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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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Inventory and solubility of fission products in molten lead to quantify source term in lead-cooled fast reactors
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Combining physics-based and data-driven techniques for reliable hybrid analysis and modeling using the corrective source term approach
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Experimental validation of Monte Carlo NaI(Tl) detector efficiency responses with surrogate 137Cs environmental contamination source term
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Techniques for solving static Klein-Gordon equation with self-interaction λϕ4 and arbitrary spherical source terms
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Fuel Assembly Reference Information for SNF Radiation Source Term Calculations
This report consolidates publicly available information related to pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear fuel physical characteristics and operating conditions. The intent of this report is to provide reference information for use in shielding analyses of spent nuclear fuel transportation packages and storage casks.
A Study on the Characteristics of the Radiation Source Terms of Spent Fuel and Various Non-Fuel Hardware for Shielding Applications
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Impacts of LEU+ and HBU Fuel on Decay Heat and Radiation Source Term
Using low-enriched uranium with 235 U enrichment slightly greater than 5.0 wt % (LEU+) fuel is desirable for light water reactors (LWR) because it can enable longer cycles and/or smaller fresh fuel batches. This report analyzes the impact of LEU+ on back-end considerations, such as its impact on the spent fuel pool, and on the thermal and shielding performance of dry storage systems. For both pressurized water reactor (PWR) and boiling water reactor LEU+ fuel, this report concludes that additional cooling time or regionalized loading plans are required to maintain the same level of thermal and shielding performance of dry storage systems loaded with LEU fuel. For PWR fuel, this report concludes that the impact of LEU+ fuel on the time-to-boil of a spent fuel pool (SFP) is minimal and does not require considerable changes in SFP management.
Mapping Arctic Marine Greenhouse Gas Emissions for Use as a Source Term in Global Climate and Earth System Models.
Abstract not provided.
MELCOR Integrated Non-LWR Accident Progression and Source Term Analysis.
Abstract not provided.
Accident Source Terms
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Complete Development of Critical Capabilities for TRISO Fission Product Source Term Calculations and Quantify Mechanisms for Pd Penetration of SiC
Overall fission product (FP) release will be an important consideration for the licensing and deployment of advanced reactors utilizing tristructural isotropic (TRISO) fuels. This work focuses on enhancing and applying the BISON models needed to predict FP transport within TRISO particles and particle failure probability, both of which factor directly into release predictions. Specifically, this report details (1) the development of the models needed to predict palladium (Pd) conservation at the engineering scale and the application of those models to characterize Pd fluxes for input into a mechanistic multiscale model for Pd penetration; (2) the refinement of sorption mass transfer models and the development of models for trapping in porous layers, which were applied and compared to particle scans from AGR-2 to provide proof of concept for a method of particle-scale validation that may reduce uncertainties compared to compact-scale validation using data from integral effects tests; (3) the development of a failure-statistics-informed, mesh-independent methodology for applying smeared cracking, enabling further study of the localized multiphysics behaviors associated with cascading particle failure mechanisms; and (4) the preliminary characterization of those coupled multiphysics particle failure behaviors using smeared, nonretentive diffusivities to provide a baseline for future study and to guide ongoing engineering applications.
Development of a CRAB-MELCOR Framework for Utilization in Support of Estimating a Realistic Mechanistic Source Term for Microreactors
Research and development activities are being performed to link important advanced non-light-water microreactor technology research and development activities sponsored by the U.S. Department of Energy’s Office of Nuclear Energy with regulatory requirements and critical licensing needs likely to impact the deployment of a domestic commercial microreactor fleet. This report presents the progress of ongoing work to perform a gap analysis of the coupling of the Nuclear Regulatory Commission’s Comprehensive Reactor Analysis Bundle code suite and MELCOR code. This key Fiscal Year 2023 research reflects microreactor regulatory needs and challenges noted by industry, Department of Energy, and Nuclear Regulatory Commission stakeholders.
Source Term Development for the MELCOR Severe Accident Code
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Advanced Reactor Source Term Modeling
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Modeling of Source Term Uncertainties in the Pedestal Region for Neutron Multiplication Monitoring in Fukushima Daiichi Fuel Debris Retrieval Operations [Slides]
Modeling approach has significant impact on expected neutron spectral flux. Presence of water in Unit 3 greatly reduces neutron signal. Neutron flux from Homogeneous and Lattice models behaved more similarly as opposed to Layered model.
Assessment of Source Term Uncertainties for Criticality Monitoring Simulations of Fukushima Daiichi Fuel Debris Retrieval Operations [Slides]
Modeling approach has significant impact on expected neutron spectral flux. Presence of water in Unit 3 greatly reduces neutron signal. Neutron flux from Homogeneous and Lattice models behaved more similarly as opposed to Layered model.