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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 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Benchmark Specifications for Select Experiments Conducted at the Kansas State University Gallium Thermal-hydraulic Experiment Facility

The Department of Energy (DOE) – Nuclear Energy University Programs (NEUP) supported the creation and operation of the Gallium Thermal-hydraulic Experiment (GaTE) facility at Kansas State University (KSU) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. GaTE was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements describing stratification behavior in the coolant were collected for use as experimental benchmarks in validation efforts for computational models. The results of these tests contribute to a greater understanding of thermal stratification behavior of liquid metal under various configurations and operating conditions. This report provides a complete description of the benchmark problem, including all necessary details and description of a set of four forced flow and four natural circulation tests and measured data for comparison with model results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Benchmark Specification for Select Experiments Conducted at the University of Wisconsin-Madison Thermal Stratification Test Facility

The Department of Energy (DOE)-Nuclear Energy University Programs (NEUP) supported the creation and operation of the Thermal Stratification Test Facility (TSTF) at the University of Wisconsin Madison (UWM) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. The TSTF was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements of the flow and temperature were collected for use as experimental benchmarks to support validation efforts for computational models. The results of these tests contribute to the greater understanding of thermal stratification behavior of liquid sodium under various configurations and operating conditions. The six TSTF tests selected for benchmarking are a set of forced circulation tests at a fixed flow rate with different Upper Internal Structure (UIS) configurations in the test section (no UIS, solid UIS, and a UIS with flow area of 4, 8, 12, and 100%). This report provides a complete description of the benchmark problems, including all key test facility details, descriptions of each test condition, and measured data for comparison with modeled results.

42 ENGINEERING↗
DOE OSTI2020

Power Generation Cycle with RELAP-7

Balance of plant of a High-Temperature gas-cooled Reactor – Pebble bed Module (HTR-PM) is modeled using RELAP-7. The model includes the primary helium loop where the heat produced by the core is transferred to a helical-coil steam generator and the secondary loop where super-heated steam is fed to a turbine to extract power. The steam is then condensed to subcooled water and pumped back into the steam generator. Results demonstrate that RELAP-7 has the capability to model different regimes encountered in two-phase flow with wall boiling, superheated steam, subcooled liquid and condensation. RELAP-7 also has the capability of modeling different fluids with their own specific models in a very flexible way, without any code change. This also demonstrates the capability of modeling the transfer of fission power into electrical power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Benchmarking MELCOR's NAC Package to ABCOVE Tests AB5 and AB6

This report presents analyses of the AB5 and AB6 ABCOVE sodium spray fire experiments with the MELCOR code. This code simulates the progression of accident events for analysis and auditing purposes of nuclear facilities during accident conditions. Historically, the ABCOVE experiments have contributed to the validation of aerosol physics and related phenomena. Given advancements in sodium-cooled reactor designs, characterization of the sodium spray combustion may further the review and validation of newly incorporated sodium properties and physics packages, namely, the sodium equations of state (EOS) and the sodium combustion (NAC) package within MELCOR. By analyzing the AB5 and AB6 experiments with and without the NAC package, sodium specificity for spray combustion and aerosol formation as well as speciation of the combustion products are reviewed with the new packages. This effort provides code users with a demonstration of the current code capabilities. This report provides the current best practices for the NAC package as well as a discussion of any issues observed while performing the presented analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Benchmarking MELCOR's NAC Package to ABCOVE Test AB7

This report presents analyses of the AB7 ABCOVE sodium spray fire experiment with the MELCOR code. This code simulates the progression of accident events for analysis and auditing purposes of nuclear facilities during accident conditions. Historically, the ABCOVE experiments have contributed to the validation of aerosol physics and related phenomena. Given advancements in sodium-cooled reactor designs, characterization of the sodium spray combustion may further the review and validation of newly incorporated sodium properties and physics packages, namely, the sodium equations of state (EOS) and the sodium combustion (NAC) package within MELCOR. Previously, the AB5 and AB6 experiments were analyzed with and without the NAC package. This work builds on the previous analyses with a demonstration of the current code capabilities of MELCOR with a more mild Na spray and pool fire scenario.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Estimation of Fission Product Transport Parameters for Cesium in the AGR-3/4 TRISO Fuel Experiment

A one-dimensional (1D) finite-element model of fission product transport in the AGR-3/4 experiment has been developed using the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and implemented in the fuel performance code, BISON. The model resolves capsule-specific geometries, materials, and temperature histories and simulates radial migration of fission products from the fuel compact through the inner ring, outer ring, and into the sink ring. Model parameters governing diffusion and sorption were estimated for key fission products – cesium (Cs), and europium (Eu) – by simultaneously fitting modeled isotopic concentration profiles and total ring inventories to a post-irradiation experimental measurement. These data include gamma scanning, liquid scintillation for Sr-90, radial deconsolidation leach-burn-leach analysis, tomographic reconstructions, and destructive physical sampling. A mortar-based interfacial sorption framework was implemented to enforce physically consistent mass transfer and flux conservation across gas gaps. Two classes of parameter sets were derived: a least-squares best-fit, and a safety-oriented conservative-fit, what applies strong penalties for underprediction of sink inventories. Across all twelve capsules, the model successfully reproduces the dominant radial transport trends for Cs, Sr, with decreasing concentrations from the compact outward through successive rings. Cs behavior is captured most consistently, while strontium predictions reveal systematic trade-offs between compact accuracy and conservative sink-ring bounding. The results demonstrate that sink ring weighted calibration provides conservative, safety relevant bounds on low temperature fission product transport, but at the cost of underpredicting compact inventories for Sr isotopes. These discrepancies highlight the need for additional physics, including fast-slow diffusion model, incorporating trapping mechanism in the transport behavior. Overall, this work establishes a robust, capsule-specific modeling framework for AGR-3/4 fission product transport and provides a defensible basis for parameter selection in source-term and fuel performance analyses for high temperature gas-cooled reactors.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Development of SAM Code Capabilities for Safety Analysis of GCR Air-ingress Events

Air-ingress following a depressurized loss-of-forced-cooling (DLOFC) event is a challenging, multiphysics safety scenario for High-Temperature Gas-Cooled Reactors (HTGRs), involving coupled gas composition transport, buoyancy-driven flow redistribution, graphite oxidation, and structural heat-up. Despite its importance — air ingress is a key scenario identified in the PIRT process for the HTGRs — existing system-level safety codes have lacked the integrated capability to simulate the complete event sequence with high confidence. This report documents the development, validation, and demonstration of three new capabilities in the SAM code to address this gap: (1) a multi-component gas mixture flow model with binary diffusion to track the helium-air composition and its effect on system density and flow; (2) a 0-D graphite oxidation model based on the Roes correlation, including oxygen consumption and exothermic heat release; and (3) an isentropic critical flow model for accurate representation of primary system depressurization through a break. These capabilities are validated against two benchmark experiments. The NSTF heavy-gas ingress experiment validates the multi-component flow model: SAM correctly reproduces the rapid buoyancydriven flow stagnation and subsequent natural circulation recovery driven by composition-dependent density changes. The NACOK graphite oxidation experiment validates the oxidation model: SAM predicts a bottom-level graphite weight loss of 25%, in close agreement with the measured 24%, and reproduces the strong axial nonuniformity and block-geometry dependence of oxidation, at a level comparable to the SPECTRA and TINTE codes. The validated capabilities are then exercised together in an integrated, reactor-scale simulation of a DLOFC air-ingress transient in a simplified HTR-PM pebble-bed reactor. In a single calculation spanning approximately 8 days, SAM reproduces the complete accident sequence: rapid depressurization, densityand diffusion-driven air ingress over ˜15 hours, onset of buoyancy-driven natural circulation, exothermic graphite oxidation with a peak fuel temperature at ˜62 hours, and eventual passive cooldown. These results demonstrate that SAM now provides the nuclear community with a preliminarily validated, modern systemlevel tool for HTGR air-ingress safety analysis, filling a recognized capability gap. Future extensions to broaden species tracking, improve oxidation chemistry, and refine the reactor model are discussed.

Yang, Gang↗
DOE OSTI2021

Resurgence of a Nation’s Radiation Science Driven by Its Nuclear Industry Needs

This article describes the radiation facilities and associated sample preparation, management, and analysis equipment currently in place at the Dalton Cumbrian Facility, a facility which opened in 2011 to support the UK’s nuclear industry. Examples of measurements performed using these facilities are presented to illustrate their versatility and the breadth of research they make possible. Results are presented from research which furthers our understanding of radiation damage to polymeric materials, radiolytic yield of gaseous products in situations relevant to nuclear materials, radiation chemistry in light water reactor cooling systems, material chemistry relevant to immobilization of nuclear waste, and radiation-induced corrosion of fuel cladding elements. Applications of radiation chemistry relevant to health care are also described. Research concerning the mechanisms of radioprotection by dietary carotenoids is reported. An ongoing open-labware project to develop a suite of modular sample handling components suited to radiation research is described, as is the development of a new neutron source able to provide directional beams of neutrons.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Rare Earth Carbide (Nd-C and Ce-C) Synthesis and Characterization to Inform Phase Equilibrium in Advanced Nuclear Fuels

As advances are being made regarding the performance of nuclear fuels, uranium carbides, and composites, such as (U,Zr)C and UO2 + UCx, have recently gained significant interest for deployment in nuclear space propulsion and high temperature gas-cooled reactors, respectively. However, the phase equilibria of several fission products in carbide systems remain unknown and may impact the overall fuel performance, specifically for particle nuclear fuels that are designed for commercial nuclear energy. Furthermore, comprehensive thermodynamic data on Rare Earth (RE) carbides, such as the Nd-C and Ce-C binary systems, remain limited. Presented in this study are the synthesis methods and characterizations of several Nd-C and Ce-C compositions. The findings from this research provide insights on the stability of RE-C binaries that form in irradiated nuclear fuels and address a critical knowledge gap in the current state of thermodynamics for two key RE-C systems.

Cavazos, Steven J. (ORCID:0009000130329363)↗
NASA NTRS1971

Shock and vibration tests of a SNAP-8 NaK pump

The pump used for reactor cooling in the SNAP 8 space power system was subjected to the expected vehicle launch vibration, and shock loading in accordance with the SNAP 8 environmental specification. Subsequent disassembly revealed damage to the thrust bearing pins, which should be redesigned and strengthened. The unit was operational, however, when run in a test loop after reassembly.

Stromquist, A. J.↗
NASA NTRS1971

After heat distribution of a mobile nuclear power plant

A computer program was developed to analyze the transient afterheat temperature and pressure response of a mobile gas-cooled reactor power plant following impact. The program considers (in addition to the standard modes of heat transfer) fission product decay and transport, metal-water reactions, core and shield melting and displacement, and pressure and containment vessel stress response. Analyses were performed for eight cases (both deformed and undeformed models) to verify operability of the program options. The results indicated that for a 350 psi (241 n/sq cm) initial internal pressure, the containment vessel can survive over 100,000 seconds following impact before creep rupture occurs. Recommendations were developed as to directions for redesign to extend containment vessel life.

Parker, W. G.↗
NASA NTRS1973

Exploratory study of several advanced nuclear-MHD power plant systems.

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

Williams, J. R.↗
NASA NTRS1975

Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

Peterman, D. D.↗
NASA NTRS1975

A preliminary systems-engineering study of an advanced nuclear-electrolytic hydrogen-production facility

An advanced nuclear-electrolytic hydrogen-production facility concept was synthesized at a conceptual level with the objective of minimizing estimated hydrogen-production costs. The concept is a closely-integrated, fully-dedicated (only hydrogen energy is produced) system whose components and subsystems are predicted on ''1985 technology.'' The principal components are: (1) a high-temperature gas-cooled reactor (HTGR) operating a helium-Brayton/ammonia-Rankine binary cycle with a helium reactor-core exit temperature of 980 C, (2) acyclic d-c generators, (3) high-pressure, high-current-density electrolyzers based on solid-polymer electrolyte technology. Based on an assumed 3,000 MWt HTGR the facility is capable of producing 8.7 million std cu m/day of hydrogen at pipeline conditions, 6,900 kPa. Coproduct oxygen is also available at pipeline conditions at one-half this volume. It has further been shown that the incorporation of advanced technology provides an overall efficiency of about 43 percent, as compared with 25 percent for a contemporary nuclear-electric plant powering close-coupled contemporary industrial electrolyzers.

Escher, W. J. D.↗