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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
NASA NTRS1972

Manned space flight nuclear system safety. Volume 4: Space shuttle nuclear system transportation. Part 1: Space shuttle nuclear safety

An analysis of the nuclear safety aspects (design and operational considerations) in the transport of nuclear payloads to and from earth orbit by the space shuttle is presented. Three representative nuclear payloads used in the study were: (1) the zirconium hydride reactor Brayton power module, (2) the large isotope Brayton power system and (3) small isotopic heat sources which can be a part of an upper stage or part of a logistics module. Reference data on the space shuttle and nuclear payloads are presented in an appendix. Safety oriented design and operational requirements were identified to integrate the nuclear payloads in the shuttle mission. Contingency situations were discussed and operations and design features were recommended to minimize the nuclear hazards. The study indicates the safety, design and operational advantages in the use of a nuclear payload transfer module. The transfer module can provide many of the safety related support functions (blast and fragmentation protection, environmental control, payload ejection) minimizing the direct impact on the shuttle.

Source record↗
NASA NTRS1973

Feasibility of an on-line fission-gas-leak detection system

Calculations were made to determine if a cladding failure could be detected in a 100-kW zirconium hydride reactor primary system by monitoring the highly radioactive NaK coolant for the presence of I-131. The system is to be completely sealed. A leak of 0.01 percent from a single fuel pin was postulated. The 0.364-MeV gamma of I-131 could be monitored on an almost continuous basis, while its presence could be varified by using a longer counting time for the 0.638-MeV gamma. A lithium-drifted germanium detector would eliminate radioactive corrosion product interference that could occur with a sodium iodide scintillation detector.

Lustig, P. H.↗
NASA NTRS1972

Preliminary design of reactor power systems for the manned space base.

The results of design integration studies of uranium-zirconium hydride (UZr-Hx) reactor power systems for the NASA space base study program are presented. The power conversion systems investigated include the Brayton cycle, the organic Rankine cycle, the SNAP-8 mercury Rankine cycle, and thermoelectric (PbTe). The proposed space base has a 10-year life and requires 100 kWe of power. Two 50-kWe power systems with a nominal replacement life of 5 years are utilized. Parametric design data such as life, weight, radiator area, reactor outlet-temperature, reactor thermal power, and power conversion system efficiency are presented and used for the design and integration of the system with the space base.

Mckhann, G. G.↗
NASA NTRS1993

U.S./CIS eye joint nuclear rocket venture

An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

Clark, John S.↗
NASA NTRS2016

A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

Gomez, C. F.↗
NASA NTRSDate not supplied

Moderator Considerations for Space Nuclear Power and Propulsion Systems

Nuclear reactors have the potential to provide high energy density to enable sustainable surface power and advanced propulsion methods needed for human exploration activities at the moon and mars. Current mission planning is surveying different reactor types for space power and propulsion application. Of these reactor types, the use of a moderator within the reactor can enable reduced enrichment, reduce overall fuel loadings, and minimize the critical size of the reactor compared to unmoderated reference systems. This proceeding summarizes some moderator materials identified for space reactor applications: zirconium hydride, yttrium hydride, beryllium, and beryllium oxide, and the unique design considerations inherent to surface power and nuclear thermal propulsion reactor designs. It was found that there are four key considerations during the moderator selection and design process: nuclear properties, thermophysical & mechanical properties, manufacture & readiness, and environmental compatibility. Surface power reactors can benefit from moderators which minimize overall system mass and are capable of surviving high temperature irradiation environments for years with little degradation. Nuclear thermal propulsion reactors can benefit from moderators which are capable of retaining structural integrity under multiple burns while being exposed to a wide temperature range (40 < T < 500+ K). Moderator materials which exhibit good stability under irradiation and high temperature operation, minimize fuel pitch, and are high readiness are desirable for near term implementation.

nuclear thermal propulsion↗
DOE OSTI2022

Update on NNL TSL Validation Efforts [Slides]

This presentation covers how a new sub-thermal neutron transmission capability at RPI LINAC is now in use to help validate TSL evaluations. Additionally, NNL has contributed multiple material evaluations to ENDF/B-VIII.1. Lastly, this presentation covers new methods to improve TSL physics and accuracy are being formulated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Flux Characterization of Washington State University’s Epithermal Neutron Beam Facility

The neutron flux of the epithermal neutron beamline facility at the Washington State University (WSU) Research reactor was characterized as part of a research effort between Pacific Northwest National Laboratory (PNNL) and WSU to better understand the effects that neutron interaction has on the release of hydrogen from zirconium hydride. This effort was to confirm that the spectrum was epithermal, and that the magnitude of the flux was adequate for the needs of the PNNL research efforts.

Zipperer, Travis J. [BATTELLE (PACIFIC NW LAB)]↗
DOE OSTI2025

COR-0011 Rev 6 MARVEL Project Code of Record

This Code of Record identifies the codes, standards, and procedures necessary to design, develop, construct, and startup the Microreactor Applications Research Validation and Evaluation (MARVEL) Project at the Materials and Fuel Complex (MFC) Transient Reactor Test (TREAT) Facility and the Idaho National Laboratory (INL). The MARVEL Project is an INL test microreactor funded by the United States Department of Energy (DOE) via the Microreactor Program (MRP). The goal of the project is to establish an operational nuclear applications test bed that can generate combined heat and power to enable integration and R&D with end-user technologies, as well as allow microreactor technologists to test next-generation control systems. The microreactor is a thermal reactor utilizing Uranium Zirconium Hydride (UZrH) fuel with review and authorization by the Department of Energy Idaho Operations Office (DOE-ID) for National Environmental Policy Act (NEPA) compliance, safety review, and supplemental readiness assessments for startup and operation. To enable rapid deployment, the MARVEL reactor will reside in the Transient Reactor Test (TREAT) Facility and utilize the existing operating Category B reactor facility, approved facility safety basis, operating crews, and recent re-start experience.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2021

Yttrium Hydride Post Irradiation Examination Plan

The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

PLN-6268 Yttrium Hydride Post Irradiation Examination Plan

The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Effect of neutron beamline irradiation on hydrogen equilibrium vapor pressure over hydrided Zircaloy-4

Here, the effect of neutron irradiation on the hydrogen equilibrium vapor pressure over hydrided samples of nickel-plated Zircaloy-4 was evaluated at the Washington State University Research Reactor’s epithermal beamline facility. To enable this study, a measurement capability was designed and assembled to obtain low pressure (<0.1 Pa) measurements over the temperature range of interest (i.e., ~300° - 400°C). Measurement results were in good agreement with tends extrapolated from previous data obtained at higher temperatures as well as recent thermodynamic data. No irradiation enhancement of vapor pressure was observed under the incident flux of epithermal neutrons over this temperature range.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Update on NNL TSL Evaluations and Validation [Slides]

12 new/revised TSL evaluations were contributed to ENDF/B-VIII.1. The thermal diffusion length $L$ is an integral property of a single material’s absorption and scattering cross sections (both differential and integral). No other neutron reactions or materials are involved. The MC21-calculated $L$ for water is consistent with the spread of experimental data and is sufficiently sensitive to different H-H 2 O TSL physics models to use the method as a TSL integral performance benchmark. Modern high-quality diffusion experiments at elevated $T$ would allow direct low-cost physics benchmarking of water TSLs when public elevated-$T$ critical benchmarks are limited. NNL is working with LLNL and RPI to develop experimental PNDA capability. An ICSBEP Fundamental Physics benchmark of the Nassar and Murphy PNDA experiment is being developed to provide an example of how to evaluate these type of benchmarks.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1972

Manned space flight nuclear system safety. Volume 1: Executive summary. Part 2: Space shuttle nuclear system safety

The nuclear safety integration and operational aspects of transporting nuclear payloads to and from an earth orbiting space base by space shuttle are discussed. The representative payloads considered were: (1) zirconium hydride-Brayton power module, (2) isotope-Brayton power module, and (3) small isotope power systems or heat sources. Areas of investigation also include nuclear safety related integration and packaging as well as operational requirements for the shuttle and payload systems for all phases of the mission.

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