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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

NA-22 Quarterly Report: April-June, 2020

At LLNL we contribute to two projects in NNSA-IAEC Science and Technology Working Group Area V: Basic Science of Waste Management & Subsurface Science. The motivation behind Topic Area V is to evaluate the scientific and safety case for an intermediate borehole to house Israel’s nuclear waste. WM1 (SNL lead) is Thermomechanical Damage (initially called Damage Accumulation and Wellbore Stability). This work focuses on evaluating the damage to the host rock induced by excavation and heat released from a possible radioactive waste repository/borehole. WM2 (LLNL lead) is Radionuclide Facilitated Transport in Carbonate Rock (initially called Colloidal Transport of Radionuclides). This work focuses on better understanding the transport of radionuclides and colloids from a nuclear waste repository/borehole in a vadose zone environment through experimentation and modeling. In the second quarter (this report) our work has been delayed in part due to COVID-19. LLNL went into shelter in place (SIP), minimum safe operations mid-March. Operations at LLNL have been slowly ramping up with 50% of the workforce back onsite in limited capacity and telecommuting widely used. The good news is that our labs opened up the last week in June and experiments are expected to start up again for WM2 Radionuclide Facilitated Transport in July. LLNL’s contribution to WM1 Thermochemical Damage modeling efforts also started back up in late June. We expect to be able to catch up on much of our work by the end of next quarter. The three national laboratories communication on a bi-monthly basis to make sure Area V projects remain on track and we have a monthly conference call with our Israeli counterparts to make sure we are communicating priorities and coordinating project details.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

NA-22 Quarterly Report: April-June, 2020

At LLNL we contribute to two projects in NNSA-IAEC Science and Technology Working Group Area V: Basic Science of Waste Management & Subsurface Science. The motivation behind Topic Area V is to evaluate the scientific and safety case for an intermediate borehole to house Israel’s nuclear waste. WM1 (SNL lead) is Thermomechanical Damage (initially called Damage Accumulation and Wellbore Stability). This work focuses on evaluating the damage to the host rock induced by excavation and heat released from a possible radioactive waste repository/borehole. WM2 (LLNL lead) is Radionuclide Facilitated Transport in Carbonate Rock (initially called Colloidal Transport of Radionuclides). This work focuses on better understanding the transport of radionuclides and colloids from a nuclear waste repository/borehole in a vadose zone environment through experimentation and modeling. In the second quarter (this report) our work has been delayed in part due to COVID-19. LLNL went into shelter in place (SIP), minimum safe operations mid-March. Operations at LLNL have been slowly ramping up with 50% of the workforce back onsite in limited capacity and telecommuting widely used. The good news is that our labs opened up the last week in June and experiments are expected to start up again for WM2 Radionuclide Facilitated Transport in July. LLNL’s contribution to WM1 Thermochemical Damage modeling efforts also started back up in late June. We expect to be able to catch up on much of our work by the end of next quarter. The three national laboratories communication on a bi-monthly basis to make sure Area V projects remain on track and we have a monthly conference call with our Israeli counterparts to make sure we are communicating priorities and coordinating project details.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Effect of zeolite type, temperature, and pH on Stage III glass alteration behavior for two nuclear waste glasses

We report long-term laboratory scale static alteration tests (up to 1500 d) were conducted on two low-activity waste glasses (LAWA44 and IDF18-A161) at S/V = 2000 m –1 . Tests were seeded with either analcime, clinoptilolite, zeolite P1, or zeolite P2 to induce Stage III glass alteration behavior. The effect of several parameters on glass Stage III behavior was studied with consideration of several variables, including temperature (22 °C – 90 °C) and pH (floating or initially set using KOH at values from 9.5 to 11.5). Temporal solution data and solids characterization at the end of the experiments are reported. Solution data demonstrated the occurrence of a sustained rate acceleration at temperatures down to 40 °C and the activation energy for these Stage III rates was determined for both glasses. Three major secondary phases were identified at the end of the experiments for both glasses: zeolite P1, zeolite P2, and/or analcime (along with minor phases: zeolites, clays, carbonates). The identification of the crystalline phases at the end of the experiment often differed from the zeolite seed phase. In addition, the use of the Avrami equation showed unimpeded zeolite growth (3-dimensional) in certain conditions, suggesting that in certain conditions the glass dissolution rate is controlled by zeolite formation kinetics whereas in other conditions the glass corrosion process impeded zeolite formation kinetics. The morphological evolution of the samples revealed the growth of the zeolites which completely fill the intergranular space in the powder bed. These results are discussed with regards to previous seeded and unseeded Stage III dissolution rate experiments performed on low- and high-activity nuclear waste glasses.

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DOE OSTI2021

Characterization of the Sulfur-Saturated Melt Versions of the HFG1 Study Glasses

The U.S. Department of Energy (DOE) is responsible for building the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is temporarily stored in 177 underground tanks. The Office of River Protection (ORP) has requested that the Savannah River National Laboratory (SRNL) contribute to current glass studies to support successful startup of the WTP, due to SRNL’s recognized capabilities and expertise for glass waste form development. As part of this effort, the Pacific Northwest National Laboratory (PNNL) is developing, batching, and fabricating simulated high-level waste (HLW) glasses to generate the associated property data needed to validate or identify areas of potential characterization improvements in the current glass property/ composition models. Currently, the models cover only a small fraction of the waste compositions projected in the Hanford tank farm; therefore, the models must be expanded to include high fluoride compositions to successfully complete the WTP mission. SRNL support of this work is defined in the Task Technical and Quality Assurance Plan (TTQAP). This report provides results from the chemical analyses of a series of sulfur-saturated melt (SSM) versions of simulated nuclear waste glasses fabricated at PNNL. The glasses were selected as part of a broader study of the influence of glass composition on chemical durability, sulfur retention, and other properties. The glasses were designated the High Fluoride Glasses-1 (HLG1).

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DOE OSTI2024

Regulatory Spike Testing of RPP-WTP LAW and HLW Glasses for Compliance with Land Disposal Restrictions, VSL-03R3760-1, Rev. 1

The United States Department of Energy’s (DOE’s) Hanford site is the current location of storage of about 50 million gallons of mixed waste. This waste is stored in underground tanks at the Hanford site and is both a listed and characteristic waste, as defined in 40 CFR Part 261, and a dangerous waste according to Chapter 173-303 of the Washington Administrative Code (WAC). The waste is also subject to the Land Disposal Restrictions (LDR) (40 CFR Part 268 and WAC 173-303-140). The River Protection Project - Waste Treatment Plant (RPP-WTP) will provide DOE with a means for treating this waste by vitrification (for subsequent disposal). The tank waste will be partitioned into low and high activity fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed of in an engineered facility on the Hanford site while the IHLW product will be directed to the national deep geological disposal facility for high level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment.

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DOE OSTI2020

Flammable Gas Generation and Control at the Idaho Completion Project Legacy Combustible Gas Generation - 20191

On April 11, 2018, four drums containing transuranic waste at the Idaho National Laboratory underwent over-pressurization, ejecting their lids and spreading radiological waste within a facility. An investigation has found that waste in the drums generated methane gas, which contributed to the event. Subsequent to the investigation, the potential for drums to have methane and other flammable analytes whose concentrations could approach or exceed the lower flammability limit (LFL) and the adequacy of the controls to prevent or mitigate a possible deflagration was evaluated. An extensive review of the historical records was performed to determine how many drums exceed the LFL. The historical record identified a small quantity of drums that exceeded the LFL for xylene, hydrogen, and methane. The primary codified applicable code or standard for handling drums is 29 CFR 1910.120 (j) 'Handling Drums and Containers.' It is used throughout the commercial and government sector. This code is integrated throughout the Fluor Idaho safety management programs and procedures to control the Standard Industrial Hazards (SIHs) associated with drum and container handling. Example requirements include: - Drums and containers are inspected, and their integrity assured prior to being moved. - Site operations are organized to minimize the amount of drum or container movement. - When there is a reasonable possibility of flammable atmospheres being present, material handling equipment and hand tools are of the type to prevent sources of ignition. Drums and containers under pressure, as evidenced by bulging or swelling, are not moved until the cause for excess pressure is determined and appropriate containment procedures have been implemented to protect employees from explosive relief of the drum. Drums which exceed the WIPP FGA limit have NCRs associated with them. Once a drum fails flammable gas sampling, an NCR is initiated. The NCR drives placement into NCR dense pack rows. A standing order drives placement of the FGA failures to a single planer segregation location. These drums are not allowed for further processing or shipment until the NCR has been cleared. Once in the single planer rows, follow-up FGA testing is performed by CCP. The NCR is dispositioned by follow-up sampling occurring after adequate time has been provided for the gas to diffuse through the filter assembly, or, in some cases, the drums are repackaged into additional drums to reduce the source term. To date, the ICP project has performed approximately 1.5 million drums moves without a deflagration event due to combustible gas generation. Flammable gas generation in a transuranic waste drum is not unique to the ICP but is common across the DOE complex. Based on the experience at Idaho, application of industry standards is sufficient to control the risk of drum deflagration due to drum movements. (authors)

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DOE OSTI2020

Iodine Co-Precipitation with Calcium Carbonate in the Presence of Silica Ions - 20360

Historical releases of I-129 to the subsurface at the U.S. Department of Energy Hanford Site have resulted in large dilute plumes that cover an area of over 50 km{sup 2}. The most concentrated I-129 plume is associated with the 200 West Area in the 200-UP-1 operable unit of the Hanford Central Plateau, where peak concentrations have been measured up to 30 times the drinking water standard of 1 pCi/L. The mobility of iodine depends on many factors, including chemical speciation, pH, redox conditions, as well as the presence of organic matter, calcium carbonate minerals and microbial activities. Speciation measurements estimate that the majority of the iodine (∼76%) in Hanford groundwater exists as iodate. Information on processes that affect iodate mobility in the Hanford subsurface support evaluation of environmental management options for the I-129 plumes. Iodate can be incorporated into calcium carbonate, which is a potential mechanism of naturally attenuating radioiodine in groundwater. However, the silica content in porewater may impact the contaminant-calcium carbonate incorporation process. Silica is one of the most abundant elements in nature and can have an impact on chemical weathering of alkaline-earth carbonates in the environment. Hence, this research aims to advances knowledge of the iodine co-precipitation process with calcium carbonate in the presence of silica. This information supports the technical basis evaluation of natural attenuation for I-129. Samples for these studies were prepared using calcium carbonate-forming solutions that included certified grade sodium meta-silicate nonahydrate (Na{sub 2}SiO{sub 3}), calcium chloride dehydrate (CaCl{sub 2}.2H{sub 2}O), sodium carbonate (NaCO{sub 2}), and sodium hydroxide (NaOH) solutions for pH adjustments. Triplicate samples were prepared with silica concentrations in the solution of 0 mM, 0.5 mM and 20 mM, mixed with an iodate standard (1000 ppm of IO{sub 3}{sup -} in H{sub 2}O) and sampled over a one-week period. Preliminary data analyses suggest that less iodine remains in solution in the presence of silica (average remaining aqueous fraction 0.23±0.03 at 20 mM of Si) relative to samples without silica (0.52±0.06); an indication that more iodine has been sorbed or incorporated into precipitates. However, the mechanisms for this behavior are still under investigation and need to be interpreted relative to pore water compositions that are representative of Hanford site-specific conditions. Scanning electron microscope equipped with energy dispersive spectroscopy (SEM-EDS) analyses were performed to observe the solid phase morphology in these experiments and showed rhombohedral calcite crystals covered with amorphous Si floccules. (authors)

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DOE OSTI2023

A novel protocol to recycle zirconium from zirconium alloy cladding from used nuclear fuel rods

A potential new zirconium recycling protocol has been demonstrated using unirradiated Zircaloy-based claddings and depleted uranium oxide for the chemical removal of zirconium alloy cladding in used nuclear fuel (UNF) rods from light water reactors. This protocol is based on the application of three new scientific findings. First, a new lower temperature chlorination reaction of zirconium with sulfur monochloride is described. Second, the high solubility of zirconium chloride in thionyl chloride is used to separate it from uranium oxide fuel and fission and activation contaminants. Finally, in the third step, the zirconium chloride is purified by simple recrystallization from thionyl chloride. Utilizing sulfur chloride solvents for a lower temperature liquid-based chemical digestion and purification of zirconium alloy claddings reduces technical complications experienced by current high temperature gas phase chlorination strategies, such as contamination of product streams. This novel protocol has been demonstrated on a 3–50 g scale of unirradiated zirconium alloy, with no significant change in the time required for complete chlorination (3–4 h). This zirconium chlorination protocol has also been performed in the presence of depleted uranium oxide pellets. The depleted fuel pellets do not affect zirconium chlorination, and the chlorination protocol does not chemically or physically alter the fuel pellets. Finally, a preliminary description of an industrial protocol to recycle nuclear grade zirconium from UNF rods is presented.

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DOE OSTI2021

Product Consistency Test Results for the LAW ALG Glasses

This report provides the results of the Product Consistency Test leachates from the Low-Activity Waste Algorithm glasses, a series of simulated nuclear waste glasses designed and fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister centerline cooled versions of the glasses. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford.

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DOE OSTI2024

Towards informatics-driven design of nuclear waste forms

Informatics-driven approaches, such as machine learning and sequential experimental design, have shown the potential to drastically impact next-generation materials discovery and design.

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DOE OSTI2025

Brine Availability Test in Salt (BATS) (FY25 Final Closeout)

This fiscal year 2025 (FY25) closeout report summarizes the final experiments conducted and data collected in phase 2 of the Brine Availability Test in Salt (BATS 2). The design and interpretation of the test was previously funded by the US Department of Energy’s Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Disposition Program, under the Disposal Research and Development (R&D) program of the Office of Spent Fuel & Waste Science and Technology (SFWST) but they are not funding it anymore (since May 2025). The experiment was located underground at the Waste Isolation Pilot Plant (WIPP), southeastern New Mexico, which is a DOE Office of Environmental Management (DOE-EM) site managed by the Carlsbad Field Office (CBFO). DOE-EM funds the WIPP Test Coordination Office (TCO), which provided critical implementation support for the continued execution of BATS. A high-level summary of processes expected in a salt repository are given in Kuhlman & Mills (2025).

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DOE OSTI2020

Containers and refractory metal coating therefore for containing radioactive materials

Fabricating structural components for a spent nuclear fuel container using the steps of forming cylindrical or rectangular channels to produce a structural component for a spent nuclear fuel container and applying a coating that includes tantalum-based material to the cylindrical or rectangular channels.

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DOE OSTI2020

Calculating Potential Radiological Emissions for Waste Management Activities at INL - 20068

At Idaho National Laboratory (INL), work involving radionuclides is evaluated for potential emissions from a project in order to comply with the National Emission Standards for Hazardous Air Pollutants (NESHAP) regulations, 40 CFR 61 Subpart H. Emission calculations are documented in an Air Permitting Applicability Determination (APAD) to analyze unmitigated and mitigated emissions and determine if an Application to Construct (ATC) or continuous monitoring is required. To calculate the unmitigated and mitigated emissions, a spreadsheet was developed to provide ease in determining potential emissions by providing the maximum operating temperature and the material being used. The spreadsheet aids in determining the potential emissions for research projects and waste management activities at Materials and Fuels Complex (MFC) and other locations across the INL site. Furthermore, it can also be used for periodic confirmatory measurements (PCM) to justify low emissions. Elements that factor into the unmitigated and mitigated calculations include the amount of each radionuclide used (in curies or grams), specific activity (if amount is given in grams), the temperature the material is heated to in Celsius, the dose conversion factor which is derived from Clean Air Act Assessment Package - 1988 (CAP-88) modeling, and the number of HEPA filters used for mitigated measures. The main drivers for calculating the unmitigated emissions for a project are the amount used per radionuclide, the maximum operating temperature, and the location of the work. The maximum operating temperature determines the airborne release factor which is dependent on the physical state of the radionuclide. Prior to October 2017, if the radionuclide was heated to greater than 100 deg. C, the radionuclide was assumed to be a gas, which has the highest airborne release factor. This assumption would be overly conservative for radionuclides with high melting and boiling points, which provided a challenge to demonstrate low emissions. In October 2017, the Environmental Protection Agency (EPA) approved an alternative method for INL. This method allows the airborne release factor to be determined by using the melting point and 90% of the boiling point of the radionuclide. This methodology was included in the spreadsheet to allow unmitigated emission calculations for APADs to be completed more efficiently and effectively. Results show a reduction in time completing air emission calculations as well as lower total emissions across all facilities at INL. MFC annual emissions were reduced by 62% from the previous year and Research and Education Campus (REC) facilities were reduced by 38% due to implementation of the approved alternative method. Time spent on APADs, PCMs, and documentation for the annual NESHAP report was also reduced significantly. The spreadsheet provided in Table I provides the potential emission calculations for the 'Advanced Retrieval and Disposition Techniques for Remote Handled Mixed Low Level Waste (RH MLLW) at the Radioactive Scrap and Waste Facility (RSWF)' project. Calculations show the Potential Effective Dose Equivalent (PEDE) at RSWF to be 7.27 E-04 mrem/yr (7.27 E-09 Sv/yr) which is well below the 0.1 mrem/yr threshold. (authors)

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DOE OSTI2021

Characterization of the Sulfur-Saturated Melt Versions of the LAW Phase 5 Glasses

This report provides results from the chemical analyses of a series of sulfur-saturated melt versions of the LAW Phase 5 glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. Results from the chemical analyses of the wash solution resulting from the preparation of these glasses are also included. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford.

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DOE OSTI2024

Proceedings of the 13 th US/German Workshop on Salt Repository Research, Design, and Operation

This report summarizes the proceedings of the 13 th US/German Workshop on Salt Repository Research, Design, and Operation hosted by Sandia National Laboratories on June 20-23, 2023, in Santa Fe, New Mexico, USA. Over 60 participants attended, representing Germany, United States, the Netherlands, Australia, and the United Kingdom, along with the IAEA. The purpose of the US/German Workshop is to foster in-person collaboration and dialogue amongst salt repository researchers and nuclear waste disposal implementers across international organizations. The workshop included five sessions of topical presentations and two breakout sessions to promote additional discussion on compelling topics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Artificial Intelligence a D and D Enabler - 20552

The paper addresses the development of a specific search engine dedicated to Decommissioning and Dismantling (D and D) projects. The solution combines advanced model based system engineering methodologies and a data centric approach. It is based on an appropriate use of the last Natural Language processing techniques, the most innovative artificial intelligence algorithms and open source framework to handle large volume and diversity of data met in the case of a nuclear infrastructure we have to dismantle. (authors)

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DOE OSTI2021

Interactive web mapping tools and custom subsurface cross-sections for interdisciplinary geologic investigation

Using Python-based geospatial analytics, open-source web mapping technologies, geophysical data models, and subsurface stratigraphy models from the Regional Geology Geologic Framework Model database assembled by Los Alamos National Laboratory, we developed a suite of web-based geologic investigation tools to identify and understand subsurface structures and geophysical properties concerning salt and shale formations within the contiguous United States. Coupled with a web map interface, these tools allow for the interactive visualization of various geologic data and demonstrate the ability to quickly generate custom subsurface cross-sections, borehole charts, and diagrams for azimuthal orientation data. These capabilities were developed for stakeholder and researcher use to facilitate informed decision making for spent nuclear waste disposition. However, these capabilities provide a flexible model for a variety of subsurface investigation needs, and we have demonstrated this flexibility by adapting these tools to meet visualization needs for various subsurface models within a web-based platform.

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DOE OSTI2025

Results of Re-evaluation of FEPs Related to Higher Fissile Content in HLW Glass at SRS

One of the objectives of the United States Department of Energy Office of Nuclear Energy’s Office of Spent Fuel and High-Level Waste Disposition is to better understand the technical bases, risks, and uncertainties associated with the safe and secure disposition of spent nuclear fuel and high-level radioactive waste. Domestic defense and research activities have generated a few thousand metric tons of spent nuclear fuel and high-level radioactive waste, much of which has been or will be processed and vitrified into high-level waste glass. The Nuclear Waste Policy Act 1982 makes the Department of Energy responsible for disposal of these materials.

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