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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Modeling Impact-Induced Reactivity Changes Using DAG-MCNP5

There is a long literature studying the criticality of space reactors immersed in water/sand after a launch accident; however, most of these studies evaluate nominal or uniformly compacted system configurations. There is less research on the reactivity consequences of impact, which can cause large structural deformation of reactor components that can result in changes in the reactivity of the system. Predicting these changes is an important component of launch safety analysis. This paper describes new features added to the DAG-MCNP5 neutronics code that allow the criticality analysis of deformed geometries. A CAD-based solid model of the reactor geometry is used to generate an initial mesh for a structural mechanics impact calculation using the PRONTO3D/PRESTO continuum mechanics codes. Boundary conditions and material specifications for the reactivity analysis are attached to the solid model that is then associated with the initial mesh representation. This geometry is then updated with the deformed finite element mesh to perturb node coordinates. DAG-MCNP5 was extended to accommodate two consequences of the large structural deformations: dead elements representing fracture, and small overlaps between adjacent volumes. The dead elements are removed during geometry initialization and adjustments are made to conserve mass. More challenging, small overlaps where adjacent mesh elements contact cause the geometric queries to become unreliable. A new point membership test was developed that is tolerant of self-intersecting volumes, and the particle tracking algorithm was adjusted to enable transport through small overlaps. These new features enable DAG-MCNP5 to perform particle transport and criticality eigenvalue calculations on both deformed mesh geometry and CAD geometry with small geometric defects. Detailed impact simulations were performed on an 85-pin space reactor model. In the most realistic model that included NaK coolant and water in the impact simulation, the eigenvalue was determined to increase 2.7% due to impact.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Evaluation of a Potassium Filled UO2 Core in a Beryllium- ??eflected Space Reactor

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965[1,2] have been evaluated for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and were determined to be acceptable benchmark experiments and are included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook)[3]. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.”[4] The third configuration in this series of experiments was a beryllium-reflected mockup of a stainless-steel-clad, highly enriched uranium (HEU)-O2 fuel, potassium-cooled, space power reactor. Reactivity measurements, cadmium ratio spectral measurements, and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for the International Reactor Physics Experiment Evaluation Project (IRPhEP) and are published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)[5]. More recently, two additional critical configurations have been evaluated and added to the existing evaluation report in the ICSBEP Handbook. The new critical configurations measure the effect of filling the core with potassium. The critical configuration consisted of 30.48 cm tall fuel tubes of 93.15 wt.% enriched uranium dioxide (UO2) pellets. Each fuel tube had 26 pellets with a total mass of 295.8 g UO2 per tube. There were 253 tubes were arranged with 1.506-cm center-to-center, triangular spacing and surrounded on all side by a beryllium reflected. The reactivity of the system was measured with the core tank empty and filled with potassium. The system reactivity was increased from 13.4 ¢ to 32 ¢. This change in system reactivity represents the worth of the potassium. Simple benchmark models for the empty and potassium filled cores have been derived and evaluated and found to be acceptable as benchmark experiments. The benchmark experiment eigenvalues are 1.0001 ± 0.0008 and 1.0011 ± 0.0009 for the empty and filled cores, respectively. When calculated with MCNP5 and ENDF/B-VII.0 the calculated eigenvalues are 0.98906 ± 0.00002 and 0.98883 ± 0.00002, respectively. These results are concerning due to the drop of the calculated system reactivity when potassium is added to the system rather than the expected increase in system reactivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Kilopower Space Reactor Launch Safety - Maximum Credible Fissions for a Criticality Accident

This note examines potential accident conditions for the Kilopower space reactor going critical during a launch accident. The goal of this analysis is to postulate the total number of fissions that could be generated during an accident event. The total number of fissions is an approximate surrogate for what will be the maximum release (or source term) for the accident. From the source term an approximation of dose (based upon the fission products generated and released) can be estimated. This note will only attempt to app roximate the total number of fissions to the right order of magnitude. Dose consequences from the number of fissions will be documented in a separate note.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Kilopower Space Reactor Launch Safety - Maximum Credible Dose for a Criticality Accident

This note examines potential consequences for the Kilopower small space reactor going critical during a launch accident. The goal of this analysis is to postulate the dose at a distance that could be generated during a criticality accident event (with emphasis on potential public doses.) This note does not examine local consequences to an individual who would be close enough to the reactor (within several 10’s of meters) to receive a lethal dose from the direct shine of neutrons and gamma radiation present when the reactor is critical.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Evaluation of a potassium filled UO{sub 2} core in a beryllium-reflected space reactor - 14371

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965[1,2] have been evaluated for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and were determined to be acceptable benchmark experiments and are included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook)[3]. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study 'power plants for the production of electrical power in space vehicles'.[4] The third configuration in this series of experiments was a beryllium-reflected mockup of a stainless-steel-clad, highly enriched uranium (HEU)-O{sub 2} fuel, potassium-cooled, space power reactor. Reactivity measurements, cadmium ratio spectral measurements, and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for the International Reactor Physics Experiment Evaluation Project (IRPhEP) and are published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)[5]. More recently, two additional critical configurations have been evaluated and added to the existing evaluation report in the ICSBEP Handbook. The new critical configurations measure the effect of filling the core with potassium. The critical configuration consisted of 30.48 cm tall fuel tubes of 93.15 wt.% enriched uranium dioxide (UO{sub 2}) pellets. Each fuel tube had 26 pellets with a total mass of 295.8 g UO{sub 2} per tube. There were 253 tubes were arranged with 1.506-cm center-to-center, triangular spacing and surrounded on all side by a beryllium reflected. The reactivity of the system was measured with the core tank empty and filled with potassium. The system reactivity was increased from 13.4 c to 32 c. This change in system reactivity represents the worth of the potassium. Simple benchmark models for the empty and potassium filled cores have been derived and evaluated and found to be acceptable as benchmark experiments. The benchmark experiment eigenvalues are 1.0001 ± 0.0008 and 1.0011 ± 0.0009 for the empty and filled cores, respectively. When calculated with MCNP5 and ENDF/B-VII.0 the calculated eigenvalues are 0.98906 ± 0.00002 and 0.98883 ± 0.00002, respectively. These results are concerning due to the drop of the calculated system reactivity when potassium is added to the system rather than the expected increase in system reactivity. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study "power plants for the production of electrical power in space vehicles." The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O 2 fuel mockup of a potassium-cooled space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO 2 ) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO 2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Godiva IV Simulated Radiation Field Characterization and Variance Reduction

Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1974

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Lantz, E.↗
DOE OSTI2013

Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies

The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O 2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O 2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %$\delta$k/k to 0.0004 %$\delta$k/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

KRUSTY: Advancing Space Nuclear Power for Future Missions [Slides]

For more than 70 years, LANL has conducted subcritical and critical experiments in support of space reactors, small modular reactors (SMRs), and criticality safety programs. These experiments were performed first at Los Alamos Critical Experiments Facility (LACEF) at TA-18 and in more recent years at the National Criticality Experiments Research Center (NCERC) in Nevada.

LACEF↗
DOE OSTI2012

Critical Configuration for Beryllium Reflected Assemblies of U(93.15)O 2 Fuel Rods (1.506-CM Pitch and 7-Tube Clusters)

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Critical configuration and physics measurements for graphite reflected assemblies of U(93.15)O 2 fuel rods (1.506-CM pitch)

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods.

1.506-cm pitch UO2 Assembly↗
DOE OSTI2020

Results of the KRUSTY Warm Critical Experiments

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a prototypic nuclear-powered test of a 5-kW(thermal) Kilopower space reactor. This paper presents results from the KRUSTY warm critical experiments, which were completed prior to the final system test. The first set of criticals comprised cold or zero-power criticals; i.e., the core was not heated by fission power. These were followed by three warm criticals, where fission power heated the core to 200°C, 300°C, and 450°C, respectively. These criticals provided the data, confidence, and regulatory framework that were needed to proceed with the KRUSTY nuclear system test. The criticals also provided valuable data for the benchmarking of codes applicable to all nuclear systems. Finally, a comparison of KRUSTY results to pretest predictions is provided, and overall, the models matched the experimental results very closely.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same set of highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium top reflector is evaluated in HEU-MET-FAST-069, and two additional highly enriched uranium annuli with beryllium cores are evaluated in HEU-MET-FAST-059. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast neutron spectra assemblies were determined to be acceptable benchmark experiments. The calculated eigenvalues for both the detailed and the simple benchmark models are within ~0.26 % of the benchmark values for Configuration 1 (calculations performed using MCNP6 with ENDF/B-VII.1 neutron cross section data), but under-calculate the benchmark values by ~7s because the uncertainty in the benchmark is very small: ~0.0004 (1s); for Configuration 2, the under-calculation is ~0.31 % and ~8s. Comparison of detailed and simple model calculations for the potassium worth measurement and potassium mass coefficient yield results approximately 70 – 80 % lower (~6s to 10s) than the benchmark values for the various nuclear data libraries utilized. Both the potassium worth and mass coefficient are also deemed to be acceptable benchmark experiment measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Direct current-self-sustained non-ambipolar plasma at low pressure

For decades, non-ambipolar diffusion has been observed and studied in laboratory plasmas that contain a double layer. However, self-sustained non-ambipolar plasma has yet to be demonstrated. This article reports the method and results for achieving such a condition at low pressure, with a wide power range (as low as 6 W). The findings reveal that to achieve self-sustained non-ambipolar plasma, both the balance between electron and ion heating and the space-potential gradient are critical. The plasma reactor developed in this work has potential applications that include microelectronic surface processing and space propulsion, via space-charge-neutral plasma-beam thruster, when operated in the high power regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Small CaH{sub 2} Moderated Thermal Reactor for Surface Power Generation II: Core Neutronics Design

National Aeronautics and Space Administration (NASA) is considering to adopt a small nuclear reactor for the planned Mars surface mission. The required specifications of the reactor system are: 10- to 100-kW electric power with power density above 5 We/kg; 8-year stable operation. In the proposed study, stretch goals for this surface power system were set as: electric power of 20 kWe with total system weight below 1000 kg. Furthermore, nuclear security and nuclear safety in the case of launch accidents must be considered for surface power systems, therefore, selecting low enriched uranium (LEU, = 20 wt% of {sup 235}U/U) is preferable due to its lower criticality. However, the reactor core weight tends to increase in the case of a reactor with LEU fuel, because relatively larger-mass core is required to achieve the critical. The essential feature of the space reactor core neutronics design is to balance between the nuclear security and safety, and the mass reduction of the core. Some feasible methods to reduce the core mass in the case of the LEU core are: (1) optimized neutron moderation, (2) reduce neutron leakage, and (3) increase multiplication of neutron. A solid moderator has been proposed by Kimura et.al.: neutron leakage control and neutron multiplicity will be featured in the current study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗