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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 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Coarse-Grained-Particle Method for Simulation of Liquid–Solids Reacting Flows

A coarse-grained-particle method (CGPM) that tracks fewer particles and their collisions by lumping several real particles into computational parcels is developed. We demonstrate that the accuracy of CGPM can be improved by properly selecting interpolation schemes used to compute flow variables at particle and computational grid locations. The computational speed of CGPM was shown to be similar to that of a widely used and less accurate particle in cell method (PIC). Finally, a chemical reaction mechanism based on rare earth elements (REE) leaching from coal byproducts was implemented with CGPM to study the effects of several flow and design parameters on the leaching process. A counter-current reactor was studied and optimized to maximize the mass fraction of REE in the liquid solution. As a result, the CGPM can now be employed to cheaply and accurately solve industrial-scale problems containing millions of computational parcels.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Strictly monolayer large continuous MoS{sub 2} films on diverse substrates and their luminescence properties

Despite a tremendous interest on molybdenum disulfide as a thinnest direct band gap semiconductor, single step synthesis of a large area purely monolayer MoS{sub 2} film has not yet been reported. Here, we report a CVD route to synthesize a continuous film of strictly monolayer MoS{sub 2} covering an area as large as a few cm{sup 2} on a variety of different substrates without using any seeding material or any elaborate pretreatment of the substrate. This is achieved by allowing the growth to take place in the naturally formed gap between a piece of SiO{sub 2} coated Si wafer and the substrate, when the latter is placed on top of the former inside a CVD reactor. We propose a qualitative model to explain why the MoS{sub 2} films are always strictly monolayer in this method. The photoluminescence study of these monolayers shows the characteristic excitonic and trionic features associated with monolayer MoS{sub 2}. In addition, a broad defect related luminescence band appears at ∼1.7 eV. As temperature decreases, the intensity of this broad feature increases, while the band edge luminescence reduces.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Evaluation of Thermal Neutron Scattering Cross Sections for CaH 2

Solid metal hydrides have long been considered viable moderators for nuclear reactor designs due to their moderating ratios, hydrogen densities, high dissociation temperatures, and mechanical properties. CaH 2 is an orthorhombic saline-hydride that has recently been investigated and shows promise for use as a moderator in microreactors. At present, there is no Thermal Scatting Law (TSL) evaluation for CaH 2 in the ENDF/B-VIII.0 database. An evaluation does exist in the JEFF-3.3 database, performed by Serot, however it was limited by the implemented methods and thus both drastically over-predicts the incoherent elastic contribution and completely ignores the coherent elastic contribution to scattering from the metal ions. The thermal neutron scattering cross sections of CaH 2 are evaluated in this work to establish accurate data for use in reactor design. The present evaluation corrects the inaccuracies of the JEFF-3.3 data, and yields three distinct libraries: Ca in CaH 2 , H 1 in CaH 2 , and H 2 in CaH 2 .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Luminescent silica microagglomerates, synthesis, and environmental testing

Abstract Tracking mass through harsh environments requires surrogate particles that withstand the event and endure until sampling. Silica-covered quantum dots have been shown to withstand a range of environmental pHs from months to years; in this work they are shown to endure in anticipated local environments. Two methods of particle synthesis were employed to produce luminescent silica with particle diameters 0.1–4 μm. These tracer particles scale for mass production, tolerate harsh environments, and endure in debris. They could be deployed in places such as chemical explosions, industrial processes, geologic test beds, oil and gas fields, nuclear reactors, and geothermal plants to track mass under harsh conditions. Graphical abstract

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

A System-Level Model for Estimating Residual Strain and Life of Nuclear Reactor Coolant System Components Under Connected-System-Thermal–Mechanical Boundary Conditions

Background: Environmental-assisted fatigue (EAF) is a major issue for the long-term survival of nuclear power plant fleets in the U.S. and worldwide. Multi-material welded regions (e.g., nozzles) and other high-stress regions of reactor coolant system (RCS) components are prone to EAF-related damage. Objective: The discussed work describes a system-level finite element (FE) model of RCS components of a pressurized water reactor (PWR). This is with the goal of predicting the stress hotspots, strain residuals, strain amplitudes and the resulting fatigue lives. Methods: Herein, the FE model was developed considering system-level loading conditions (under connected system thermal-mechanical boundary conditions). Thermal-mechanical stress analysis was performed considering thermal stratification and a design-basis reactor loading cycle. Based on the FE model results, the strain residuals, strain amplitudes and resulting fatigue lives of RCS components were predicted. Results: The results show that some of the RCS components can have significantly different strain amplitudes, residual strain, and fatigue lives, despite having similar geometry and material. Higher residual strain can lead to accelerated cyclic hardening of material and the associated effect of EAF. The simulated component-level strain profile (under realistic multi-axial-multi-physics loading cycle) can guide the selection of appropriate test inputs for conducting laboratory-scale EAF tests, which is a focus of future works. Conclusions: Despite similar geometry and material the RCS component can have significantly different strain profiles and resulting fatigue lives. The discussed approach can help to identify and prioritize the RCS components for conducting expensive nondestructive evaluation (NDE) inspections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Improving Thermal Model Prediction Through Statistical Analysis of Irradiation and Post-Irradiation Data from AGR Experiments

As part of the High Temperature Reactors (HTR) R&D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Babcock & Wilcox Critical Experiment Calculations to Support MPACT Verification

This report describes a series of benchmark calculations performed using the method of characteristics radiation transport code MPACT to verify its radial transport and axial buckling calculation capabilities. The benchmarks are based on a series of Babcock & Wilcox critical assembly configurations analyzed in what are commonly known as the B&W-1484 benchmark experiment and the B&W-1810 benchmark experiment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

MOOSE Optimization Module Overview: Application to Residual Stress Inversion in Nuclear Fuel Plates

Manufacturing methods, material selections and unique plate geometries used in low enriched uranium plate fuels for the High Performance Research Reactor (HPRR) have led to concerns about the residual stresses effect on bond strength of the fuel/cladding interfaces and overall fuel performance under irradiation. The current HPRR plate fuel test program performed residual stress calculations using the contour method with experimental displacement data obtained by incrementally cutting the plate fuel. In this work, we use a gradient based inverse optimization module developed within the MOOSE framework to estimate normal and shear residual stresses within the plate fuels using displacement data taken from a cut placed in the HPRR fuel plate. These results are verified against residual stress calculations obtained from the contour method. The inversion protocol used in this work utilizes two meshes to separately discretize the physics and parameter space. This allows a fine mesh to be used to fully resolve the physics and geometry of the forward problem. A coarser mesh is used to resolve the parameter space with spatially varying mesh density determined from a sensitivity analysis. The adaptive resolution of the parameter space essentially has the effect of regularization as it is designed to achieve the bias-variance tradeoff.

42 ENGINEERING↗
DOE OSTI2024

Photonuclear Physics in SCALE

Radiation shielding modeling involves a wide range of applications, such as power reactors, critical experiments, portable sources, linear accelerators, and fusion systems. Particle energy ranges vary from thermal energies to fast energy spectra. MAVRIC, as the radiation shielding sequence of the SCALE code system [1] developed and maintained by Oak Ridge National Laboratory, is capable of modeling both neutron and photon interactions in matter and is widely used by the US Nuclear Regulatory Commission, Department of Energy, the National Nuclear Security Administration, and radiation shielding practitioners all around the world. Although most applications don’t involve highly energetic photons that can cause photonuclear reactions and create coupled chains of high energy, ionized electrons and bremsstrahlung radiation within a material medium, when present, these reactions can have a significant impact upon the downstream particle fluxes from incident photons above 20 MeV. For example, linear accelerators used for medical isotope production are examples of applications in which high-energetic photons are generated and used for driving or controlling other components of a complex facility. Radiation shielding analyses of such a facility would require accounting for any and all significant sources of radiation. Unfortunately, MAVRIC included no data and methods to model either photonuclear reactions or the bremsstrahlung photons until this work, which enabled accounting for both. Photonuclear reaction cross sections were imported from the Evaluated Nuclear Data File libraries, and corresponding methods were implemented in the SCALE code system; see Section 2 for how the nuclear data are imported and used. Bremsstrahlung photons were modeled by assuming thick-target approximation whereby all generated electrons and positrons due to photon interactions are assumed to be absorbed within the same material in which they were born, yielding a quick particle transport simulation method by using fast lookup tables for generating subsequent gammas; see Section 4 for details. A suite of verification models was used to test implemented methods and associated data. Tests yielded satisfactory results for the new features. Enhancements will be released as part of the next major SCALE release (v7.0), which is anticipated to be deployed in 2025, with beta releases available in the interim.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Molecular dynamics simulation of thermal transport in UO 2 containing uranium, oxygen, and fission-product defects

Uranium dioxide (UO 2 ) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes. The model is validated by comparison to low-temperature experimental measurements on single-crystal hyperstoichiometric UO 2+x samples and high-temperature literature data. Furthermore, this work will enable more accurate fuel-performance simulations and will extend to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Modeling MTS pyrolysis and SiC deposition kinetics using principal component analysis and neural networks

Accurate chemical kinetics modeling is crucial for improving the efficiency of chemical processing and synthesis of ceramic matrix composites. Detailed kinetic models are computationally expensive due to the large number of transported chemical species, while the simplified physics-based models, such as single-step global mechanisms, are efficient but often overlook key chemical intermediates and pathways. Recent deep learning approaches promise accurate and cost-effective models. Yet, they require additional closures for the transported nonlinear latent variables, complicating integration with existing solvers. In this work, we develop a hybrid linear—nonlinear reduced model for silicon carbide deposition from methyltrichlorosilane precursor by combining principal component analysis (PCA) and autoencoder (AE) neural network (NN) approaches. PCA is used to identify a smaller set of linear transport variables, enabling direct reuse of conventional transport solvers. NNs then reconstruct the full chemical state from these reduced variables. We demonstrate the method on a chemical vapor deposition reactor—comprising a gas-phase pyrolysis plug flow reactor and a heterogeneous surface reactor—over a wide range of temperatures, pressures, and residence times. Our PCA–AE model achieves high accuracy with only five transported scalars, achieving an eightfold cost reduction compared to detailed mechanisms, in both a priori (using data from the test set only) and a posteriori (coupled with a differential equation solver). In conclusion, notable errors arise primarily near training domain boundaries and for long residence times, indicating the need for domain shift indicators and better long-horizon predictions in future reduced chemistry model development.

autoencoder neural networks↗
DOE OSTI2014

Characterization of tungsten films and their hydrogen permeability

Prediction of tritium migration and its retention within fusion reactors is uncertain due to a significant role of the structural disorder that is formed on the surface layer after plasma exposure. Tungsten films deposited by any of the suitable methods are always disordered and contain a high density of hydrogen traps. Experiments on such films with hydrogen isotopes present a suitable complementary method, which improves the picture of the hydrogen interaction with fusion relevant materials. The authors report on the morphology, composition, and structure of tungsten films deposited by the thermionic vacuum arc method on highly permeable Eurofer substrates. Subsequently, hydrogen permeation studies through these films were carried out in a wide pressure range from 20 to 1000 mbars at 400 °C. The final value of the permeation coefficient for four samples after 24 h at 400 °C was between P = 3.2 × 10{sup −14} mol H{sub 2}/(m s Pa{sup 0.5}) and P = 1.1 × 10{sup −15} mol H{sub 2}/(m s Pa{sup 0.5}). From the time evolution of the permeation flux, it was shown that diffusivity was responsible for the difference in the steady fluxes, as solubility was roughly the same. This is confirmed by XRD data taken on these samples.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Deadtime Determination of Helium Detector Using MCNP Code

In detector systems, the minimum time separation between the events is called detector's deadtime. The The intrinsic properties of the detector and the pulse processing circuitry's characteristics are the sources of deadtime. Researchers have been working on improving a detector deadtime model that can implicitly characterize a detection system's behavior while reducing counting errors. There are two commonly known dead-time models: the 'Paralyzing' and the 'Non-paralyzing' models. In reality, detection systems fit neither of these idealized models perfectly, instead falling somewhere between the two models. The paralyzing model is mathematically expressed by eq. 1, where m is the measured count rate, n is the true count rate and t is dead-time. This model assumes that each event during the dead-time will reset it to a fixed duration, thus extending the deadtime. The dead-time extension depends on the count rate. According to the non-paralyzing model, deadtime is fixed after each detected event, and all events occurring during deadtime are lost. The fraction of time during which an apparatus is sensitive is 1-mt. Therefore, the fraction of the true number of events can be recorded. In this study, a Canberra {sup 3}He neutron detector (0.5NH1/1K) deadtime value was estimated using a combination of experimental subcritical assembly measurements and theoretical MCNP simulations. The deadtime of the Canberra 10 mm diameter {sup 3}He tube detector was calculated by comparing measured counts from different locations in the subcritical assembly at Missouri University of Science and Technology Reactor (MSTR) with MCNP (Monte Carlo N-Particle Code) simulations. The deadtime calculations are provided for both the paralyzing and non-paralyzing models for the Canberra 10 mm diameter {sup 3}He tube detector. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Electrolytic Oxide Reduction of Plutonium Oxide Surrogates

Our research team has investigated electrolytic oxide reduction (EOR) as an improved metal production process versus the currently employed multiple cycle direct oxide reduction (MCDOR) process for the conversion of plutonium oxide to plutonium metal. Inherent advantages in safety, waste generation, efficiency, and operational ease make EOR an attractive alternative to traditional metal production practices. Phase I of the project consisted of the construction of a prototype electrochemical cell, and proof of concept experiments towards the reduction of SnO 2 and CeO 2 compounds. Phase II of the project was focused on the optimization of the cerium system in preparation for future plutonium EOR experiments. We observed the electrolytic reduction of cerium oxide resulting in mixed degrees of success, and discovered this was a result of parasitic side reactions and partial reduction. We employed cyclic voltammetry (CV) to identify the products of these side reactions and characterize partially reduced species. We also observed that, when metal was successfully produced by EOR, a large amount of passed charge was required. We concluded that this observation was due to several inefficiencies of the process relating to the reduction mechanism (e.g. inefficient use of cathodically produced calcium). We used our CV results to guide our optimization of the EOR of CeO 2 by refining our electrochemical method, cell geometry, and materials of construction. Finally, we developed a next generation cell, worked to establish a new molten salt electrochemical capability in the Plutonium Facility, and completed a strategy for plutonium experiments. Our electrochemical method to produce plutonium metal clearly supports the core mission of the laboratory. Furthermore, this project has allowed for the development of capabilities that will provide valuable data to programs interested in characterization and detection, as well as science of signatures. Additionally, establishing this capability in the plutonium facility will facilitate future safeguard studies that can assist with safeguard development in molten salt reprocessing flowsheets, and molten salt reactor designs.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Exploring Acceleration Schemes for Nonlinear Heat Equations in Fusion Reactor Modeling

Plasma modeling inside a tokamak reactor poses a challenging and often times computationally expensive endeavor. The process of running these simulations has the potential to increase the predictive capabilities of these models. Here we investigate the effect particular acceleration schemes, mainly Anderson Acceleration, have on a nonlinear transport equation solver, Tango, developed by Parker et al. We show that Anderson acceleration offers a method for decreasing the computational cost through a test problem developed by Shestakov et al, which shows promise towards utilizing Anderson Acceleration in complex transport simulations

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibration curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Evaluation of Carbon Dioxide Capture From Existing Coal Fired Plants by Hybrid Sorption Using Solid Sorbents

A novel hybrid solid sorbent technology for CO₂ capture and separation from coal combustion-derived flue gas was evaluated. The technology – Capture of CO₂ by Hybrid Sorption (CACHYS™) – is a solid sorbent technology based on the following ideas: 1) reduction of energy for sorbent regeneration, 2) utilization of novel process chemistry, 3) contactor conditions that minimize sorbent-CO₂ heat of reaction and promote fast CO₂ capture, and 4) low-cost method of heat management. This report provides key information developed during the course of the project that includes sorbent performance, energy for sorbent regeneration, physical properties of the sorbent, the integration of process components, sizing of equipment, and overall capital and operational cost of the integrated CACHYS™ system. Seven sorbent formulations were prepared and evaluated at the lab-scale for energy requirements and CO₂ capture performance. Sorbent heat of regeneration ranged from 30-80 kJ/mol CO₂ and was found to be dependent on process conditions. Two sorbent formulations (designated HCK-4 & HCK-7) were down-selected for additional fixed-bed testing. Additional testing involved subjecting the sorbents to 100 continuous cycles in the fixed-bed reactor to determine performance as a function of time. The working capacity achieved for HCK-4 sorbent ranged from 5.5-8.0 g CO₂/100 g sorbent, while the HCK-7 typically ranged from 8.0-10.0 g CO₂/100 g sorbent. Overall, there was no deterioration in capacity with continuous cycling for either sorbent. The CACHYS™ bench-scale testing system designed and fabricated under this award consists of a dual circulating fluidized-bed adsorber and a moving-bed regenerator. The system takes a flue gas slipstream from the University of North Dakota’s coal-fired steam plant. Prior to being sent to the adsorber, the flue gas is scrubbed to remove SO₂ and particulate. During parametric testing of the adsorber, CO₂ capture achieved using the 2-bed configuration with recirculation in both beds was 65-70% with a high flue gas CO₂ loading (~7%) and up to 85% with a low flue gas CO₂ loading (~4%). A sorbent regenerator system consisting of a pre-heater, desorber, and cooler is used to heat the CO₂-rich sorbent with direct and indirect steam producing a nearly 100% pure stream of CO₂. Parametric testing of the regenerator system demonstrated the impact of process conditions on both desorption rate and the heat of regeneration. Clear evidence of the use of specific process conditions that lower the overall energy of desorption was identified. This observation validates measurements made at the laboratory-scale. Several longer-term continuous tests were conducted to evaluate the performance of the sorbent/process as a function of time. Using a 2-bed configuration, sustained capture efficiency of 40-60% with a high flue gas CO₂ loading (~8%) and 70-80% with a low flue gas CO₂ loading (~4%) were achieved. However, sorbent working capacity was found to be considerably lower than laboratory-scale measurements. The low working capacity is attributed to insufficient sorbent/gas contact time in the adsorber. Sorbent properties that had a significant impact on CO₂ capture performance were identified. The results show that controlling these sorbent properties substantially improves CO₂ capture performance, with preliminary estimates indicating that relative improvement of ~30% is possible. Testing culminated with an operationally trouble-free test of 15 hours with sustainable performance. Overall, several practical strategies to increase performance of the sorbent and process were identified. The initial technical and economic assessment of the CACHYS™ process estimated the cost of CO2 capture was $36.19/ton with a 48.6% increase in levelized cost of electricity (LCOE) for the 550 MWe net plant. Using additional data gathered over the course of the project, and with revised technical and economic assumptions, the estimated cost of CO₂ capture with the CACHYS™ process is $39/ton (only includes the cost of the CO2 capture system) with an increase in LCOE of 55.9%. Overall, CACHYS™ represents a significant improvement over the benchmark MEA system, and has demonstrated progress towards achieving DOE’s goals for CO₂ capture technologies.

01 COAL, LIGNITE, AND PEAT↗
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↗