-SR directive increased the fissile mass loading limit to 2500 g/m 3 beginning with Sludge Batch 11. Thus, the 897 g/m 3 limit still applies to Sludge Batch 10 (SB10) processing. For SB5 through initial Sludge Batch 9 (SB9) processing prior to coupled operation with the Salt Waste Processing Facility (SWPF), SRNL provided DWPF a bounding glass density value that was based on a statistical evaluation of density measurements. To eliminate the need for experimental work, a composition-based density model for HLW glasses was developed at SRNL in 2019. The objective of this report is to present the bounding glass density determined with the composition-based density model for SB10 sludge-only (SO) and coupled processing.

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DOE OSTI2023

Evaluation of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

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DOE OSTI2022

GDSA - Modeling and Integration – ORNL: PFLOTRAN modeling

This report summarizes work done at Oak Ridge National Laboratory (ORNL) related to post-closure criticality consequence analysis of dual-purpose canister (DPC) direct disposal in unsaturated alluvium. PFLOTRAN software was used to model the behavior of a breached DPC with low water infiltration rates and unsaturated conditions in alluvial deposits such that a low-power criticality event occurs with a focus on thermal hydrological constraints on criticality timing and power output. This summary describes initial exploration of those thermal hydrological processes and of the long-term average power output from a criticality event that could be sustained without driving water out of the DPC package and thereby terminating the criticality event. The objectives are (1) to identify, for different groundwater infiltration rates, the time at which the DPC fills with water sufficiently to potentially initiate criticality; (2) to bound the power output that could be produced by a criticality event that would thermally drive water out of the package at various percolation rates; and (3) provide an estimate of how the size of a package breach affects the water infiltration and therefore the possibility timing of criticality.

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DOE OSTI2023

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.

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DOE OSTI2023

Effects of Mineral Impurities and Heat on Uranium(VI) Sorption onto Bentonite (Final Report)

Most nuclear waste disposal options include compacted bentonite, consisting of montmorillonite clay, as part of a barrier system to minimize contaminant mobility. Uranium (U) is the primary element in spent nuclear fuel, and from an environmental perspective, a potential contaminant of water resources. Furthermore, uranium may control nuclear fuel degradation rates and the consequent release of other radioactive contaminants based on its diffusive transport away from waste canisters. Uranium sorption onto clay and other mineral surfaces is expected to limit U(VI) mobility in these systems. However, at this point a prediction of U(VI) sorption and transport behavior in performance assessment (PA) models is complicated by a series of factors, such as: (1) the presence of various U(VI) solution species with different charges and sorption characteristics; (2) the complex microstructure of montmorillonite clay resulting in two types of clay porosities and multiple sorption sites; (3) the largely unknown effects of bentonite mineral impurities on pore water chemistry, U(VI) sorption and diffusion behavior; and (4) the potential impacts of heat, generated by the decay of spent fuel, on mineralogical and microstructural transformations, and any subsequent effects on radionuclide sorption and mobility. In this project, we investigated the effects of calcite impurities on U(VI) sorption and diffusion onto montmorillonite before and after mineral exposure to heat. Overall, our main goals were to: 1. Gain a mechanistic understanding of the effects of calcite impurities on U(VI) sorption and diffusion in montmorillonite/bentonite at variable chemical conditions, and before and after mineral exposure to heat; and to 2. Work towards the development of new surface complexation and diffusion models to predict U(VI) sorption and mobility in montmorillonite/bentonite while accounting for the impacts of calcite impurities and heat. In order to achieve these goals, we pursued the following research objectives and tasks: 1. Characterize the effects of calcite impurities on U(VI) sorption onto Na-montmorillonite, and evaluate if and how these effects change due to solids’ exposure to heat (Task 1). 2. Characterize U(VI) surface speciation in solid samples from Task 1, with a particular focus on U(VI) precipitates and co-precipitates (Task 2). 3. Test the relevance of observed changes in U(VI) sorption characteristics for U(VI) diffusive transport behavior in bentonite barriers (Task 3). 4. Evaluate the exclusion of anionic U(VI) solution complexes from clay nanopores (Task 4). 5. Building on Tasks 1-4, start the development of new surface complexation and reactive transport models that can simulate and predict U(VI) sorption and diffusion in complex bentonite systems (Task 5). A detailed description of all tasks, and their specific results and conclusions is provided in this report.

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DOE OSTI2022

Hanford Low-Activity Waste Glass Minor Component Concentration Boundary Expansion

This study investigated the effects of increasing concentrations of minor components on simulated Hanford low-activity waste (LAW) glass properties to allow for LAW glass composition envelope expansion. A test matrix of ten glasses was generated by spiking three previously studied simulated Hanford LAW glass compositions with increasing concentrations of minor component mixes up to a total concentration of 2 wt%. In this study, the ten glasses were fabricated, and key properties were measured including crystal formation after canister centerline cooling (CCC), crystallinity as a function of temperature, density, viscosity, electrical conductivity, glass durability using the Vapor Hydration Test (VHT), product consistency using the Product Consistency Test (PCT), and sulfur solubility. The data from these tests are presented in this report.

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DOE OSTI2023

Product Consistency Test Results for the LAW HPVR Glasses

This report summarizes the chemical analysis of Product Consistency Test (PCT) leachates received from Pacific Northwest National Laboratory (PNNL). The leachates are from a series of simulated nuclear waste glasses designated Low-Activity Waste High PCT and VHT Response (LAW HPVR) glasses that were designed and fabricated at PNNL. The glasses included both quenched and canister centerline cooled glasses. The reported data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The elemental release for the study glasses is reported as normalized concentration (NC i ). NC i of several elements was computed for both the target and measured glass compositions, which were similar, resulting in no significant differences when computing the NC i values. The majority of the glasses exhibited NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment Plant (WTP) low-activity waste constraint of 4 g/L. Several blank solutions included with the study glasses had K, Na, and Si concentrations above the analytical detection limit. Additionally, several of the reference glasses that were included with the study glasses had Na concentrations that were higher than expected. The unexpected analyte concentrations in the reference glass and blank solutions were insignificant with respect to the reported NC i values, and are noted for completeness and their experimental relevance.

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DOE OSTI2020

Analysis of Respirator Cartridge Performance Testing on the 702-AZ Primary Exhauster for the Hanford AY/AZ Tank Farms during a Waste-Disturbing Event

Washington River Protection Solutions (WRPS) conducted tests using two types of chemical cartridges for use in air-purifying respirators (APR) to determine the period of time that the cartridges would provide adequate performance1 for APRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the 702-AZ Primary Exhauster for the Hanford AY-AZ tank farms. Unlike prior cartridge testing on the 702-AZ Primary Exhauster, the recent tests were performed during a waste-disturbing event. The Occupational Safety and Health Administration (OSHA) identifies cartridge testing as a valid approach for establishing a cartridge change schedules. Testing is commonly applied in situations where mixtures of COPCs exist, and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to assure measurement and/or control of the key variables OSHA identified as important to estimate the cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Testing was conducted from February 10–11, 2017, on a slipstream from the 702-AZ exhauster fed to a respirator cartridge test stand developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose respirator cartridges, SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina), were assessed on separate days. Sample media (sorbent tubes) and canisters (e.g., Summa) sampling were used to collect samples of the vapor stream entering and exiting the respirator cartridge and were subsequently analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life.

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DOE OSTI2023

Feasibility Study for Disposal Control Rod Assemblies Using UNF-ST&DARDS As-Loaded Zion Dual Purpose Cask Models

This report documents an initial evaluation to support future use of disposal control rod assemblies (DCRAs) for post-closure criticality control in dual purpose canisters (DPCs). The work described herein is an extension of previous efforts performed by Walker using inputs generated by the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNFST&DARDS) for the Zion site DPCs with as-loaded isotopic compositions. The results of this analysis demonstrate that there are multiple pathways to using DCRA for post-closure criticality control. Various configurations of DCRA material, diameter, number of rods per DCRA, and number and location of DCRAs within a DPC were shown to be effective in varying degrees for the set of DPCs analyzed by Walker. Because of the variations in DPC as-loaded isotopic compositions considered in an array of DCRA parametric sweeps, it can be concluded that a DPC-specific methodology is feasible (i.e., a one-size-fits-all approach may not be needed). Instead, the utility program created for this work can be expanded to develop capabilities to provide DPC-specific DCRA arrangements to limit cost and weight and to allow for operational considerations.

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DOE OSTI2022

Thermal Modelling of Advanced Test Reactor Fuel in a Generalized Dry Storage System

Star-CCM+, a computational fluid dynamics (CFD) software was used to conduct modeling and simulation of the thermal performance of a dry storage configuration consisting of Department of Energy Standardized Canisters (DOESCs) loaded with aluminum-clad spent nuclear fuel (ASNF.) The configuration includes nine DOESCs loaded with Advanced Test Reactor ASNF contained within a stainless steel overcanister centered in a ventilated, concrete overpack. The simulations were used to estimate the maximum temperatures reached by backfill gases inside the overcanister and DOESCs.

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DOE OSTI2022

Preparation & shipping of ten neutron irradiated Eurofer97 steel variants to HFIR for SANS experiments

At ORNL, ten variants of Eurofer97 steel were irradiated in the high flux isotope reactor (HFIR) to ~2.94 – 3.24 dpa at 300 ± 30 °C, as part of the EUROfusion Lot-IV collaboration. The irradiations were performed in non-instrumented rabbit capsules in the flux trap region, which included SS-J3 type tensile samples and M4-CVN multi-notch bend bars. Their full set of microstructure and mechanical properties were previously reported in multiple previous publications. In this project, half-broken irradiated and nonirradiated SS-J3 tensile samples from ten alloys, code-named H, I, J, K, L, M, N, O, P and E, were prepared and shipped for facilitating small angle neutron scattering (SANS) experiments at the HFIR general purpose (GP)-SANS beam line. This report summarizes the completed tasks which included canister moves at the Irradiated Materials Examination and Testing (IMET) hot-cell facility to retrieve the samples from long-term storage, loading of the ten irradiated samples at IMET inside lead (Pb) piglets that were specifically provided by ORNL for SANS experiments and radiological shipment from IMET facility to the HFIR hot cells for performing SANS experiments. In addition to the irradiated samples, ten nonirradiated half-broken pieces from the Eurofer97 variants were also provided to HFIR for SANS.

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DOE OSTI2023

UNF-ST&DARDS Enhancements for RCCA Data in As-loaded Dual Purpose Cask Models

This report summarizes the work performed to enable detailed modeling of rod cluster control assembly (RCCAs) in dual purpose canisters (DPCs) in as-loaded configurations using the Used Nuclear Fuel – Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). The goal of this project was to evaluate the reactivity impact that RCCAs have on k eff of DPCs with pressurized water reactor (PWR) fuel to potentially use the additional margin in future post-closure criticality safety analysis. This preliminary evaluation determined the number of DPCs that currently require compensatory actions prior to emplacement in a repository because of their high reactivity under post-closure criticality scenarios, which could be made acceptable by including the as-loaded RCCAs as specified in the Unified Database (UDB). This report briefly describes the modeling methods currently used within UNF-ST&DARDS and the modifications made to automate inclusion of the as-loaded RCCAs in the DPCs for post-closure criticality calculations for the loss of neutron absorber (NA) scenarios, or NA models. For the Zion site, the loss of basket scenarios, or degraded basket (DB) models, are also included. Discussion is also provided regarding the UDB data compared to available site-specific loading map data for the sites, which are specifically evaluated herein. This report compares the modifications to a similar evaluation for the Zion by site conducted by Walker as an initial validation. After the partial validation, all existing PWR sites with applicable NA models within the UDB were evaluated.

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DOE OSTI2023

Non-Fouling, Low Cost Electrolytic Coagulation & Disinfection for Treating Flowback and Produced Water for Reuse

Executive Summary This Final Report is composed of two major sections. The first section presents experimental results obtained in the laboratory during Budget Period 1. The second section presents results from the field test conducted during Budget Period 2. Laboratory Results This research investigated a novel electrochemical process for producing a ferric iron coagulant for use in treating flowback and produced water from hydraulic fracturing and oil production operations. The treatment system improves the effectiveness and lowers the cost of coagulation processes using Fe3+ as the coagulant. The electrolytic coagulant generation (ECG) system uses an electrochemical cell to produce acid and base from oilfield brine solutions. The acid is used to dissolve scrap iron to provide a Fe3+ coagulating agent. The base is used to neutralize the treated water. Compared to conventional electrocoagulation (EC), the main advantage of the ECG system is an order of magnitude lower cost for the source of iron. The second advantage over conventional EC is that it can deliver Fe3+ doses greater than 1 mM, since it is not limited by the amount of dissolved oxygen in the water required to oxidize ferrous to ferric iron. The capital costs for conventional EC and the ECG system are similar, but the operational costs for the ECG system are an order of magnitude lower than conventional EC. The combined costs for iron and electrical energy for treating 1 m3 of FPW with 1 mM Fe3+ is estimated to be $0.87 for conventional EC, and $0.087 for the ECG system. The estimated all-in cost for treating FPW with a 2 mM Fe3+ dose is $0.73/m3 ($0.12/bbl). Field Test Results The field test was conducted at the Paul Foster Central Tank Battery (CTB) in Lea County, New Mexico from November 10, 2022 through December 15, 2022. The feed water to the system was produced water from the Tatanka 1H formation. After approximately two weeks of testing, the initial batch of produced water had been treated and no untreated produced water was available. Thus, after this time, the feed water to the system consisted of previously treated water (i.e., recycled water). The recycled water had nearly all colloidal particles removed, and had a much lower alkalinity due to precipitation of carbonate minerals during the first pass through the system. Although the recycled water was not an ideal test solution due to its low particulate concentrations, its lower alkalinity did allow us to identify the main problem with the treatment system. The main problem with the treatment system was caused by the high alkalinity of the initial feed water (5.4 meq/L) that consumed a significant fraction of the electrochemically generated acid. This resulted in pH values exiting the iron contact tank that were too high to dissolve enough iron to effectively treat the produced water. Tests performed with recycled water with lower alkalinity did not have this issue, and dissolved iron concentrations greater than 20 mM could be achieved. One consistent observation was that effluent water from the iron contact tank was always free of particulates, even when fed with circumneutral solutions. This suggests that there is no need to dissolve high concentrations of iron if all the water to be treated is passed through the scrap iron canister. In this case, dissolved O2 and hypochlorous acid can promote sufficient iron corrosion to provide an effective coagulating agent – even in neutral pH water. This solves the problem resulting from highly alkaline produced water. Modifications to the design of the treatment system were made based on the field test results. These modifications will add minimal additional cost, and were tested in bench-scale laboratory experiments. In short-term testing, the modified treatment process was able to remove colloidal FeS particulates to levels below detection. Long-term, steady state testing will be required to determine whether the modified process is suitable for commercial treatment systems.

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DOE OSTI2023

A Vacuum Drying Study of Simulated Failed Nuclear Fuel (FY23)

This report documents work performed under the Office of Spent Fuel and Waste Disposition’s (SFWD’s) Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 4 milestone M4SF-23OR010203013. Over the 70 years of light-water reactor (LWR) operation, fuel rod failures have occurred during reactor operation. Based on the Unified Database that SFWD maintains, there are ~5,400 LWR fuel rods that developed leaks during reactor operation of the >19 million rods operated. The Unified Database indicates that ~4,200 of these rods remain in the spent fuel pool, and ~1,200 are in dry storage canisters. It is expected that all failed rods are “waterlogged” in the spent fuel pool, meaning the inside of the rods contain some volume of water. There are two related objectives of this work. The first is to develop a simple and inexpensive experiment to better understand removal of water from inside failed spent nuclear fuel (SNF) rods, and the second is to develop a model to predict drying of simulated failed fuel in the simple experiment. Once a model captures the phenomenon in the simple experiment, it can be extended to predict drying under more prototypical conditions.

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DOE OSTI2020

Guide to CFD-Chemical Model for Spent Fuel Storage

This outlines the steps that were used to create the CFD model in Star-CCM+ to simulate the 50 year thermal history of ATR fuel stored in DOE sealed canisters. Then this goes through the steps to set up the processing the model results for the one-way coupled Cantera chemical model. The next part describes the process for setting up the 50-year Cantera model in Python. The goal of this is to document the procedure for quality assurance, as well as in the event another person will need to run the process to generate additional results.

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DOE OSTI2023

Evaluation of Flaw Detection with SLDV Methods Before/After Mechanical Patch

This report provides a summary of the testing and the experiments to evaluate the utility of laser ultrasonic testing (UT) in detecting and characterizing defects on the surface of steel plates before and after the application of an engineered composite patch for crack repair. Included are the detection and characterization of defects from a section of welded plate material taken from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister which had been pre-flawed and exposed to deliquescent sea salt conditions for 2 years. The information includes a review of previously reported metallurgical characterization, and new work on detailed surface characterization of the corrosion attack using laser UT.

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DOE OSTI2023

ASNF Dry Storage Pilot with HFIR Fuel: Concept Plan

The U.S. Department of Energy (DOE) currently manages a large inventory of aluminum-clad spent nuclear fuel (ASNF) from U.S. High Performance Research Reactors and also some fuel from foreign research reactors. A transition to new dry storage configurations is desired to facilitate decisions related to potential storage of future ASNF and transitioning current ASNF out of aging storage facilities. DOE has developed designs for a set of DOE Standard Canisters (DSCs) to provide sealed, road-ready dry storage suitable for interim storage, transportation, and potentially disposition in a future repository. Extensive experiments and modeling activities have been carried out to develop the technical basis for such storage, with results to date indicating this storage approach should be safe and viable following suitable drying processes.

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DOE OSTI2023

Analysis of the Sludge Batch 9 (Macrobatch 11) DWPF Pour Stream Glass Sample

A pour stream (PS) sample taken near the end of processing of Sludge Batch 9 (SB9, Macrobatch 11) with the Nitric-Formic Acid Flowsheet was analyzed by the Savannah River National Laboratory to comply with the Defense Waste Processing Facility (DWPF) Glass Product Control Program. The SB9 PS sample was taken on 5/28/2022 during Melter Feed Batch 815 associated with glass canister S04741, and was transferred to SRNL in Primary Container PC0136.

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