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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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A High-Throughput Hot Surface Ion Source for Electromagnetic Radioisotope Separation
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Ionization Efficiency Measurement and Optimization of a Thermal Ion Source for Radioisotope Electromagnetic Separation
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Official Release of The Radioisotope Power System Dose Estimation Tool (RPS-DET)
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Modeling Ionizing Radiation for Spaceflight Dynamic Radioisotope Power Systems
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Total Radioisotope Yield Calculator
An initial number of fissions on a fissile/fissionable material, defined by ENDF library availability, is used to produce an initial inventory of fission products (FPs). The FPs are decayed using full, analytical solutions to the Bateman equations to calculate the FP inventory at a user-specified time. Output can contain information about activity, mass, dose rate, and gamma-ray intensities based on user-defined parameters, given as .txt or .xlsx file formats. Full execution and display of results takes 5-10 seconds, depending on the complexity of the input parameters.
The Sensitivity of Three Terrestrial Bacteria to a Multi-Mission Radioisotope Thermoelectric Generator
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Production of Medically Desirable Radioisotopes in the EIRENE Molten Salt Reactor
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Radiological Design and Planning for the Radioisotope Processing Facility
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An Interactive Virtual Radioisotope Processing Facility
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A 15-Mev/nucleon Iso-Cyclotron for Security and Radioisotope Production
Cargo inspection systems exploit the broad bremsstrahlung spectrum from a 6-10 MeV, low-duty cycle electron accelerator which in the presence of significant backgrounds presents challenges in image and material identification. An alternative approach is to use ions which can excite nuclear states either directly, or through generation of secondary high-energy signature gammas produced from nuclear interactions in a target. RadiaBeam is designing a compact sector isocyclotron 1.25 m in radius, with high-gradient cavities to accelerate multi-ion species up to 15-20 MeV/u with large turn-to turn, centimeter-level separation for low-loss extraction without lossy foil stripping. A strong-focusing radial field profile will be optimized in a separated-sector format for control over machine tune simultaneous with isochronous orbist requirements for high-current (~0.5 milliamp) operation. Innovation in injection will be introduced to replace the high-loss central region. Non-security applications of the cyclotron include medical isotope production, ion radiobiology, as well as material science research and ion instrumentation development.
Radioisotope production at SNS (RIPS)
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Evaluation of Dose- and Risk-Based Groundwater Cleanup Levels for Low Energy Beta Radioisotopes
Low-energy beta-emitting radionuclides that were released historically during reactor operations and plutonium separations activities at some U.S. Department of Energy sites have migrated into the groundwater, forming contaminant plumes that are subject to federally regulated remediation actions. At the Hanford Site, the low-energy beta-emitting radionuclides include iodine-129, technetium-99, chlorine-36, carbon-14, and tritium (H-3). All are highly mobile in the subsurface, and except for tritium, and have very long half-lives—thousands to millions of years. The geochemistry and transport behavior of these contaminants in the subsurface present significant challenges for remediation of groundwater to federal drinking water standards (DWS)—the appropriate or relevant and applicable requirements (ARARs) for cleanup. For some of the low-energy beta-emitter contaminants, particularly iodine-129, cleanup and restoration of groundwater to DWS may not be attainable within a reasonable timeframe using currently available treatment technologies.
Radioisotope Thermoelectric Generator Accident Pre-Assessed Scenario
This scenario was drafted for inclusion in a revision of Federal Radiological Monitoring and Assessment Center (FRMAC) Assessment Manual, Volume 2 - Pre-Assessed Default Scenarios. The contents of this scenario were reviewed and approved by the FRMAC Assessment Working Group in March 2020. The scenario is being issued separately from the full volume ahead of the Mars 2020 launch. The full volume will be published in the future once all scenarios are complete.
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.
Preliminary Radioisotope Screening for Off-site Consequence Assessment of Advanced Non-LWR Systems
Currently a set of 71 radionuclides are accounted for in off-site consequence analysis for LWRs. Radionuclides of dose consequence are expected to change for non-LWRs, with radionuclides of interest being type-specific. This document identifies an expanded set of radionuclides that may need to be accounted for in multiple non-LWR systems: high temperature gas reactors (HTGRs); fluoride-salt-cooled high-temperature reactors (FHRs); thermal-spectrum fluoride-based molten salt reactors (MSRs); fast-spectrum chloride-based MSRs; and, liquid metal fast reactors with metallic fuel (LMRs) Specific considerations are provided for each reactor type in Chapter 2 through Chapter 5, and a summary of all recommendations is provided in Chapter 6. All identified radionuclides are already incorporated within the MACCS software, yet the development of tritium-specific and carbon-specific chemistry models are recommended.
Quantitative Assessment for Advanced Reactor Radioisotope Screening Utilizing a Heat Pipe Reactor Inventory
This report documents a method for the quantitative identification of radionuclides of potential interest for accident consequence analysis involving advanced nuclear reactors. Based on previous qualitative assessments of radionuclide inventories for advanced reactors coupled with the review of a radiological inventory developed for a heat pipe reactor, a 1 Ci activity airborne release was calculated for 137 radionuclides using the MACCS 4.1 code suite. Several assumptions regarding release conditions were made and discussed herein. The potential release of a heat pipe reactor inventory was also modeled following the same assumptions. Results provide an estimation of the relative EARLY and CHRONC phase dose contribution from advanced reactor radionuclides and are normalized to doses from equivalent releases of I-131 and Cs-137, respectively. Ultimately, a list of 69 radionuclides with EARLY or CHRONC dose contributions at least 1/100th that of I-131 or Cs-137, respectively – 48 of which are currently considered for LWR consequence analyses – was identified of being of potential importance for analyses involving a heat pipe reactor.