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A novel approach to mitigating the potential release of radioisotopes under fire conditions - enhancing fire resiliency of radiological contamination fixatives during deactivation & decommissioning activities

Savannah River National Laboratory (SRNL), in close collaboration with the Florida International University Applied Research Center (FIU ARC), successfully executed a technology development activity on behalf of the Department of Energy, Office of Environmental Management (DOE EM). The purpose of the activity was to improve the capability of fixative technologies in immobilizing residual contamination when exposed to thermal stressors as postulated in accident scenarios in Basis for Interim Operations (BIO) documents across the DOE EM complex. The effort resulted in the test and evaluation of a down-selected, commercial-off-the-shelf (COTS) intumescent technology in a radioactive environment at the Savannah River Site (SRS) Building 235-F Plutonium Fuel Form (PuFF) Facility and highlighted the potential use of this technology as a stand-alone, fire retardant fixative. Furthermore, this activity highlighted significant shortfalls in common fixatives currently used to support decommissioning activities and brought to the forefront the need for a methodical, uniformed approach to certifying fixative technologies for operational use in Deactivation and Decommissioning (D&D) activities. An essential component of this research was to evaluate the state of industry fixatives currently in use and set the foundation for comparison to alternative technological solutions. A baseline of five (5) commonly used fixatives and decontamination gels (hereafter collectively referred to as fixatives) was conducted, and notable shortfalls and deficiencies were revealed, particularly when exposed to thermal, water, and other environmental stressors. At temperatures as low as 300 °F - 400 °F (148.89 °C – 204.44 °C), all commonly used fixatives melted from the substrates within 3-5 minutes of exposure, resulting in contaminant transport. Significant degradation in terms of mass loss, desiccation, and off-gassing occurred, and the chemical breakdown of the polymer was so complete researchers assessed there would likely have been a release of residual contamination. Additional vulnerabilities became evident when exposed to water immersion and high humidity; existing fixatives took up water, swelling and frequently delaminating from the substrate, once again increasing the likelihood of a contaminant release.

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BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project

The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.

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Transportation and Systems Analysis Collaborations in Support of a Federal Consolidated Interim Storage Facility

The U.S. Department of Energy’s Integrated Waste Management (IWM) program under the Office of Nuclear Energy is planning for the future transportation, storage, and eventual disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from nuclear power plant sites across the United States. To better enable informed decision-making regarding the back end of the nuclear fuel cycle, the IWM program has been sponsoring the development and application of system analysis tools capable of analyzing various options for managing SNF and HLW. With these tools, integrated waste management system (IWMS) architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analyses and assessments typically use these modeling and simulation tools to investigate implications of changes in various assumptions and parameters such as acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, start and stop dates of facilities, etc. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation toolkit that is used for a system-level simulation and analysis focused on SNF management in the United States. An analyst using NGSAM has the ability to define several factors like the number of storage facilities, capacity at each facility, transportation schedules, shipment rates, and other conditions. NGSAM’s primary purpose is to provide a system analyst with capabilities to model the IWMS and gain insights into SNF and HLW management alternatives including the impact of system choices, associated cost estimates, and development of integrated yet flexible approaches. IWM is also developing the Stakeholder Tool for Assessing Radioactive Transportation (START). START is a web-based geospatial tool developed to provide visualization and initial evaluation of transportation options associated with future SNF and HLW shipment planning and operations. This includes characterizing safety, economic, and environmental conditions on and in proximity to shipment origins as well as along prospective transportation routes. Information from the START tool can be presented/shared as maps, graphics, geospatial files, and tabular form to enable easy data representation and export functionality. The system analysis, NGSAM, and START teams had been working closely for several years before formalizing this collaboration. START provides the SNF routing information for use in NGSAM. System analysts use the NGSAM tool to generate results (schedule, costs, infrastructure acquisitions, shipment rates, etc.). Subject-matter experts and system analysts work together to inform how NGSAM should model the waste management system. Specifically, this paper discusses the collaborations between the system analysis, NGSAM, and START teams and their accomplishments over the past year. Some examples of these efforts include calibrating and adding data to the START output files to meet NGSAM needs, gaining a better understanding of START data used in NGSAM, as well as how updates in START data could result in an improved NGSAM analysis. Detailed examples of various tasks that have been performed by the team will be discussed in the full paper. Continued efforts in this direction are expected in the coming years.

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Transportation and System Analysis Collaborations in Support of a Federal Consolidated Interim Storage Facility

The U.S. Department of Energy’s Integrated Waste Management (IWM) program under the Office of Nuclear Energy is planning for the future transportation, storage, and eventual disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from nuclear power plant and waste custodian sites across the United States. To better enable informed decision-making regarding the back end of the nuclear fuel cycle, the IWM program has been sponsoring the development and application of system analysis tools capable of analyzing various options for managing SNF and HLW. With these tools, integrated waste management system (IWMS) architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analyses and assessments typically use these modeling and simulation tools to investigate implications of changes in various assumptions and parameters such as acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, start and stop dates of facilities, etc. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation toolkit that is used for a system-level simulation and analysis focused on SNF management in the United States. NGSAM’s primary purpose is to provide a system analyst with capabilities to model the IWMS and gain insights into SNF and HLW management alternatives including the impact of system choices, associated cost estimates, and development of integrated yet flexible approaches. An analyst using NGSAM can define several factors like the number of storage facilities, capacity at each facility, transportation schedules, shipment rates, and other conditions. IWM is also developing the Stakeholder Tool for Assessing Radioactive Transportation (START). START is a web-based geospatial tool developed to provide the visualization and initial evaluation of transportation options associated with future SNF and HLW shipment planning and operations. This includes characterizing safety, economic, and environmental conditions at and in proximity to shipment origins as well as along prospective transportation routes. Information from the START tool can be presented/shared as maps, graphics, geospatial files, and tabular form to enable easy data representation and export functionality. The system analysis, NGSAM, and START teams have been working closely for several years before formalizing this collaboration. START provides the SNF transportation routing information for use in NGSAM. System analysts use the NGSAM tool to generate results (schedule, costs, infrastructure acquisitions, shipment rates, etc.). Subject-matter experts and system analysts work together to inform how NGSAM should model the waste management system. Specifically, this paper discusses the collaborations between the system analysis, NGSAM, and START teams and their accomplishments over the past year. Some examples of these efforts include calibrating and adding data to the START output files to meet NGSAM needs, gaining a better understanding of START data used in NGSAM as well as how updates in START data could result in an improved NGSAM analysis. Detailed examples of various tasks that have been performed by the team will be discussed in the full paper. Continued efforts in this direction are expected in the coming years.

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Preliminary Analysis of Advanced Reactor Spent Nuclear Fuel Storage, Transportation, and Disposal

Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.

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Preliminary Analysis of Advanced Reactors Storage, Transportation, and Disposal

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP[1]) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes the efforts by the Spent Fuel & Waste Science and Technology (SFWST) in evaluating characteristics and packaging options for advanced reactor spent nuclear fuel forms. The fuel forms were categorized into three types: (1) tristructural isotropic (TRISO), (2) metallic, and (3) fuel salt. This work emphasized TRISO and metallic SNF and waste streams because of the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations. Preliminary information for the spent-fuel salts discharged from molten-salt reactors (MSRs) is also examined to provide a baseline for future efforts. All calculations and assumptions used publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. The paper also includes a literature review and analysis on the storage, transportation, and disposal evaluations and experiences from reactors with similar fuel forms. Advanced-reactor vendors cite past experiences with Fort St. Vrain for TRISO and the Experimental Breeder Reactor II (EBR-II) for metallics that have major influences on fuel design. Finally, the paper includes preliminary concepts of operation for advanced-reactor SNF. This encompasses storage, transportation, potential treatment, and disposal activities from both a per-canister and systems-integration perspective.

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Nuclear Material Accounting and Control (NMAC) for a Commercial Fuel Debundling Facility in the United States

The concept of fuel debundling proposes that all non-fuel components of a fuel assembly be removed prior to beginning the recycling process. It was introduced to reduce the quantity of and cost of disposing high level radioactive waste during recycling in the United States. However, nuclear material accountancy and control programs could also benefit from fuel debundling due to a reduction in self-shielding material for individual fuel rods compared to a fuel assembly. Current NRC regulations and guidance point to an independent fuel debundling facility with a material-possession-only license as the best option for implementing this concept into a used fuel recycling scheme in the United States. This allows the facility to comply only with simple NMAC requirements under Subpart B of 10 CFR Part 74 without the need to submit an NMAC plan to the NRC, exempting it from more tedious and complex NMAC requirements. However, a recycling facility may desire a Pu mass estimate from measurements of each individual fuel rod to improve accountancy into the dissolver. Many technologies and methodologies were considered for performing NMAC in an independent fuel debundling facility in the United States. Few were able to meet anticipated throughput demands of up to 3,300 rods/day and achieve measurement uncertainties suitable for input accountancy for a recycling facility. The best option was a combination of passive gamma-ray spectroscopy and neutron coincidence counting to meet all requirements for the proposed Fuel Debundling Detector (FUDD). Follow-on work will include a full uncertainty analysis to inform what assumptions can be made to meet input accountancy objectives before beginning experimental activities.

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Radionuclide Waste Disposal: Development of Multi-scale Experimental and Modeling Capabilities (Final Report)

The DOE EPSCoR Implementation project “Radioactive waste management: Development of multi-scale experimental and modeling capabilities” helped to develop a team of scientists and engineers from Clemson University, South Carolina State University, and the University of South Carolina to address the disposition of nuclear wastes and study the transport of radioisotopes from a waste repository in the near and far field. The project involved 20 faculty from the three institutions as well as 13 postdoctoral fellows, 32 graduate students, and 32 undergraduate students and was active from 2014-2019. Additionally, we forged new collaborations with nine researchers from DOE laboratories SRNL, LLNL, and ANL as well as the University of Manchester and the China Academy of Engineering Physics. The overarching goal of the project was to understand the conditions under which important classes of co-reactants, ranging from counter ions in crystal lattices to dissolved oxygen in pores, control the chemistry and transport characteristics of radionuclides in engineered waste forms and natural soils. Our approach was to characterize the time and length scales over which non-equilibrium states are maintained by rate-limiting, or rate-enhancing, reactions between radionuclides and co-reactants due to interactions between physical mass-transfer processes (i.e., advection, diffusion) and (biogeo) chemical reactions. We have focused our project on three specific classes of reactions relevant to radionuclide transport at DOE legacy sites: ion exchange/substitution, ligand complexation, and redox-mediated reactions. Understanding radionuclide migration requires detailed knowledge of how changes to a system – whether engineered or natural – drive the behavior of co-reactants, which in turn provide the geochemical context controlling radionuclide transport. Student engagement and training were a primary focus of the project in order to create a pipeline of researchers who could work in the area of nuclear waste disposition to support the state and the nation. Over the duration of the project we worked with 32 undergraduate, graduated 18 M.S. students and 14 Ph.D. students, and advised 13 postdoctoral fellows. The Ph.D students and postdocs have primarily taken positions at DOE laboratories, academia, and industry. Through our collaborative team, numerous follow on projects have been started with over $5M in sponsored research. Additionally, thus far the team has published 53 peer reviewed papers and given over 75 technical presentations.

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System Analysis Support of Consent-Based Siting Efforts - 23400

To better enable informed decision-making regarding the back end of the nuclear fuel cycle, the Integrated Waste Management (IWM) program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the development and application of system analysis tools capable of analyzing various system options for the management of spent nuclear fuel (SNF) and high-level radioactive waste. With these tools, integrated waste management system (IWMS) architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analyses and assessments typically use these modeling and simulation tools to investigate things like the implications of changes in various assumptions and parameters such as acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, and different assumed system operation start dates. Some example system analysis results have been published previously. With the passage of the consolidated appropriations acts for fiscal years 2021 and 2022, Congress provided funding and direction for the DOE to move forward with interim storage to support near-term action in managing the nation’s SNF as an important component of an IWMS. Consent-based siting is a phased, adaptive, and collaborative approach to siting facilities that focuses on the needs and concerns of people and communities. Communities participate in the siting process by working carefully through a series of phases and steps with the DOE (as the implementing organization). Each step and phase helps a community determine whether and how hosting a facility to manage SNF is aligned to the community’s goals. By its nature, a consent-based siting process must be transparent, flexible, adaptive, and responsive to community concerns. Thus, the phases and steps are intended to serve as a guide, not a prescriptive set of instructions. System analysis research analysts interface with DOE’s consent-based siting for a federal interim storage program and can perform preliminary system analyses as requested to support various aspects of the consent-based siting team’s work. This paper provides an overview of IWM program’s system analysis activities that support the ongoing consent-based siting process. The questions that decision-makers, planners, communities, and stakeholders might have during the consent-based siting process are similar to questions that have been explored by system analysts previously. Specifically, this paper discusses work that will inform consent-based siting efforts in the areas of community impacts, cost estimates, various SNF acceptance rates by potential facilities, potential consolidated interim storage facility (CISF) capacities and configurations, and the implications of a CISF on the overall waste management system. It is expected that system analysis will continue to be performed to support the consent-based siting process in the future.

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Data Summary Report for Soil and Slab Sampling at Former Building 175

Lawrence Livermore National Security (LLNS) and the U.S. Department of Energy (DOE) are in the process of returning the area where Building 175 existed to beneficial use for future site development. The completed assessment will help LLNS and DOE to outline project scope and costs associated with the removal, characterization, and disposal of materials generated when the concrete slab for the former building and any associated remaining subsurface structures are demolished. Slab removal under the Transition & Disposition process requires full screening for potential site contamination to determine appropriate future land use. Borehole locations and depths were designed to correctly assess DOE’s future liability for characterizing residual vadose zone contamination. If a residual source area is identified at depth, DOE will need to evaluate the additional cost of future subsurface cleanup (with limited access) due to planned development. This is necessary to leave the location in a “ready-to-build” status. As a result, Phase II sampling requirements may in some cases extend beyond the construction zone required for slab removal to ensure the necessary due diligence. To avoid the potential spread of contamination and/or the creation of an environmental release by impacting the integrity of known contamination areas (e.g., pit within room 102 and the east to west boundary seam), these locations were not sampled for this project. During demolition of the slab, these locations and their underlying soils should be further evaluated. The assessment was conducted in accord with the May 2022 Lawrence Livermore National Laboratory Former Building 175 Assessment Soil Sampling and Analysis Plan / Quality Assurance Plan. Thirty-six, direct-push borings were advanced, along with collecting concrete (where present) and soil samples for laboratory analyses. Thirty-two of the borings were advanced to a depth of 25 feet below ground (bgs), and four borings were advanced to a depth of 55 feet bgs. The concrete slab for former Building 175 ranged from approximately 8 inches to over two feet in thickness. Soils encountered during the investigation consisted primarily of clayey silt, with interbeds of sandy gravel and silty sand to the total depth explored of 55 feet bgs. Field photoionization detector readings – checking for volatile organic compound (VOC) vapors in soils, ranged from zero (0) to a peak of 33 parts per million at 25 feet bgs in boring PC-B175- 028. The cause for the peak reading is unknown, however, no visually discolored or odorous soils were encountered during the assessment and all VOC results were below Lawrence Livermore National Laboratory’s (LLNL’s) Soil Screening and Management Plan (SSMP) soil screening levels (SSL). Groundwater was not encountered during the assessment and is expected to occur at roughly 65 feet bgs in the project area. No metals were detected at concentrations of concern in the concrete core samples collected and analyzed. Low gross alpha and beta activity concentrations were detected in the concrete core samples collected and analyzed. The detected activity concentrations would appear to be from naturally occurring radioactive isotopes, i.e., potassium 40, present in the raw materials used to make concrete and not artificially added. Tritium was not detected above the testing laboratory's Method Detection Limit (MDL) in any of the concrete core samples collected and analyzed. The radioactive isotopes - actinium 228, bismuth 214, lead 212, lead 214, potassium 40, radium 226, radium 228, and thallium 208, were detected at low activity concentrations in the concrete core samples analyzed. Based on the detected activity concentrations, the isotopes would appear to be naturally occurring in the raw materials used to make concrete and not artificially added. Acetone – a common laboratory contaminant, was detected in nine of the soil samples collected and analyzed. Concentrations of two other volatile organic compounds (benzene and tetrachloroethene) were detected in five of the soil samples analyzed. All concentrations were below SSLs. Total petroleum hydrocarbons as diesel range organics were detected at 43.9 milligrams per kilogram (mg/kg) in the 20-foot bgs duplicate sample from boring PC-B175-007. Total petroleum hydrocarbons as motor oil were detected at 7.20 mg/kg in the 15-foot bgs routine sample from boring PC-B175-001, and at 20.7 mg/kg in the 20-foot bgs duplicate sample from boring PC-B175-007. The SSL for diesel-range TPHs is 260 mg/kg. LLNL does not have an SSL for motor oil range TPHs. No indications of a release, e.g., visibly stained, or odorous soil, were present at the boring locations. No samples collected contained polychlorinated biphenyls (PCBs) in concentrations above the laboratory reporting limit. Arsenic was detected above its SSL of 8.51 mg/kg in the 25-foot bgs routine sample collected from boring PC-B175-023. Nickel was detected above its SSL of 86.0 mg/kg in the 10-foot bgs routine sample from boing PC-B175-005, the 20-foot bgs routine sample from boring PC-B175-025, and the 25-foot bgs routine sample from boring PC-B175-034. The detected concentrations, however, were well below ten times (10x) their respective STLCs. Gross alpha and/or gross beta were detected above their respective SSL activity concentrations in three routine soil samples. Retesting (two per sample) of the samples showed that the initial reported activity concentrations were anomalous. Tritium was not detected in any of the routine or duplicate soil samples collected and analyzed during the investigation. The radioactive isotopes - actinium 228, bismuth 212 and 214, lead 212 and 214, potassium 40, radium 224, 226 and 228, thallium 208, thorium 234 and uranium, were detected at low activity concentrations in the soil samples analyzed for radioactive constituents. The detected isotopes and their associated activity concentrations are typical of those naturally occurring in the marine-type sedimentary deposits underlying the Livermore Valley. No radiological controls are necessary for the soil evaluated.

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CoDCon Project (Final Report)

The co-decontamination (CoDCon) project was established in FY 2016 with the objectives of (a) evaluating the uncertainty in the uranium (U)/plutonium (Pu) ratio in a mixed U/Pu product from a tributyl phosphate (TBP)–based solvent extraction flowsheet, and (b) developing and demonstrating on-line optical spectroscopy for real-time monitoring of key components (e.g., Pu, U, and HNO 3 concentrations) in the process solutions. We were interested in assessing the accuracy and precision to which a specific uranium-to-plutonium (U/Pu) ratio can be achieved, which for the purposes of this project was set at a U/Pu mass ratio of 7/3. The uncertainty associated with achieving this specific target U/Pu ratio was investigated during five flowsheet tests using laboratory-scale solvent extraction equipment. In addition, optical spectroscopic techniques were incorporated into the CoDCon solvent extraction testing system, allowing real time monitoring of all input and output process streams. Two CoDCon flowsheet tests were performed in FY 2018 using a simple dissolved fuel simulant containing only U (~1 M) and Pu (~15 mM) in nitric acid (HNO 3 ; ~3 M). In FY 2019, two additional flowsheet tests were performed. For the first of these (CoDCon Run 3), the dissolved fuel simulant was similar to that used in the first two tests, with the inclusion of 1 mM neptunium (Np). The second test conducted in FY 2019 (CoDCon Run 4) used a more representative dissolved fuel simulant, including addition of non-radioactive fission product elements. A fifth CoDCon flowsheet test (CoDCon Run 5) was conducted in FY 2020, with the following additional objectives: (1) routing of the technetium (Tc) in the simulated dissolved fuel solution to the solvent extraction raffinate, and (2) routing of the Np in the simulated dissolved fuel solution to the U/Pu product. All tests used a bank of sixteen 2 cm centrifugal contactors. The tests involved first loading the solvent (30 vol% TBP dissolved in n-dodecane) with U and Pu (and Np, for Run 5), then the Pu (and Np) was stripped from the loaded solvent with a U(IV) solution (~50 mM) and the flowsheet conditions were adjusted such that some U partitioned into the Pu-containing product stream. The amount of U accompanying the Pu was monitored in real time using optical spectroscopic techniques coupled with chemometric modeling. Based on the real-time spectroscopic measurement of the U/Pu ratio, adjustments were made to the flowrate of the fresh TBP solvent phase used to scrub U from the aqueous Pu-containing product. This proved to be a very effective way to control the U/Pu mass ratio in the product. This report presents the results of the CoDCon Run 5 test. The flowsheet tested in Run 5 was substantially different than that run in the prior tests, especially the solvent loading section of the flowsheet. Two key changes were made. First, based on the objective to extract all the Np and route it with the U/Pu product, pentavalent vanadium [V(V)] was added to the feed and scrub solutions. The purpose of the V(V) was to convert all the Np to the +6 oxidation state, which is extractable by TBP. Second, a high acid (8 M HNO 3 ) scrub was added to the flowsheet to scrub the Tc from the solvent. This was followed by a low acid scrub (0.05 M HNO 3 ) to reduce the residual HNO 3 concentration in the solvent prior to the Pu stripping step. The output from the low acid scrub was collected separately, rather than routing towards the raffinate. The modifications to the solvent loading part of the flowsheet were only partially successful. The treatment with V(V) was effective at converting the Np to Np(VI). Only 1.3% of the Np remained in the raffinate solution. However, ~40% of the Np stripped out of the solvent in the low acid scrub step; nearly 20% of the Pu also was stripped from the solvent during the low acid scrub. For further development, either modifications to the flowsheet, or concentration and recycle of the low acid stream into the

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Presence of aromatic-rich organic matter and its characterization in grout materials: Implications for radionuclide immobilization

Grout materials are commonly used to immobilize low-level radioactive waste. Organic moieties can be unintentionally present in common ingredients used to make these grout waste forms, which may result in the formation of organo-radionuclide species. These species can positively or negatively affect the immobilization efficiency. However, the presence of organic carbon compounds is rarely considered in models or characterized chemically. Here, we quantify the organic pool of grout formulations with and without slag, as well as the individual dry ingredients used to make the grout samples (ordinary Portland cement (OPC), slag and fly ash), including total organic carbon (TOC) and black carbon, followed by aromaticity evaluation and molecular characterization via Electro Spray Ionization Fourier-Transform Ion Cyclotron Resonance Mass Spectrometry (ESI-FTICRMS). All dry grout ingredients contained significant amounts of organic carbon, ranging from 550 mg/kg to 6250 mg/kg for the TOC pool, with an averaged abundance of 2933 ± 2537 mg/kg, of which 60 ± 29% was composed of black carbon. The significant abundance of a black carbon pool implies the presence of the aromatic-like compounds, which was further identified by both phosphate buffer-assisted aromaticity evaluation (e.g., >1000 mg-C/kg as aromatic-like carbon in the OPC) and dichloromethane (DCM) extraction with ESIFTICRMS analysis. Besides aromatic-like compounds, other organic moieties were also detected in the OPC, such as carboxyl-containing aliphatic molecules. While the organic compound only consists of minor fractions of the grout materials investigated, our observations of the presence of various radionuclide-binding organic moieties suggests the potential formation of organo-radionuclides, such as radioiodine, which might be present at lower molar concentrations than TOC. Here, evaluating the role of organic carbon complexation in controlling the disposed radionuclides, especially for those radionuclides with strong association with organic carbon, has important implications for the long-term immobilization of radioactive waste in grout systems.

36 MATERIALS SCIENCE↗

Crystallization of a Neptunyl Oxalate Hydrate from Solutions Containing Np V and the Uranyl Peroxide Nanocluster U 60 Ox 30

Uranyl peroxide nanoclusters are an evolving family of materials with potential applications throughout the nuclear fuel cycle. While several studies have investigated their interactions with alkali and alkaline earth metals, no studies have probed their interactions with the actinide elements. This work describes a system containing U 60 Ox 30 , [((UO 2 )(O 2 )) 60 (C 2 O 4 ) 30 ] 60- , and neptunium(V) as a function of neptunium concentration. Ultra-small and small angle X-ray scattering were used to observe these interactions in the aqueous phase, and X-ray diffraction was used to observe solid products. The results show that neptunium induces aggregation of U 60 Ox 30 when the neptunium concentration is ≤ 10 mM, whereas (NpO 2 ) 2 C 2 O 4 ·6H 2 O(cr) and studtite ultimately form at 15–25 mM neptunium. Finally, the latter result suggests that neptunium coordinates with the bridging oxalate ligands in U 60 Ox 30 , leaving metastable uranyl peroxide species in solution, and is an important finding given the potential application of uranyl peroxide nanoclusters in the recycling of used nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hydrogen uptake in graphite matrix at high temperature

Tritium management is a critical challenge for the next generation of nuclear reactors, such as Fluoride Salt Cooled High Temperature Reactors (FHRs) and High Temperature Gas-cooled Reactors (HTGRs), due to the higher production rate (up to 10,000 times) than conventional Light Water Reactors (LWRs). Graphitic materials employed as moderator, reflector, and fuel pebbles offer a potential pathway for tritium recovery by serving as a sink for tritium. Prediction of uptake capacity under reactor relevant conditions remains a challenge due to a lack of low partial pressure data and significant inter-grade variability of graphite. This study addresses these gaps by providing a comprehensive characterization of hydrogen (as a tritium surrogate) uptake and release behavior in the A3-3 graphite matrix (GM) used in fuel pebbles. Uptake measurements are performed at reactor relevant temperatures of 600- 800 °C, 1-200 Torr hydrogen pressure, and 15-120 min equilibration time, followed by thermal desorption spectroscopy up to 1100 °C. Uptake experiments at different equilibration times demonstrate the role of kinetics in hydrogen uptake, which can be modeled as a diffusion-with-trapping process. In the thermodynamics limit, the Sips adsorption model is shown to capture the uptake in A3-3 GM well. Our campaign provides a set of new results for hydrogen uptake in A3-3, including limiting uptake capacity at 600 °C, apparent diffusion coefficient at 600 °C, and the first estimates of the FHR/HTGR relevant (600 °C, 20 Pa partial pressure) equilibrium uptake capacity and time to saturation. Desorption data highlights a new site for hydrogen uptake, not observed in nuclear graphite, which we attribute to the non-graphitized binder. Using the Kissinger method, we estimate activation energy for release from the desorption peaks, confirming the activation energy for release from the basal planes and providing the first estimate for the activation energy of release from the binder.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Nuclear waste attributes of near-term deployable small modular reactors

The nuclear waste attributes of near-term deployable SMRs were assessed using established nuclear waste metrics, which are the DU mass, SNF mass, volume, activity, decay heat, radiotoxicity, and decommissioning LLW volumes. Metrics normalized per unit electricity generation were compared to a reference large PWR. Three SMRs, VOYGR, Natrium, and Xe-100, were selected because they represent a range of reactor and fuel technologies and are active designs deployable by the decade’s end. The SMR nuclear waste attributes show both some similarities to the PWR and some significant differences caused by reactor-specific design features. The DU mass is equivalent to or slightly higher than the PWR. Back-end waste attributes for SNF disposition vary, but the differences have a limited impact on long-term repository isolation. SMR designs can vary significantly in SNF volume (and thus heat generation density). However, these differences are amenable to design optimization for handling, storage, transportation, and disposal technologies. Nuclear waste attributes from decommissioning vary depending on design and decommissioning technology choices. Given the analysis results in this study and assuming appropriate waste management system and operational optimization, there appear to be no major challenges to managing SMR nuclear wastes compared to the reference PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The importance of ion kinetic energy for interference removal in ICP-MS/MS

Here, the effect of ion kinetic energy on gas phase ion reactivity with ICP-MS/MS was investigated in order to explore tuning strategies for interference removal. The collision/reaction gases CO 2 , N 2 O and O 2 were used to observe the ion product distribution for 48 elements using an Agilent tandem ICP-MS (ICP-MS/MS) as a function of reaction gas flow rate (pressure) and ion kinetic energy. The kinetic energy of the incident ion was varied by adjusting the octopole bias (V oct ). The three gases all form oxides (MO + ) as the primary product with differing reaction enthalpies that result in distinct differences in the ion energies required for reaction with product ion distributions that vary with Voct. Consequently, by varying the ion kinetic energy (i.e., V oct ), differences in interference reactivity can be used to achieve maximum separation. Three practical application examples were reported to demonstrate how the ion kinetic energy can be varied to achieve the ideal ion product distribution for interference resolution: CO 2 for the removal of 238 U in Pu analyses, CO 2 for the removal of 40 Ar 16 O vs. 56 Fe, and O 2 for the removal of Sm in Eu analyses, analogous to Pu/Am. The results demonstrate how the starting ion energy defined by V oct is an important factor to fully leverage the utility of any given reaction gas to remove interferences in the mass spectrum using ICP-MS/MS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Two Neptunium(III) Mellitate Coordination Polymers: Completing the Series Np–Cf of Trans-Uranic An(III) Mellitates

For this work, two neptunium(III) mellitates, 237 Np 2 (mell)(H 2 O) 9 ·1.5H 2 O (Np-1α) and 237 Np 2 (mell)(H 2 O) 8 ·2H 2 O (Np-1β), have been synthesized from 237 NpCl 4 (dme) 2 by reduction with KC 8 and subsequent reaction with an aqueous solution of mellitic acid (H 6 mell). Characterization by single-crystal X-ray crystallography and UV–vis–NIR spectroscopy confirms that the neptunium is in its +3 oxidation state and both polymorphs are isostructural to the previously reported plutonium mellitates. Of the two morphologies, Np-1α is indefinitely stable in air, while Np-1β slowly oxidizes over several months. This is due to the change in the energy of the metal-ligand charge-transfer absorption exhibited by these compounds attributed to differing numbers of carboxylate bonds to Np(III), where in Np-1β the energy is low enough to result in spontaneous oxidation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Unraveling Gibbsite Transformation Pathways into LiAl-LDH in Concentrated Lithium Hydroxide

The intercalation of lithium (Li+) ions into metallic hydroxides (e.g. gibbsite, Al(OH)3) facilitates the synthesis of layered double hydroxides. Caustic environments, such as in lithium hydroxide (LiOH) solutions, present additional pathways beyond the intercalation of ions, specifically dissolution/precipitation processes. Herein, the phase transformation Al(OH)3 to lithium aluminum hydroxide dihydrate, Li[Al(OH)3)]2OH•2H2O in partially deuterated LiOH solutions is tracked through multinuclear, magic angle spinning, high-field nuclear magnetic resonance spectroscopy (27Al and 6Li MAS-NMR) , X-ray diffraction, and scanning electron microscopy of harvested solids. At high magnetic fields (19.975 T), 27Al MAS-NMR enabled detection and quantification of the reactive interface between Al(OH)3 and Li[Al(OH)3)]2OH•2H2O. 6Li MAS-NMR spectra collected on the harvested solids exhibited exchange mediated narrowing as the reaction time progressed, consistent with transport of Li+ ions between the interface and the bulk Li[Al(OH)3)]2OH•2H2O. Complimentary in-situ 27Al MAS-NMR at lower magnetic fields (7.046 T) utilized unique rotors capable of simultaneous resolution of solution and solid state Al-species in caustic environments. The reaction between Al(OH)3 (0.5 M) and LiOH (3 M) initially exhibited greater dissolution and formation of Al(OH)4- than the precipitation. Over several hours, the supersaturated Al(OH)4- solution precipitated Li[Al(OH)3)]2OH•2H2O. Simultaneous, in-situ observations of the solid component tracked the formation of Li[Al(OH)3)]2OH•2H2O evident by an increase in signal and narrowing of the Oh resonance. These results provides insight into Al precipitation in alkaline media with applications ranging from the reduction of Al content in Hanford site, high-level radioactive waste (HLRW) to the precipitation of LDH materials for application as catalyst materials, drug delivery devices, and ion-ion exchangers.

Graham, Trenton R.↗