Search NASA⌕ Search

SEARCH · Search NASA

Results for “neutron moderator”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Chiral Effective Field Theory and the High-Density Nuclear Equation of State

Born in the aftermath of core-collapse supernovae, neutron stars contain matter under extraordinary conditions of density and temperature that are difficult to reproduce in the laboratory. In recent years, neutron star observations have begun to yield novel insights into the nature of strongly interacting matter in the high-density regime where current theoretical models are challenged. At the same time, chiral effective field theory has developed into a powerful framework to study nuclear matter properties with quantified uncertainties in the moderate-density regime for modeling neutron stars. In this article, we review recent developments in chiral effective field theory and focus on many-body perturbation theory as a computationally efficient tool for calculating the properties of hot and dense nuclear matter. We also demonstrate how effective field theory enables statistically meaningful comparisons among nuclear theory predictions, nuclear experiments, and observational constraints on the nuclear equation of state.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Direct thermal neutron detection by the 2D semiconductor 6 LiInP 2 Se 6

Highly efficient neutron detectors are critical in many sectors, including national security, medicine, crystallography and astronomy. The main neutron detection technologies currently used involve 3 He-gas-filled proportional counters and light scintillators for thermalized neutrons. Semiconductors could provide the next generation of neutron detectors because their advantages could make them competitive with or superior to existing detectors. In particular, solids with a high concentration of high-neutron-capture nuclides (such as 6 Li, 10 B) could be used to develop smaller detectors with high intrinsic efficiencies. However, no promising materials have been reported so far for the construction of direct-conversion semiconductor detectors. Here we report on the semiconductor LiInP 2 Se 6 and demonstrate its potential as a candidate material for the direct detection of thermal neutrons at room temperature. This compound has a good thermal-neutron-capture cross-section, a suitable bandgap (2.06 electronvolts) and a favourable electronic band structure for efficient electron charge transport. We used α particles from an 241 Am source as a proxy for the neutron-capture reaction and determined that the compact two-dimensional (2D) LiInP 2 Se 6 detectors resolved the full-energy peak with an energy resolution of 13.9 per cent. Direct neutron detection from a moderated Pu–Be source was achieved using 6 Li-enriched (95 per cent) LiInP 2 Se 6 detectors with full-peak resolution. Here, we anticipate that these results will spark interest in this field and enable the replacement of 3 He counters by semiconductor-based neutron detectors.

36 MATERIALS SCIENCE↗
NASA NTRS2007

Ground-Based Testing of TiB2 and Al2O3/TiB2 Response to Space Environment

Two materials, titanium diboride and an alumina/titanium diboride composite, exhibit characteristics favorable for use in multiple space applications. These characteristics include low mass (4.52 gm/cc), high strain rate impact resistance, high temperature use (3000oC M.P.), thermal and electrical conductivity, thermal shock resistance, and high visible-range reflectivity. Additionally, the presence of boron in these materials gives them the potential to shield against neutron radiation as well as charged radiation. These materials are flying on MISSE 6 to assess material changes resulting from exposure to the space environment. This study provides a preliminary, ground-based examination of these materials' interactions with individual components of the space environment, in particular atomic oxygen (AO) and neutron radiation, in order to better predict and understand post-flight results. Individual specimens are exposed to ground state AO and surface oxidation is measured. Equivalent exposures of up to 13 months show no rapid oxidation, however evidence indicates some surface oxidation occurring. Other samples are placed near a polyethylene moderated, one Ci Am/Be neutron source to determine their shielding capability. Comparisons between exposed and shielded indium foil, which is activated by transmitted neutrons, measure each material's ability to shield neutrons. Preliminary results indicate a significant shielding benefit provided by both materials.

Jefferies, Sharon A.↗
DOE OSTI2023

Thermal Spectrum Molten Salt-Fueled Reactor Reference Plant Model

This report details a fully coupled neutronics thermal hydraulics reference plant model for a thermal spectrum molten salt-fueled reactor. The Multiphysics model is developed on the Nuclear Regulatory Commission's Comprehensive Reactor Analysis Bundle available on the Idaho National Laboratory high-performance computer, which natively and seamlessly couples Griffin and Sam codes of the Multiphysics Object-Oriented Simulation Environment (MOOSE) based applications. Griffin provides the reactor physics capabilities, k-eigenvalue, delayed neutron precursor concentrations, and transient solutions for stationary and flowing fuels. Sam provides the solution of mass, momentum, and energy conservation equations of the whole system with a set of system components for one-dimensional single-phase flow. The neutronics feedback model relies primarily on fuel and moderator temperatures, fuel salt density, and delayed neutron precursors distributions. This report presents our results for the coupled steady state core and an unprotected loss of flow event. Although this model is prototypical regarding capabilities in the Comprehensive Reactor Analysis Bundle, its results are consistent with experimental data of the pump transient tests that were performed at zero power of the Molten Salt Reactor Experiment (MSRE).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Physical and Thermomechanical Properties of Yttrium Hydride from Large Scale Bulk Metal Hydriding Furnace

Given the superior thermal stability and highly attainable hydrogen density, yttrium hydride is an excellent high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. Yttrium hydride has been selected as the moderator material for the Transformational Challenge Reactor, which was launched at Oak Ridge National Laboratory (ORNL) in 2019. However, fabrication of large-scale crack-free yttrium hydride is challenging and very limited efforts have been committed to the characterization of bulk yttrium hydride in response to the need to establish a complete database of the thermomechanical properties of YHx. In this report, the challenges associated with fabricating large-scale crack-free yttrium hydride are discussed herein. In response to those challenges, a hydriding system was designed and constructed at ORNL and was used to successfully fabricate crack-free yttrium hydride in complex geometries at large scales. This was accomplished by precisely controlling the hydrogen’s partial pressure and the retort temperature, which was informed by the well-established thermodynamic properties of the binary H-Y system. Hydrogen content in as-fabricated hydride was determined by the weight change method and vacuum hot extraction technique, complemented by the X-ray diffraction (XRD). In addition, significant efforts are being dedicated to establishing a complete database of the thermomechanical properties of as-fabricated yttrium hydride. In FY2020, we investigated the thermophysical properties of yttrium hydrides as a function of temperature (room temperature to 700°C) and hydrogen concentration (H/Y ratio ranges from 1.52 to 1.93). The results indicate that at the temperatures below 300 °C, the hydrogen content did not have a significant influence on the thermal expansion, while the specific heat capacity, the thermal diffusivity, and the calculated thermal conductivity were slightly higher for the higher H/Y ratio. Between 300°C and 700 °C, a reversible second-order endothermic transition in all measured thermal properties was observed. It was also found that the onset temperatures of the observed transition varied, with the composition having inverse dependence on the hydrogen content. An attempt was made to explain the behavior of the thermophysical properties at higher temperatures by considering the order– disorder transition as a result of hydrogen redistribution. In addition, nanoindentation was employed to determine the elastic modulus and hardness and to capture the crystal orientation dependence of these parameters. Vickers hardness was also reported. The final section of the report introduces ongoing neutron irradiation campaign of yttrium hydride.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1995

Effect of neutron irradiation on the London penetration depth for polycrystalline Bi(1.8)Pb(0.3)Sr2Ca2Cu3O10 superconductor

Magnetization studies of polycrystalline Bi(1.8)Pb(0.3)Sr2Ca2Cu3O10 superconductor, prior to and after neutron irradiation, showed an increase in J(sub c) due to irradiation damage. Analysis of the equilibrium magnetization revealed significant increases in other more fundamental properties. In particular, the London penetration depth increased by approximately 15 percent following irradiation with 8 x 10(exp 16) neutrons/sq cm. Corresponding changes were observed in the upper critical magnetic field H(sub c2). However, the most fundamental thermodynamic property, the superconductive condensation energy F(sub c), was unaffected by the moderate level of neutron-induced damage.

Ossandon, J. G.↗
NASA NTRS2011

Optical Sensors for Monitoring Gamma and Neutron Radiation

For safety and efficiency, nuclear reactors must be carefully monitored to provide feedback that enables the fission rate to be held at a constant target level via adjustments in the position of neutron-absorbing rods and moderating coolant flow rates. For automated reactor control, the monitoring system should provide calibrated analog or digital output. The sensors must survive and produce reliable output with minimal drift for at least one to two years, for replacement only during refueling. Small sensor size is preferred to enable more sensors to be placed in the core for more detailed characterization of the local fission rate and fuel consumption, since local deviations from the norm tend to amplify themselves. Currently, reactors are monitored by local power range meters (LPRMs) based on the neutron flux or gamma thermometers based on the gamma flux. LPRMs tend to be bulky, while gamma thermometers are subject to unwanted drift. Both electronic reactor sensors are plagued by electrical noise induced by ionizing radiation near the reactor core. A fiber optic sensor system was developed that is capable of tracking thermal neutron fluence and gamma flux in order to monitor nuclear reactor fission rates. The system provides near-real-time feedback from small- profile probes that are not sensitive to electromagnetic noise. The key novel feature is the practical design of fiber optic radiation sensors. The use of an actinoid element to monitor neutron flux in fiber optic EFPI (extrinsic Fabry-Perot interferometric) sensors is a new use of material. The materials and structure used in the sensor construction can be adjusted to result in a sensor that is sensitive to just thermal, gamma, or neutron stimulus, or any combination of the three. The tested design showed low sensitivity to thermal and gamma stimuli and high sensitivity to neutrons, with a fast response time.

Boyd, Clark D.↗
DOE OSTI2022

Neutron Leakage Spectra Sensitivities for ICSBEP Benchmarks

Neutron leakage spectra have been measured, simulated, and investigated by many groups. These spectra have many uses, including determining shielding requirements and calculating dose for radiation protection purposes, validating nuclear data, and determining material composition. It has even been proposed to use measurements of neutron leakage spectra to determine the soil composition of Mars. As part of the Los Alamos National Laboratory (LANL) Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) Laboratory Directed Research Development (LDRD) project, the authors are working on developing methods to calculate sensitivities to neutron leakage spectra. This may allow nuclear data evaluators to better use neutron leakage spectra data to constrain or adjust nuclear data. This may also be particularly useful as many neutron leakage spectra measurements do not require fissile material and can be performed with well characterized neutron sources. Neutron leakage spectra measurements are usually performed by placing a detector at the outside of a nuclear system. This may be a reactor, a neutron generator, or a neutron source. Certain detection systems can use pulse height data to infer neutron energies, other detector systems rely on other ways of determining neutron energy (for example, a Bonner sphere with multiple moderator thicknesses can be used to measure neutron spectrum). These measurements typically rely on some unfolding of the measured results. For pulsed systems like the "Livermore Pulsed Spheres," time-of-flight information can also help to constrain the neutron energy spectra data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter

This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252 Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Development of the MPACT 69-group Library for Magnox Reactor Analysis using CASL VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnoxreactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Kim, Kang Seog↗
DOE OSTI2020

Feasibility Study for a Subcritical Assembly at the Oak Ridge National Laboratory

A conceptual design has been completed for a new subcritical assembly at Oak Ridge National Laboratory (ORNL). This subcritical assembly will support the Nuclear Criticality Safety Program (NCSP) training and education program to augment current subcritical assembly capabilities at Sandia National Laboratories (SNL) and at the National Criticality Experiments Research Center (NCERC). The proposed subcritical assembly uses legacy AGN-201M research reactor fuel plates that are available at the Y-12 National Security Complex. This subcritical assembly contains approximately 620 grams of 235 U as UO 2 particulates distributed homogeneously in radiation-stabilized polyethylene. The fuel plates will have a graphite neutron reflector to obtain a core multiplication, $M$, from 10 to 50, corresponding to a k eff of 0.9 to 0.95, respectively. The subcritical assembly will be able to support at least four experiments for the training courses: the addition of fissile material to increase neutron multiplication (mass), a core separation experiment (interaction), the effect of adding moderators to the core (moderation), and the effect of adding neutron absorbers to the core (poison/absorption). The proposed ORNL subcritical assembly will provide additional student bandwidth to support NCSP training courses and to provide for some backup capacity in the event that nuclear facility operations are disrupted at SNL or NCERC due to some unforeseen issue. The subcritical assembly will be designed to be an inherently safe subcritical assembly—subcritical under all normal and abnormal conditions—and it will provide the capability to conduct hands-on training to support NCSP and general nuclear criticality safety (NCS) staff training and qualification goals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Integral Experiment Final Design for Thermal/Epithermal eXperiments with 233U3O8 ZPR Fuel Elements and Polyethylene (IER-329 CED-2 Report)

This report documents the integral experiment final design for IER-329, Thermal/Epithermal eXperiments (TEX) with 233 U 3 O 8 Zero Power Reactor (ZPR) Fuel Elements and Polyethylene. IER-329 (TEX-23) is the third baseline design in the Thermal/Epithermal eXperiments (TEX) series led by Lawrence Livermore National Laboratory (LLNL) and funded by the DOE’s Nuclear Criticality Safety Program. This experiment was designed to address the existing 233U integral benchmark experiments over prediction in k eff by up to 2% for thermal systems and under-prediction in k eff by up to 4% for intermediate systems. The TEX-23 design utilizes assemblies of 233 U 3 O 8 Zero Power Reactor (ZPR) fuel elements moderated and reflected by high-density polyethylene. This design is similar to the existing TEX-Pu and TEX-HEU designs, utilizing an adjustable moderator to tune the neutron energy spectra and modular layers for the incorporation of diluent materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Proton-rich Production of Lanthanides: The vi Process

The astrophysical origin of the lanthanides is an open question in nuclear astrophysics. Besides the widely studied s, i, and r processes in moderately to strongly neutron-rich environments, an intriguing alternative site for lanthanide production could in fact be robustly proton-rich matter outflows from core-collapse supernovae under specific conditions—in particular, high-entropy winds with enhanced neutrino luminosity and fast dynamical timescales. In this environment, excess protons present after charged-particle reactions have ceased can continue to be converted to neutrons by (anti)neutrino interactions, producing a neutron-capture reaction flow up to A ∼ 200. This scenario, christened the νi process in a recent paper, has previously been discussed as a possibility. Here, we examine the prospects for the νi process through the lenses of stellar abundance patterns, bolometric light curves, and galactic chemical evolution models, with a particular focus on hypernovae as candidate sites. We identify specific lanthanide signatures for which the νi process can provide a credible supplement to the r/i processes.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Comparison of the Neutronic Performance of IRP-1 and IRP-2 at SNS

The Spallation Neutron Source (SNS) operated from start-up, in May 2006, through December, 2017 before replacing the Inner Reflector Plug (IRP). This component is the most costly and time-consuming “regular maintenance” replacement foreseen in the operation of the First Target Station (FTS). The IRP includes beryllium reflector blocks, steel and aluminum support structures, aluminum moderator vessels (which in turn include gadolinium and cadmium neutron poisons) and water coolant systems. The lifetime of this multi-million dollar component is driven by poison and decoupler burnup in moderator vessels. The moderators include these neutron poisons to best match neutron beam performance with scattering instrument needs. The burnup that comes with use changes the performance of the moderators until they no longer server their intended purpose. The reasonable desire to maximize the IRP lifetime means that the moderator performance at the Beginning Of Life (BOL) is significantly “over poisoned” and different from the performance at moderator End Of Life (EOL). Upon installation of the second IRP article, IRP-2, the SNS Neutronics team characterized the performance of the neutron beamlines in order to assess the differences between them, primarily by comparing the EOL performance of IRP-1 to the BOL performance of IRP-2. While IRP-2 was largely similar to IRP-1 in its conceptual design, there were some differences in the specific design of IRP-2 to support enhanced manufacturability, and the change from IPR-1 to IRP-2 coincided with the replacement of the light water in the IRP cooling loop with heavy water, as had been intended in the SNS design but not implemented during construction because of heavy water availability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Handbook on the Material Properties of Yttrium Hydride for High Temperature Moderator Applications

Yttrium hydride is an optimal choice for a high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. A complete database of the thermomechanical properties of yttrium hydride is not yet available, although it is much needed to understand and predict the moderator performance during service in reactors. This milestone report presents the properties of unirradiated bulk yttrium hydride as a function of hydrogen concentration—including density, crystal structure, specific heat capacity, thermal diffusivity, thermal conductivity, hardness, elastic/shear moduli, Poisson’s ratio, fracture strength, microstructure, and thermal stability. This information provides a baseline measurement for the subsequent neutron irradiation response study of yttrium hydride. The recommended empirical treatment of the data is suggested. In addition, other properties (i.e., hydrogen retention, thermal hydrogen migration, and irradiation response) requiring investigation are discussed. Also included are the Preliminary post-irradiation examination (PIE) data of yttrium hydride irradiated at 600 and 900°C to 0.1 dpa in the High Flux Isotope Reactor (HFIR). The thermophysical properties have insignificant change following this low dpa irradiation for both irradiation temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Design of the cryogenic moderator system for the second target station

The Second Target Station (STS) at Oak Ridge National Laboratory will be a 700 kW pulsed spallation neutron source designed to provide the world's highest brightness cold neutron beams. In order to produce the required neutron performance, two compact liquid hydrogen moderators are located adjacent to the tungsten spallation target and must be supplied with less than 20 K hydrogen and a para hydrogen fraction of 99.8% or greater. The Cryogenic Moderator System (CMS) will consist of a single hydrogen loop feeding the two moderators in series cooled by a helium refrigerator with a cooling capacity of 2.5 kW at 17 K. The hydrogen loop consists of a hydrogen circulator, hydrogen helium heat exchanger, ortho-para converter, accumulator, transfer lines and heater. The design of the hydrogen loop is based on the CMS design of the First Target Station at the Spallation Neutron Source and some of the component designs may be reused. General hydrogen temperature control is provided by controlling the flowrate of helium to the heat exchanger. The hydrogen loop will have a constant flowrate of 0.5 L/s and remove a nuclear heat load of about 850 W from the two moderators, which is deposited both directly in the hydrogen and the adjacent hydrogen containing structures. Because the nuclear heat load is accelerator driven, the hydrogen system must remain stable when the heat load is removed instantaneously during beam trips. System stability is maintained passively with the accumulator and actively with the heater. Ionizing radiation which interacts with the liquid hydrogen drives backconversion of the hydrogen from parahydrogen to orthohydrogen. The STS moderator performance is very sensitive to small fractions of orthohydrogen requiring an ortho-para converter to maintain the hydrogen supplied to the moderators at near equilibrium parahydrogen concentration. STS CMS is in the early stage of preliminary design and current focus is evaluating component sizing and system stability during beam transients.

Janney, Jim↗