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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Security by Design Economics Analysis for Advanced Reactors and Small Modular Reactors Project Interim Report for FY2021

Advanced Reactor and Small Modular Reactor (AR/SMR) designs have the potential to provide clean, reliable baseload energy. Ensuring the capability to deploy these reactors in an economically viable fashion is of interest to industry. A large portion of the expected operating costs of AR/SMRs involves the security of the plant. Security by Design (SeBD) is the practice of including features in the design and construction of the site, with the intent to decrease the operating costs related to security. Quantifying the increase or decrease in the overall lifetime cost to the plant as a result of SeBD is of paramount importance in understanding the disadvantages and benefits of such activities. The National Nuclear Security Administration’s (NNSA) Office of International Nuclear Security (INS) is funding the development of a methodology whereby the capital expenses and operating expenses, as well as the physical security effectiveness, of SeBD can be quantified for AR/SMRs. This report is an interim report on the progress of the work performed by Sandia National Laboratories (SNL), Idaho National Laboratory, and Oak Ridge National Laboratory (ORNL). It is the second annual report on this work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Reactor Physics Simulations of the High Flux Isotope Reactor Permanent Beryllium Reflector Number 5 Design

This study evaluates reactor physics aspects of the proposed High Flux Isotope Reactor (HFIR) permanent beryllium reflector number 5 design. HFIR is a pressurized, light water–cooled, and beryllium reflected research reactor that operates at 85 MW for cycle lengths of approximately 24 to 26 days. A new permanent reflector design is highly desirable for increased versatility in irradiation experiments, to arrange the vertical experiment facilities (VXFs) to minimize their impact on neutron scattering if loaded with neutronabsorbing experiments, to enhance reflector thermal-structural-hydraulic performance, and to simplify the complex fabrication process. The current number 4 and concept number 5 reflectors have 22 and 28 VXFs, respectively, and are modeled and analyzed in this study with the MCNP5 and SCALE 6.1.3 nuclear simulation codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Generation of localized reactor point kinetics parameters using coupled neutronic and thermal fluid models for pebble-bed reactor transient analysis

The systems analysis of anticipated operating occurrences and design basis accidents for pebble-bed reactor systems requires knowledge of neutron point kinetics equations (PKE) parameters. Typically, the generation of PKE parameters is performed in a global manner using standalone neutronics calculations, without the inclusion of thermal fluid distributions. We utilize Griffin and Pronghorn for generating global and local PKE parameters which includes the use of thermal fluid distributions to account for localized effects. This work establishes a methodology for calculating PKE parameters for a pebble bed reactor with a coupled neutronics/thermal fluids analysis. PKE parameters generated on a global and local basis are compared against a diffusion solve for a typical load-following transient. Locally-generated neutron kinetic parameters are able to reduce the maximum error in the transient power level from 5% to below 1.5%; along with this, bulk temperature errors were reduced from 12 K to 4 K.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Comparative analysis of energy deposition modes available in Serpent 2 within the framework of the supercritical water reactor - Fuel qualification test reactor physics benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor (SCW-SMR) technology is in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. As a main purpose of the project, proper estimates of energy deposition and its spatial distribution are prerequisites for the accurate analysis of safety related parameters of the SCW-SMR concept under development. A supercritical water reactor fuel computational benchmark model, provided by Canadian Nuclear Laboratories, was applied for detailed comparison of different energy deposition calculation options available in the Serpent 2 Monte Carlo code. The effect of energy deposition options on the normalization of the results as well as on the spatial distribution of the energy deposition are discussed. Consistent energy deposition calculation methods are presented between three Monte Carlo codes, viz., Serpent 2, MCNP6 and OpenMC. Although resource-intensive, the use of the coupled neutron-photon transport mode of Serpent 2 is recommended for accurate spatial and quantitative characterization of energy deposition in the SCW-SMR fuel assemblies, accounting for both neutron and photon heating of all the materials. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Design and characterization of an electromagnetic-resonant cavity microwave plasma reactor for atmospheric pressure carbon dioxide decomposition

Plasma processes are ideally suited for the conversion of renewable electricity into gas-phase reactivity, such as for the decomposition of carbon dioxide (CO 2 ). The design, development, and characterization of a microwave plasma reactor for atmospheric pressure undiluted carbon dioxide decomposition are presented. The reactor operates as an electromagnetic-resonant cavity in which the generated plasma forms a bulb attached to a converging-diverging nozzle and stabilized by streams of tangentially-injected processing gas. Electromagnetic wave confinement, residence time, and critical gas vorticity constitute fundamental reactor sizing and operation parameters. Experimental findings demonstrate that flow rate plays a dual role in plasma stabilization and process performance, whereas deposited power has a minor role in CO 2 decomposition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Reactor Performance Improvement Options to Sustain High Flux Isotope Reactor Leadership into the Future

The mission of the Neutron Sciences Directorate (NScD) at the U.S. Department of Energy’s Oak Ridge National Laboratory (ORNL) is the undertaking of high-impact research into the structure and properties of materials across the spectrum of biology, chemistry, physics, materials science, and engineering. NScD operates two world-leading neutron scattering facilities including the High Flux Isotope Reactor (HFIR) and the Spallation Neutron Source. HFIR achieved full power in 1966, and over a half century later, it continues to serve a variety of national missions. HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world to support scientific missions including cold and thermal neutron scattering, isotope production, and materials irradiation research. To sustain leadership in neutron sciences into the future, ORNL is exploring areas in which HFIR can be improved to enhance its performance. Many improvement areas are being explored such as upgrading the cold source and neutron scattering facilities; however, the improvement areas focused on in this paper include replacing the reactor pressure vessel, upgrading the neutron reflector, and converting from high-enriched uranium to low-enriched uranium fuel.

Chandler, David↗
DOE OSTI2020

Characterization of Uranium Foil Irradiations at the WSU TRIGA Reactor Using a New Reactor Model in SCALE [Slides]

A depleted uranium foil and a highly enriched uranium foil that were irradiated at the Washington State University TRIGA reactor were used to measure the relative production of 237 U and 239 U, two key isotopes of interest in nuclear forensics, as well as their decay daughters 237 Np and 239 Pu. Total fissions in the two foils were also estimated using 99 Mo measurement. A reactor model of the TRIGA was built in the SCALE neutron transport code using dimensions obtained from an existing MCNP geometry. The model was successfully used to simulate foil fissions and the reaction rate ratios in 238 U, using the laboratory-measured isotope ratios as a benchmark. The flux distribution in the foils was determined. Finally, using the continuous energy Tsunami-3d module of SCALE, a sensitivity/uncertainty analysis was performed to determine the effects on the reaction ratio caused by other reactions in the reactor and by uncertainties in the known cross-sections.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Assessment of ORIGEN Reactor Library Development for Pebble-Bed Reactors Based on the PBMR-400 Benchmark

This report provides an evaluation of present SCALE capabilities for modeling depletion of pebble-bed reactor systems, using the PBMR-400 benchmark as a test case. A specific aim of this work is to understand the system characteristics required to generate production-quality reactor data libraries for rapid depletion calculations with ORIGEN. This report includes a discussion of present SCALE capabilities for modeling doubly heterogeneous fuels, prior SCALE work modeling pebble bed–type reactors, and a detailed neutronic analysis of the PBMR-400 core for both fresh and equilibrium-composition core conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Material Test Reactors; Annular Core Research Reactor Testing

The Material Test Reactors Fifth Workshop reviews the successes and failures of material test protocols and lessons learned. This presentation introduces the Annular Core Research Reactor (ACRR), the history of nuclear reactors at Sandia National Labs and the experiment review process. Included are actual examples of experimental activities at ACRR.

Martin, Lonnie E. [Sandia National Laboratories (S↗
NASA NTRS1991

Fission fragment assisted reactor concept for space propulsion: Foil reactor

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

Wright, Steven A.↗
DOE OSTI2024

Thermal Gradient and Neutron Irradiation Experiment Design for Fusion Reactor Materials in the Advanced Test Reactor

This work outlines a hypothetical coupled thermal gradient and neutron irradiation experiment in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Although the ATR is a thermal spectrum test reactor and doesn’t inherently produce a flux spectrum dominated by the high-energy neutrons typical in a fusion reactor, it’s multitude of experiment positions and dynamic flux environment make it a suitable platform for investigating fusion related issues.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Design Basis Model for Hosting Small Modular Reactors

An aggressive transition from fossil fuels to other types of energy implies the need to construct a large number of nuclear power plants in the near future. However, the real and perceived risks of nuclear energy remain a significant impediment to this transition. This paper describes a comprehensive work process that combines the rigor of model-based systems engineering (MBSE) with 1) the Idaho National Laboratory's (INL) decades of experience with small reactors and with 2) modern project delivery processes. The objective is to reduce the risk of building new facilities or converting existing facilities to nuclear power generation.

42 ENGINEERING↗
DOE OSTI2018

Reactor physics phenomena in additively manufactured control elements for the High Flux Isotope Reactor

Additive manufacturing is under investigation as a novel method of fabricating the control elements (CEs) of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory with greater simplicity, eliminating numerous highly complex fabrication steps and thereby offering potential for significant savings in cost, time, and effort. This process yields a unique CE design with lumped absorbers, a departure from traditionally manufactured CEs with uniformly distributed absorbing material. Here, this study undertakes a neutronics analysis of the impact of additively manufactured CEs on the HFIR core physics, seeking preliminary assessment of the feasibility of their practical use. The results of the MCNP transport simulations reveal changes in the HFIR reactor physics arising from geometric and nuclear effects. Absorber lumping in the discrete CEs yields a large volume of unpoisoned material that is not present in the homogeneous design, in turn yielding increases in free thermal flux in the CE absorbing regions and their immediate vicinity. The availability of additional free thermal neutrons in the core yields an increase in fission rate density in the fuel closest to the CEs and a corresponding increase in neutron multiplication on the order of 100 pcm. The absorption behavior exhibited by the discrete CEs is markedly different from the homogeneous CEs due to several competing effects. Self-shielding arising from absorber lumping acts to reduce the effective absorption cross section of the discrete CEs, but this effect is offset by geometric and spectral effects. The operational performance of the discrete CEs is found to be comparable to the homogeneous CEs, with only limited deficiencies in reactivity worth that are expected to be operationally recoverable via limited adjustment of the CE positions and withdrawal rate. On the whole, these results indicate that the discrete CEs perform reasonably similarly to the homogeneous CEs and appear feasible for application in HFIR. In conclusion, the physical phenomena identified in this study provide valuable background for follow-up design studies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗