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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 1,099 records · Page 61
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 8-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both to be focused on obtaining fission product transport data to support improvements in modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These goals were combined in the AGR-3/4 irradiation experiment, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body, in which the outer graphite ring included a floor and cap to fully enclose the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to obtain data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compact 1-4 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a “fuel body”, in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2011

Alternative Fuel Research in Fischer-Tropsch Synthesis

NASA Glenn Research Center has recently constructed an Alternative Fuels Laboratory which is solely being used to perform Fischer-Tropsch (F-T) reactor studies, novel catalyst development and thermal stability experiments. Facility systems have demonstrated reliability and consistency for continuous and safe operations in Fischer-Tropsch synthesis. The purpose of this test facility is to conduct bench scale Fischer-Tropsch (F-T) catalyst screening experiments while focusing on reducing energy inputs, reducing CO2 emissions and increasing product yields within the F-T process. Fischer-Tropsch synthesis is considered a gas to liquid process which reacts syn-gas (a gaseous mixture of hydrogen and carbon monoxide), over the surface of a catalyst material which is then converted into liquids of various hydrocarbon chain length and product distributions1. These hydrocarbons can then be further processed into higher quality liquid fuels such as gasoline and diesel. The experiments performed in this laboratory will enable the investigation of F-T reaction kinetics to focus on newly formulated catalysts, improved process conditions and enhanced catalyst activation methods. Currently the facility has the capability of performing three simultaneous reactor screening tests, along with a fourth fixed-bed reactor used solely for cobalt catalyst activation.

Surgenor, Angela D.↗
DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Determination of Average Burnup in AGR-3/4 Compacts 1-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite. These two goals were combined in the AGR-3/4 irradiation, which consisted of 12 independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. Two capsule types were used: a standard capsule and a so-called fuel body, in which the outer graphite ring included a floor and cap that fully enclosed the fuel. The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from the matrix and graphite at accident temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

IER 519 Experiment Execution Report: TEX Hanford

The TEX-Hanford experiments were performed at the National Criticality Experiments Research Center (NCERC) at the Device Assembly Facility (DAF) at the Nevada Nuclear Security Site (NNSS). Three configurations were measured, with the measurements occurring in December 2025 and January 2026. The Fe-14 configuration was executed in December 2025; the Fe-11 configuration began its handstack on the Planet critical assembly machine in December 2025, but was taken critical in January 2026; the Fe-16 configuration was executed entirely in January 2026. The TEX-Hanford experiments were performed on the Planet critical assembly machine utilizing using PANN (Plutonium Aluminum No Nickel) ZPPR (Zero Power Physics Reactor) plutonium plates as fuel. The configurations were moderated with varying thicknesses of HDPE and iron absorber. This document details the configurations that were measured, preliminary reactivity measurements of the measured configurations, data files from the neutron detection systems, and results from the coordinate-measuring machine (CMM) measurements of the final stack height measurements each configuration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2015

In-Space Propulsion, Logistics Reduction, and Evaluation of Steam Reformer Kinetics: Problems and Prospects

Human space missions generate waste materials. A 70-kg crewmember creates a waste stream of 1 kg per day, and a four-person crew on a deep space habitat for a 400+ day mission would create over 1600 kg of waste. Converted into methane, the carbon could be used as a fuel for propulsion or power. The NASA Advanced Exploration Systems (AES) Logistics Reduction and Repurposing (LRR) project is investing in space resource utilization with an emphasis on repurposing logistics materials for useful purposes and has selected steam reforming among many different competitive processes as the preferred method for repurposing organic waste into methane. Already demonstrated at the relevant processing rate of 5.4 kg of waste per day, high temperature oxygenated steam consumes waste and produces carbon dioxide, carbon monoxide, and hydrogen which can then be converted into methane catalytically. However, the steam reforming process has not been studied in microgravity. Data are critically needed to understand the mechanisms that allow use of steam reforming in a reduced gravity environment. This paper reviews the relevant literature, identifies gravity-dependent mechanisms within the steam gasification process, and describes an innovative experiment to acquire the crucial kinetic information in a small-scale reactor specifically designed to operate within the requirements of a reduced gravity aircraft flight. The experiment will determine if the steam reformer process is mass-transport limited, and if so, what level of forced convection will be needed to obtain performance comparable to that in 1-g.

Logistics↗
DOE OSTI2023

SPC-70646 Specification for Reactor Supplemental Shielding for Use in DOME

NRIC is developing the Demonstration of Microreactor Experiments (DOME) test bed at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC): to allow for testing of advanced reactors in support of future licensing and commercial operations. One such need is a neutron and photon radiation shielding system to protect personnel and equipment from harsh neutron and photon fluxes during reactor operations and post shut down operations, disassembly, and decommissioning.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Method for Real-Time Salt Aerosol Concentration and Size Measurements for Molten Salt Reactor Safety Assessments

Licensing a new nuclear reactor involves evaluating credible accident scenarios by developing models that simulate accident progression to predict possible outcomes and their impact on safety. The formation of radionuclide-bearing aerosols in the respirable size range can significantly affect dose consequence and threaten human health and is therefore a focus of nuclear reactor safety assessments. Recent reviews of the literature identified a lack of existing experimental data describing the mechanisms of formation and properties (size, concentration, and composition) of radionuclide-bearing aerosols that may be produced from molten salt reactor (MSR) facilities during postulated accident scenarios. Experiments on aerosol formation from molten salt systems are a high priority need that will support MSR licensing by indicating the radiological consequences of aerosol formation and providing the data required for model development and validation. The evolving conditions that occur during a MSR accident, such as a spill of molten fuel salt, may affect aerosol formation mechanisms and aerosol properties over time. There is a need to conduct experiments that simulate credible MSR accidents in a laboratory to generate aerosols with realistic characteristics and behaviors and a need to measure these aerosols in real time under accident-relevant conditions. This report describes the development of a method that can be used to quantify the size and concentration of salt aerosol particles that form from molten salt systems in real time. This method will be employed in future integral effects tests that are conducted at an engineering scale to simulate realistic MSR accidents and in future separate effects tests that will provide mechanistic insight into aerosol formation and properties to support process model development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

INL Contributions to Draft HTTF Benchmark Specifications

The High-Temperature Test Facility (HTTF) is an integral effects thermal hydraulics test facility at Oregon State University designed as a 1/4 length scale model of the Modular High-Temperature Gas-Cooled Reactor 350 MW core (mHTGR-350). In the spring and summer of 2019, several experiments were conducted at HTTF providing a valuable source of gas-cooled reactor thermal hydraulics data. Idaho National Laboratory (INL), Oregon State University, Argonne National Laboratory, and Canadian Nuclear Laboratories have partnered to use this experimental data to develop a gas-cooled reactor thermal hydraulics benchmark led by INL under the auspices of the Advanced Reactor Technologies (ART) program. This report provides some context on the benchmark, HTTF, and previous Reactor Excursions and Leak Analysis Program (RELAP)5-3D modeling of HTTF. It also provides a draft of the benchmark specifications for the Depressurized Conduction Cooldown problem, which is the INL-led benchmark problem.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fuel Performance Evaluation of THOR-C Experiments

The Temperature Heatsink Overpower Response Commissioning (THOR-C) and THOR-Metal (THOR-M) experiments will be performed as part of an ongoing project for testing sodium fast reactor fuels with the Japan Atomic Energy Agency (JAEA). The THOR-C experiments consist of fresh metallic fuel pins and have been analyzed using the ABAQUS, Ansys codes and the BISON fuel performance code. THOR-M-Loss of Flow-1 (THOR-M-LOF-1) is designed to test an EBR-II irradiated fuel pin under LOF conditions. Simulation of the THOR-MLOF-1 experiment required first simulating the base irradiation of the fuel pin in EBR-II. MFUEL module of SAS4A/SASSYS-1 [1] is a physics-based metallic fuel performance model applicable to the normal operation, transient scenarios and fuel failure modeling including scenarios with bulk fuel melting. The model has been validated using EBR-II normal operation, separate effect transient tests as well as TREAT M-Series transient tests [2]. In this study, MFUEL models has been utilized together with a new capsule heat transfer model developed in this project. The new heat transfer model was necessary due to (1) significant amount of heat losses that required 2D heat transfer, (2) the presence of a titanium heat sink, rejecting a significant amount of heat, and (3) stagnant coolant conditions, which are inconsistent with SAS4A/SASSYS-1 (SAS) heat transfer model. Updates to SAS4A/SASSYS-1 and MFUEL has been described below, followed by a preliminary validation effort using the results from THOR-C-2 fresh fuel capsule experiment. A previous study for THOR-C-2 analysis using BISON code is also utilized in this study to model this test [3]. [1] D. O’Grady, A. J. Brunett, L. Ibarra, A. Karahan, T. Kim, T. S. Sumner, R. Thomas, T. H. Fanning, “The SAS4A/SASSYS-2 Version 5.7 Safety Analysis Code System,” Argonne National Laboratory,ANL/NSE-SAS/5.7, (2023). [2] A. Karahan, T. Kim, T. Fanning, D. O’Grady, “Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1,” Argonne National Laboratory, ANL/NSE-23/11, (2023). [3] M. Mihelish, A. Zabriskie, K. Paaren, P. Medvedev, C. Jensen, “Fuel Performance Predictions for the TREAT THOR-C Experiments,” Idaho National Laboratory, INL/RPT-23-73397, Revision 0, (2023)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Test of nuclear decay rate variation due to an antineutrino flux

Unexplained variations of the decay rate parameter for weak interaction decays such as β ± decay, electron capture, as well as strong interaction α decay have been reported. Because these variations have been presented by a number of groups, at various locations, using various types of detectors, different isotopes, and over extended periods of time, some researchers have interpreted the source of these variations as not from ambient environmental factors such as temperature, pressure, and humidity but via an unexplained fundamental interaction. The state of decay rate parameter variations experiments is reviewed, and the reported results are placed into a common comparable context by defining a phenomenological cross section. After that we make decay parameter measurements as a function of time at the level of 10 -5 in the presences of an antineutrino flux with an on-off cycle time of ≈ 30 days. This level of precision requires a detailed understanding of both systematic and statistical errors; otherwise, systematic errors in the measurement may mimic the decay events of fundamental interactions. The experiment searched for variation of the $^{54}_{25}$Mn, e - capture decay rate parameter and $^{137}_{55}$Cs, β - decay rate parameter both to a level of precision of 1 part in ≈10 5 by comparing the difference between the decay rate in the presence of an antineutrino flux ≈ 3 x 10 12 $\bar{ν}$ cm -2 s -1 and no flux measurements. The experiment was located 6.53 m from the reactor core of the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Two weak interaction decays, one via electron capture and the other via β - decay, were selected because the final state and the time reverse state each contain a neutrino and antineutrino, covering arguments that the antineutrino flux may interact differently or not at all in one of the cases. The $\gamma$ spectra from both decays were collected and analyzed independently. The measured variation in the decay rate parameters are found to be δλ/λ =(0.034 ± 1.38) x10 -5 for $^{54}_{25}$Mn and δλ/λ =(0.67±1.56) x10 -5 for $^{137}_{55}$Cs. These results are consistent with no measurable decay rate parameter variation due to an antineutrino flux, yielding a 68% confidence level upper limit sensitivity for $^{54}_{25}$Mn δλ/λ ≤ 1.31 x 10 -5 or σ ≤ 1.29 x10 -25 cm 2 in cross section and for $^{137}_{55}$Cs δλ/λ ≤ 2.23 x 10 -5 or σ ≤ 5.69 x10 -27 cm 2 . The cross-section upper limit obtained in these null or no observable effect measurements are ≈ 10 4 times more sensitive than past experiments reporting positive results in 54 Mn and 137 Cs. . .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

NRC Multiphysics Analysis Capability Deployment (FY2021--Part 1)

This report details progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” The tasks completed for this report are as follows: First, Task 1d: The net radiation transfer method was implemented into MOOSE for modeling reactor cavity cooling system geometries. RCCS models for two experiments were created: (1) Natural Convection Shutdown Heat Removal Test Facility (NSTF) experiment R022, and (2) HTTR VCS mockup. For validation, computed temperature distributions were compared to measured temperatures. Next, Task 4c: An algorithm for computing the pebble bed reactor equilibrium core isotopic com-position was developed and an initial version is implemented into the reactor multi-physics code Griffin. Initial results for a simplified axisymmetric pebble bed reactor are presented. Finally, Task 7: generation of a reference plant model for molten salt cooled pebble bed reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2003

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Godfroy, Thomas↗
DOE OSTI2024

ZPR/ZPPR Critical Experiment Facilities and As-Built ZPR/ZPPR Models

Argonne National Laboratory (ANL) operated four split-table critical experiment facilities, ZPR 3, ZPR-6, ZPR-9 and ZPPR, between 1955 and 1990. ZPR is the abbreviation for Zero Power Reactor, and ZPPR was the abbreviation for Zero Power Plutonium Reactor which was changed to Zero Power Physics Reactor in the 1980s due to negative views on Plutonium breeding and proliferation concerns. ZPR-3, ZPR-6, ZPR-9 and ZPPR were four reactors in a series of zero power (or lower power reactors) designed at Argonne in the 1950s and 1960s.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1987

Processing materials in space - The history and the future

The development of materials processing in space, and some of the Soyuz, Apollo, Skylab, and Shuttle orbital materials experiments are reviewed. Consideration is given to protein crystal growth, electrophoresis, low-gravity isoelectric focusing, phase partitioning, a monodisperse latex reactor, semiconductor crystal growth, solution crystal growth, the triglycine sulfate experiment, vapor crystal growth experiments, the mercuric iodide experiment, electronic and electrooptical materials, organic thin films and crystalline solids, deep undercooling of metals and alloys, magnetic materials, immiscible materials, metal solidification research, reluctant glass-forming materials, and containerless glass formation. The space processing apparatuses and ground facilities, for materials processing are described. Future facilities for commercial research, development, and manufacturing in space are proposed.

Chassay, Roger↗