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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
NASA NTRS1973

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

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

Williams, J. R.↗
DOE OSTI2018

Development of a Two-Fluid Molten Salt Reactor Safeguards Model

This work outlines the development of a two-fluid molten salt reactor process and safeguards model. The model is split into two parts consisting of a process model and a safeguards model. The process model is based on a design by Flibe Energy, the Liquid-Fluoride Thorium Reactor, which is a two-fluid molten salt reactor that performs full salt processing on-site to remove fission products and re-fuel the reactor. The model simulates feed and consumption rates of the reactor fuel and blanket salts. The process model includes the reactor core and salt processing loops. The reactor core model has a robust architecture that allows for integration with other tools and data sets as they become available. A majority of the effort to date has been focused on the process model, and the safeguards model will be developed in detail in future work. A preliminary safeguards analysis was performed based on actinide inventories, and a preliminary materials accountancy approach was initialized. The results of this analysis are presented along with a description of the model development.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Interactive Visualization of Multi-Group Cross Sections on High-Fidelity Spatial Meshes

Core simulation tools utilizing Monte Carlo transport methods represent the 'Holy Grail' for nuclear reactor analysis. However, deterministic transport methods pose a number of advantages to Monte Carlo from a computational perspective, which seems likely to make these methods more viable in the near to intermediate term. Deterministic methods almost universally discretize the energy domain in a few to hundreds of energy groups. One of the key challenges for deterministic transport-based methods to be viable is accurate, reactor agnostic multi-group cross section generation. This paper describes tools for data processing and visualization of multigroup cross sections generated on high-fidelity spatial meshes from the OpenMC Monte Carlo code. Next generation deterministic transport reactor core modeling tools will require multi-group cross section (MGXS) generation on high-fidelity spatial meshes. High-fidelity MGXS are necessary for the accurate simulation of neutron physics in nuclear reactors with a high degree of spectral heterogeneity - e.g., large localized neutron flux gradients throughout a reactor core. Although many local (e.g., pin cell) approximations have been developed to model the scalar flux to compute spectral-averaged MGXS, these techniques cannot account for complex intra-pin and intra-assembly spatial self-shielding effects. Monte Carlo presents the most accurate method for reactor agnostic multi-group cross section generation since it does not require the use of local approximations to the flux. Monte Carlo has been proposed as a highly desirable framework for reactor agnostic cross section generation for many years. Monte Carlo methods can effectively embed all of the relevant spectral effects into the MGXS through the random particle tracking and interaction sampling processes. Previous work has applied MCNP and Serpent to compute few-group constants for diffusion calculations, and OpenMC to compute multi-group cross sections for transport-based methods. The primary difficulty with using Monte Carlo for MGXS generation is the tremendous computational resources required to adequately converge multi-group cross sections in each spatial region and energy group. Recent development of the OpenMC Monte Carlo code for high-performance computing platforms has made multigroup cross section generation from Monte Carlo more tractable in the foreseeable future. The recent release of OpenMC v0.7.1 includes a new Python-based module for MGXS generation. The MGXS generation module leverages many of the algorithms and features for 'big data' analysis of large tally datasets on complex spatial meshes. This paper describes the data management workflow and visualization tools which have been developed to explore the plethora of MGXS data generated with OpenMC. It is the authors' hope that the development of the visualization tool described in this paper will ultimately enable the design of machine learning algorithms to accelerate the convergence rate of high-fidelity multi-group cross section generated with Monte Carlo. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1974

Physics and potentials of fissioning plasmas for space power and propulsion

A brief description of two gas core reactor concepts devised for propulsion applications is presented and recent research on the emission of radiation from nonequilibrium fissioning gases is considered. The principles of UF6 and plasma core reactor experiments are discussed. It is expected that the program of nuclear gas core reactor experiments can lead in several years to reactor operation at pressures, temperatures, and power levels of technological significance to terrestrial applications.

Thom, K.↗
DOE OSTI2018

Evaluation of Secondary Gamma Environments at the Annular Core Research Reactor

An overview of experimental and computational studies of prompt secondary gamma production and transport, executed under the auspices of the Readiness in Technical Base and Facilities (RTBF) program, is presented. Relevant experiments at the Annular Core Research Reactor (ACRR) were conducted in the FY2012 -- FY2014 timeframe and pertain to the performance of various elemental calorimeters and the analytic fractionation of dose contributions to the calorimeter discs. In particular, the influence of the choice of prompt capture gamma production databases on the computed disc heating factors is discussed. Finally, the results of a polyurethane foam moderation experiment are detailed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Direct Numerical Simulation of the Turbulent Flow Through PWR Spacer Grid and Mixing Vanes

Direct numerical simulation (DNS) has been regarded as a reliable data source for the development and validation of turbulence models along with experiments. The realization of DNS usually involves a very fine mesh that should be able to resolve all relevant turbulence scales down to Kolmogorov scale. As the most computationally expensive approach compared to other CFD techniques, DNS applications used to be limited to flow studies at very low Reynolds numbers. Thanks to the tremendous growth of computing power over the past decades, the simulation capability of DNS has now started overlapping with some of the most challenging engineering problems. One of those examples in nuclear engineering is the turbulent coolant flow inside reactor cores. In Light Water Reactor (LWR) cores, the nuclear fuel rods are arranged together as rod bundles in a triangular or square pattern. The fuel rod bundles are supported by the spacer grids. The mixing vanes are also installed to promote the turbulence intensity in coolant flow. An understanding of spacer grid and mixing vanes effect is important because it has a direct impact on the heat removal efficiency from the fuel rod surface. In the presented work, the DNS of turbulent flows are performed in three different types of geometries: (a) a regular Pressurized Water Reactor (PWR) subchannel; (b) a special PWR subchannel with reduced size spacer grid and mixing vanes (SGMV); (c) a special PWR subchannel with reduced size SGMV, but with a longer domain size. The springs and dimples of the spacer grid are not represented for the simplicity in the present investigation. The mean flow velocity and turbulent kinetic energy profiles are averaged from the instantaneous DNS data in different types of geometries as well as various downstream locations with respect to SGMV region. The results will help better assess the influence of spacer grid and mixing vanes on core coolant flows.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Multi-Physics Simulations of Heat Pipe Micro Reactor

System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report presents the development of new capabilities for modeling the heat pipe type reactor systems. The need for power at remote locations away from a reliable electrical grid is an important niche for nuclear energy. Heat pipe-cooled fast-spectrum nuclear reactors are well suited for these applications. The key feature of the heat pipe reactors is the use of heat pipes for heat removal from the reactor core. The heat pipe makes use of the phase change of the working fluid and transports a large amount of heat from the evaporator to the condensation end with very small temperature drops. In contrast to the traditional nuclear reactor system that makes use of pumped loop for extracting the thermal power, the heat pipe reactors make use of hundreds of heat pipes for removing the thermal power (including the decay heat) passively. This could potentially significantly improve the reliability and safety of the reactor systems. The essential part in the analysis of a heat pipe type reactor is the modeling of heat transport inside the heat pipe. The capability of SAM was extended to enable the modeling of the conventional heat pipe. The full suite of non-LWR codes for confirmatory analysis at NRC is known as the Comprehensive Reactor Analysis Bundle (CRAB). It makes use of existing NRC codes, and integrates them with several codes developed through the DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. In this report, the heat pipe micro reactor is simulated using three submodules under CRAB. The MAMMOTH module is used to simulate the reactor kinetics behavior of the micro reactor; the SAM module is used to simulate the heat conduction in the reactor core and heat removal through the heat pipe heat exchangers and Reactor Cavity Cooling System (RCCS); and the MOOSE Tensor Mechanics module is used to simulate the thermal expansion of the reactor cores. The different sub-models are coupled together using MOOSE’s MultiApp system and executed using the BlueCrab application. The new capability has been demonstrated by a steady-state operation analysis, a failure of a single central heat pipe transient analysis, and a loss of heat sink transient analysis. The fully coupled model is shown to work well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Modeling of a Passive Endothermic Reaction Cooling System

Passive reactor core cooling systems can mitigate the effects of station blackout conditions and avoid accidents similar to what occurred at the Fukushima Dai-ichi reactors. Advanced nuclear reactor designs such as the Advanced Heavy Water Reactor (AHWR) and others include passive cooling systems. The AHWR utilizes water storage tanks that can remove decay heat for more than 50 days after station blackout. While these systems do not rely on electricity, they do rely on passive valves to function during an emergency situation. The purpose of the present work is to determine the feasibility of a passive endothermic reaction cooling system (PERCS). The PERCS relies on an endothermic reaction that does not proceed until it reaches a threshold temperature. Thus, a vessel filled with chemicals that undergo an endothermic reaction can be directly connected to the reactor core without any valves and only react once accident conditions raise the temperature past the threshold temperature. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1968

Nuclear-Rocket Propulsion

The three basic types of nuclear power-plants (solid, liquid, and gas core) are compared on the bases of performance potential and the status of current technology. The solid-core systems are expected to have impulses in the range of 850 seconds, any thrust level (as long as it is greater than 10,000 pounds (44,480 newtons)), and thrust-to-engine-weight ratios of 2 to 20 pounds per pound (19.7 to 197 newtons per kilogram). There is negligible or no fuel loss from the solid-core system. The solid-core system, of course, has had the most work done on it. Large-scale tests have been performed on a breadboard engine that has produced specific impulses greater than 700 seconds at thrust levels of about 50,000 pounds (222,000 newtons). The liquid-core reactor would be interesting in the specific impulse range of 1200 to 1500 seconds. Again, any thrust level can be obtained depending on how big or small the reactor is made. The thrust-to-engine weight ratio for these systems would be in the range of 1 to 10. The discouraging feature of the liquid-core system is the high fuel-loss ratio anticipated. Values of 0.01 to 0.1 pound (0.00454 to 0.0454 kilograms) or uranium loss per pound (0.454 kilograms) of hydrogen are expected, if impulses in the range of 1200 to 1500 seconds are desired. The gas-core reactor shows specific impulses in the range of 1500 to 2500 seconds. The thrust levels should be at least as high as the weight so that the thrust-to-weight ratio does not go below 1. Because the engine weight is not expected to be under 100,000 pounds (444,800 newtons), thrust levels higher than 100,000 pounds (448,000 newtons) are of interest. The thrust-to-engine weights, in that case, would run from 1 to 20 pounds per pound (9.8 to 19.7 kilograms). Gas-core reactors tend to be very large, and can have high thrust-to-weight ratios. As in the case of the liquid-core system, the fuel loss that will be attendant with gas cores as envisioned today will be rather high. The loss rates will be 0.01 to 0.1 pound of uranium (0.00454 to 0.0454 kilograms) for each pound (0.454 kilograms) of hydrogen.

Rom, Frank E.↗
DOE OSTI2022

Boron injection system for controlling a nuclear reaction by delivering boron into a containment vessel

A nuclear power system includes a reactor vessel that includes a reactor core mounted within a volume of the reactor vessel. The reactor core includes one or more nuclear fuel assemblies configured to generate a nuclear fission reaction. The nuclear power system further includes a containment vessel sized to enclose the reactor vessel such that an open volume is defined between the containment vessel and the reactor vessel. A boron injection system is positioned in the open volume of the containment vessel and includes an amount of boron sufficient to stop the nuclear fission reaction or maintain the nuclear fission reaction at a sub-critical state. The boron injection system is positioned to deliver the amount of boron into the open volume.

Linik, Larry↗
NASA NTRS1972

Closed loop dynamics of in-core thermionic reactor systems.

Use of a point model of an in-core thermionic converter to investigate alternative schemes for providing closed-loop reactor control. It was found that schemes based on variable-gain power regulation buffers which use the reactor current as the control variable provide complete protection from thermionic burnout and also provide a virtually constant voltage to the user. A side benefit is that the emitter temperature transients are small; even for a complete electric load drop the emitter temperature transient is less than 100 K. The current regulation scheme was selected for further study with a distributed parameter model which was developed to account for variations in thermionic and heat transfer properties along the length of a cylindrical converter. It was found that, even though the emitter temperature distribution is about 200 K along the converter length, the dynamic properties are unchanged when using the current control scheme.

Sawyer, C. D.↗
DOE OSTI2022

On-the-fly response function generation method for composite coarse mesh

The hybrid stochastic deterministic transport code COMET, based on the incident response expansion theory, is used to model reactor cores with high fidelity and formidable computational speed. COMET models a reactor core using a library of incident flux response expansion coefficients that are pre computed for all the unique lattice cells (e.g., fuel assemblies, reflector blocks, etc.) in the core. In order to further improve its computational efficiency in pre-calculating the response library a new response function generation method is developed to compute the response functions for the composite coarse meshes made of a smaller set of unique lattices on the fly within the COMET's deterministic transport core sweep. The efficiency is achieved by eliminating a number of unique lattices that can be made up from the reduced set of unique meshes on the fly. The numerical process consists of the following steps. First, the boundary condition on composite coarse mesh boundaries is projected onto the expansion basis to compute the incident flux moments on external surfaces of all the basic (reduced set of unique) coarse meshes. Secondly, the deterministic sweeping solver in COMET is used to converge on the outgoing/incoming flux expansion moments crossing interfaces between the basic coarse meshes. Thirdly, the response functions for the composite coarse meshes are constructed as a superposition on the fly. The new response function generation method was tested on 88 composite coarse meshes consisting of CANDU fuel bundles and moderator blocks. It was found that response functions generated by the new method agree very well with those generated by direct Monte Carlo calculations. The average and maximum relative differences in the surface-to-surface response coefficients computed by the two methods are 0.10% and 0.20%, respectively. Similarly, the average and maximum relative differences in the response fission densities are 0.13% and 0.43%, respectively. These discrepancies are within one standard deviation of the stochastic uncertainties. The new method is five times faster than the original direct Monte Carlo method. The size of the response function library for the new method is five times smaller than that for the original method, leading to significantly less requirement for the computer hard drive space and memory. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239 Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2 /fission and a fuel-dependent variation in the IBD yield, dσ f /dF 239 , of (–1.86±0.18)×10 –43 cm 2 /fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239 Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235 U, 239 Pu, 238 U, and 241 Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2 /fission have been determined for the two dominant fission parent isotopes 235 U and 239 Pu. A 7.8% discrepancy between the observed and predicted 235 U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
NASA NTRS1991

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Clark, John S.↗
DOE OSTI2023

Enhanced graphite neutron reflector with beryllium oxide inclusions

A nuclear reactor is disclosed including a reactor core and a reflector assembly surrounding the reactor core. The reflector assembly includes a stationary reflector component including a graphite support structure comprising a plurality of channels defined therein and a plurality of beryllium-oxide pins positioned in the channels.

Aleshin, Yuriy↗
DOE OSTI2016

Neutronics/Thermal-hydraulics coupling with RMC and CTF for BEAVRS benchmark calculation

As the computational accuracy required for design and safety analysis of current and future nuclear reactors is continuously increasing, high fidelity numerical reactor simulations including different physics raise wide concern of researchers. The interplay between neutronic and thermal-hydraulic (N-TH) effects of a nuclear reactor core plays an important role in reactor design, safety analysis and core performance. Monte Carlo method can provide high fidelity neutronics analysis of different nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. Armed with the massively parallel algorithms, the computational cost of Monte Carlo codes can be greatly reduced. Compared to single channel thermal process, sub-channel codes have enormous advantages of detailed modeling of full-core reactor or single assembly and high accuracy of calculation results. Detailed sub-channel modeling of a full-core also needs running in parallel, otherwise the computational cost will be relatively high. In this paper, a new hybrid coupling with continuous-energy Monte Carlo code RMC and the sub-channel code CTF has been developed. The coupling codes are applied to steady-state simulations of BEAVRS benchmark in hot full power condition. In order to deal with the temperature dependence of cross sections for Monte Carlo code, the on-the-fly cross sections treatment has been developed in RMC. The on-the-fly method was applied to the N-TH coupling in this work. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Fukushima Daiichi reactor source term attribution using cesium isotope ratios from contaminated environmental samples

Source term attribution of environmental contamination following the Fukushima Daiichi Nuclear Power Plant (FDNPP) disaster is complicated by a large number of possible similar emission source terms (e.g. FDNPP reactor cores 1–3 and spent fuel ponds 1–4). Cesium isotopic analyses can be utilized to discriminate between environmental contamination from different FDNPP source terms and, if samples are sufficiently temporally resolved, potentially provide insights into the extent of reactor core damage at a given time. In this work, rice, soil, mushroom, and soybean samples taken 100–250 km from the FDNPP site were dissolved using microwave digestion. Radiocesium was extracted and purified using two sequential ammonium molybdophosphate-polyacrylonitrile columns, following which 135 Cs/ 137 Cs isotope ratios were measured using thermal ionization mass spectrometry (TIMS). Results were compared with data reported previously from locations to the northwest of FDNPP and 30 km to the south of FDNPP. 135 Cs/ 137 Cs isotope ratios from samples 100–250 km to the southwest of the FDNPP site show a consistent value of 0.376 ± 0.008. 135 Cs/ 137 Cs versus 134 Cs/ 137 Cs correlation plots suggest that radiocesium to the southwest is derived from a mixture of FDNPP reactor cores 1, 2, and 3. Conclusions from the cesium isotopic data are in agreement with those derived independently based upon the event chronology combined with meteorological conditions at the time of the disaster. In conclusion, cesium isotopic analyses provide a powerful tool for source term discrimination of environmental radiocesium contamination at the FDNPP site. For higher precision source term attribution and forensic determination of the FDNPP core conditions based upon cesium, analyses of a larger number of samples from locations to the north and south of the FDNPP site (particularly time-resolved air filter samples) are needed. Published in 2016. This article is a U.S. Government work and is in the public domain in the USA.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗