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At least 289 records · Page 16

Noble Gas Surface Flux Simulations And Atmospheric Transport

Signatures from underground nuclear explosions or UNEs are strongly influenced by the containment regime surrounding them. The degree of gas leakage from the detonation cavity to the surface obviously affects the magnitude of surface fluxes of radioxenon that might be detected during the course of a Comprehensive Test Ban Treaty On-Site Inspection. In turn, the magnitude of surface fluxes will influence the downwind detectability of the radioxenon atmospheric signature from the event. Less obvious is the influence that leakage rates have on the evolution of radioxenon isotopes in the cavity or the downwind radioisotopic measurements that might be made. The objective of this letter report is to summarize our attempt to better understand how containment conditions affect both the detection and interpretation of radioxenon signatures obtained from sampling at the ground surface near an event as well as at greater distances in the atmosphere. In the discussion that follows, we make no attempt to consider other sources of radioactive noble gases such as natural backgrounds or atmospheric contamination and, for simplicity, only focus on detonation-produced radioxenon gases. Summarizing our simulations, they show that the decay of radioxenon isotopes (e.g., Xe-133, Xe-131m, Xe-133m and Xe-135) and their migration to the surface following a UNE means that the possibility of detecting these gases exists within a window of opportunity. In some cases, seeps or venting of detonation gases may allow significant quantities to reach the surface and be released into the atmosphere immediately following a UNE. In other release scenarios – the ones we consider here – hours to days may be required for gases to reach the surface at detectable levels. These release models are most likely more characteristic of “fully contained” events that lack prompt venting, but which still leak gas slowly across the surface for periods of months.

58 GEOSCIENCES↗

Argonne National Laboratory Site Environmental Report (CY2019)

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2019. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP-88 Version 4 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗

Argonne National Laboratory Site Environmental Report (CY2020)

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2020. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 Version 4.1 computer code, was used in preparing this report.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

High-Sensitivity Nuclear-Structure Studies of Rare Isotopes and Few-Electron Systems

This final technical report presents the work that was completed during the award period for DE-SC0017649 related to the structure of nuclei for both astrophysics and tests of the Standard Model, namely: Using trapped bare (and few-electron) radioactive ions to probe the nature of elec- troweak decay modes under extreme conditions for structure and astrophysics, Providing experimental nuclear structure guidance to nuclear theory in systems relevant for testing the Standard Model, and Working with the HPGe sub-group of the FRIB decay station collaboration, which includes the technical development of a next-generation γ-ray spectrometer.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Argonne National Laboratory Site Environmental Report for Calendar Year 2021

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2021. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 Version 4.1 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗

Argonne National Laboratory Site Environmental Report for Calendar Year 2022

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2022. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 Version 4.1 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗

Argonne National Laboratory Site Environmental Report for Calendar Year 2023

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2023. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗

Argonne National Laboratory Site Environmental Report for Calendar Year 2024

This report discusses the status and accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2024. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, efficiency efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups and on-site members of the public was estimated. Results are interpreted with respect to the origin of radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗

The limits of the nuclear landscape

In 2011, 100 new nuclides were discovered1. They joined the approximately 3,000 stable and radioactive nuclides that either occur naturally on Earth or are synthesized in the laboratory2,3. Every atomic nucleus, characterized by a specific number of protons and neutrons, occupies a spot on the chart of nuclides, which is bounded by drip lines indicating the values of neutron and proton number at which nuclear binding ends. The placement of the neutron drip line for the heavier elements is based on theoretical predictions using extreme extrapolations, and so is uncertain. However, it is not known how uncertain it is or how many protons and neutrons can be bound in a nucleus. Here we estimate these limits of the nuclear landscape and provide statistical and systematic uncertainties for our predictions. We use nuclear density functional theory, several Skyrme interactions and high-performance computing, and find that the number of bound nuclides with between 2 and 120 protons is around 7,000. We find that extrapolations for drip-line positions and selected nuclear properties, including neutron separation energies relevant to astrophysical processes, are very consistent between the models used.

Nazarewicz, Witold↗

The Importance of Inventory for Informing Waste Strategy - 20280

A waste inventory aims to present the best available information about the nature, quantity and location of radioactive wastes and materials at a facility, site or in a country. Quality inventory data is necessary for developing appropriate strategies and plans for managing radioactive wastes and materials, with high standards of protection for people and the environment. The inventory of wastes will change over time due to a wide range of technical, commercial or policy reasons, for example, if reactor lifetimes are extended, if decommissioning strategies change or if new technologies are introduced for managing waste. Inventory information can also improve as better waste characterisation data becomes available. At a facility or site level, good quality inventory data are needed to underpin decision making about the waste management options and to demonstrate compliance with the chosen treatment or disposal route. At a site and country level, knowledge of the inventory can drive the improvement to waste strategy; waste management asset and service planning; and implementation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Production and characterization of thin, self-supporting Si foils for use as targets in radioactive beam experiments

Development of thin, self-supporting silicon foils (both natural and isotopically-enriched) for use as targets in reaction studies with radioactive beams is detailed. Foils with a thickness of ~220 μg/cm 2 were produced using vapor deposition and were floated onto aluminum frames with 10–15 mm diameter holes. During their production, the foil thickness was measured using a quartz crystal monitor. Subsequently, the foil thickness was characterized by particle energy loss measurements and Rutherford backscattering (RBS). These measurements demonstrated that the thickness could be determined to within a 0.5% uncertainty. The elemental purity of the foils was assessed using RBS and X-ray photoelectron spectroscopy. Lastly, this analysis demonstrated that the foils have 87%–90% silicon abundance.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Buffer Material Degradation in High Level Radioactive Waste Disposal - 18454

The concept of geological disposal for high-level radioactive waste (HLW) is based on a multi-barrier system comprising the natural geological barrier that repository host rock provides and its surroundings, as well as an engineered barrier system (EBS) that includes waste form, waste canisters, buffer materials, and backfill. The multi-barrier system is expected to perform its desired functions to isolate waste from the biosphere and achieve HLW disposal. Because of the interaction processes of radionuclide decay heat with hydraulic and water-mineral reactions in the radioactive waste repository, the bentonite buffer degradation resulting from the external temperature and pressure warrants investigation. However, seldom addressed in Taiwan is the influence of thermo-hydro-chemical (T-H-C) processes on buffer material degradation in the EBS of HLW disposal repositories as related to smectite clay dehydration. Therefore, we adopted the chemical kinetic model of smectite dehydration to calculate the amount of water expelled from smectite clay minerals caused by higher temperatures of waste decay heat. We discovered that smectite clay could cause volume shrinkage because of a loss of interlayer water, which can lead to bentonite buffer compression. The results of this study may be used in advanced research on the evolution of bentonite degradation for both performance assessments and safety analyses of final HLW disposal in Taiwan. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Crystalline Ceramic Waste Forms: Reference Formulation Report

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets with CS/LN/TM combined waste stream with Mo and Zr removed. Waste streams that contain Mo must be produced in reducing environments to avoid Cs-Mo oxide phase formation. Waste streams without Mo have the ability to be melt processed in air. A path forward for further optimizing the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere, and controlled heat treatment schedules are anticipated to improve the targeted elemental partitioning.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO 2 and Al 2 O 3 were observed distributed in a network of fine grains with small residual pores. The titanate phases that incorporate M +3 rare earth elements were observed to be distinct phases (ex. Nd 2 Ti 2 O 7 with less degree of substitution as compared to the more homogeneous melt processed samples where a high degree of substitution and variation of composition within grains was observed. Liquid phase sintering was enhanced in reducing gas environments and resulted in large (10-200 microns) irregular shaped grains along with large voids associated with the melt process; SPS and HP samples exhibited finer grain size with smaller voids. Metallic alloys were observed in the bulk of the sample for SPS and HP samples, but were found at the bottom of the crucible in melt processed trials. These results indicate that for a first melter trial, the targeted phases can be formed in air by utilizing Ti/TiO 2 additives which aid phase formation and improve the electrical conductivity. Ultimately, a melter run in reducing gas environments would be beneficial to study differences in phase formation and elemental partitioning.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Crystalline Ceramic Waste Forms: Report Detailing Data Collection In Support Of Potential FY13 Pilot Scale Melter Test

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to summarize the data collection in support of future melter demonstration testing for crystalline ceramic waste forms. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. The principal difficulties encountered during processing of the “reference ceramic” waste form by a melt and crystallization process were the incomplete incorporation of Cs into the hollandite phase and the presence of secondary Cs-Mo non-durable phases. In the single phase hollandite system, these issues were addressed in this study by refining the compositions to include Cr as a transition metal element and the use of Ti/TiO 2 buffer to maintain reducing conditions. Initial viscosity studies of ceramic waste forms indicated that the pour spout must be maintained above 1400°C to avoid flow blockages due to crystallization. In-situ electron irradiations simulate radiolysis effects indicated hollandite undergoes a crystalline to amorphous transition after a radiation dose of 10 13 Gy which corresponds to approximately 1000 years at anticipated doses (2 10 10 -2 10 11 Gy). Dual-beam ion irradiations employing light ion beam (such as 5 MeV alpha) and heavy ion beam (such as 100 keV Kr) studies indicate that reference ceramic waste forms are radiation tolerant to the α-particles and β-particles, but are susceptible to a crystalline to amorphous transition under recoil nuclei effects. A path forward for refining the processing steps needed to form the targeted phase assemblages is outlined in this report. Processing modifications including melting in a reducing atmosphere with the use of Ti/TiO 2 buffers, and the addition of Cr to the transition metal additives to facilitate Cs-incorporation in the hollandite phase. In addition to melt processing, alternative fabrication routes are being considered including Spark Plasma Sintering (SPS) and Hot Isostatic Pressing (HIP).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

RANGERS: Methodology for Design and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF: Synthesis Report

Salt formations have long been recognized as a highly favorable host rock for the final disposal of high-level radioactive waste (HLW) in deep geological repositories. Their unique properties, including exceptional impermeability, self-healing capabilities, and thermal conductivity, make them a reliable natural barrier for the deep disposal of radioactive waste. This report focuses on the development and application of a methodology for assessing the integrity and per formance of the Engineered Barrier System (EBS) within salt-based repositories, a critical component of the multi-barrier system ensuring safe radioactive waste disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Measurement of natural radionuclides in Malaysian bottled mineral water and consequent health risk estimation

The radionuclides of {sup 226}Ra, {sup 232}Th and {sup 40}K were measured in ten mineral water samples, of which from the radioactivity obtained, the ingestion doses for infants, children and adults were calculated and the cancer risk for the adult was estimated. Results showed that the calculated ingestion doses for the three age categories are much lower than the average worldwide ingestion exposure of 0.29 mSv/y and the estimated cancer risk is much lower than the cancer risk of 8.40 × 10{sup −3} (estimated from the total natural radiation dose of 2.40 mSv/y). The present study concludes that the bottled mineral water produced in Malaysia is safe for daily human consumption.

61 RADIATION PROTECTION AND DOSIMETRY↗

Dose Estimation in Recycling of Decontamination Soil Due to Fukushima NPS Accident as Coastal Levees - 18138

After the Fukushima Nuclear Power Station accident, large quantities of radiocesium-contaminated soil generated from decontamination activities have been stored in the Fukushima prefecture. To complete the final disposal of decontamination soil, the Ministry of the Environment of Japan presented policy to reduce the disposal volume by using low-level decontamination soil as recycled materials. The use will be limited to civil engineering structures in public projects, such as coastal levee and road embankment, and be managed by the public authority. However, there has been no example or safety assessment of recycling of decontamination soil. To contribute to guideline development for the recycling by the Ministry of the Environment, we conducted the safety assessment for recycling of decontamination soil as coastal levee. First, we evaluated additional doses to workers and the public in the case of construction and service of coastal levee based on Japanese construction standards. From the result, the radioactive cesium concentration level of recycled materials, with which all additional doses meet the radiation criterion of 1 mSv/y, was derived to be 6,800 Bq/kg. For general designs of coastal levee in Japan, additional doses to the public in the case of service were below 10 μSv/y with derived radioactivity level. Finally, the radiation safety for collapse of coastal levee caused by natural disasters was evaluated to be on the safe side. With derived radioactivity level, additional doses in the case of disaster were confirmed to be below 1 mSv/y. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗