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Results for “neutron shield”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Fast Neutron Shielding [Slides]

• In Neutron shielding, neutron energy is the primary determinant of shield characteristics. • Fast Neutrons (E ≥1 MeV) must first me thermalized with high neutron scattering materials. • These thermal neutrons can then be absorbed or shielded with high cross sections for neutron absorption. • Rain-gauges, which utilize principles of nucleic detection, operate with thermal neutron detectors. • Systems such as He-3 detectors measure rain by correlating neutron moderation with rain level. • From this relationship its possible to determine unknown levels of substance with a known detector signal strength.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Designing a boron nitride polyethylene composite for shielding neutrons

Neutrons are encountered in many different fields, including condensed matter physics, space exploration, nuclear power, and healthcare. Neutrons interacting with a biological target produce secondary charged particles that are damaging to human health. The most effective way to shield neutrons is to slow them to thermal energies and then capture the thermalized neutrons. These factors lead us to consider potential materials solutions for neutron shields that maximize the protection of humans while minimizing the shield mass and adapt well to modern additive manufacturing techniques. Using hexagonal boron nitride (hBN) as a capture medium and high-density polyethylene (HDPE) as a thermalization medium, we aim to design the optimal internal structure of h 10 BN/HDPE composites by minimizing the effective dose, which is a measure of the estimated radiation damage exposure for a human. Through Monte Carlo simulations in Geant4, we find that the optimal structure reduces the effective dose up to a factor of 72 over aluminum (Al) and up to a factor of 4 over HDPE; this is a significant improvement in shielding effectiveness that could dramatically reduce the radiation exposure of occupational workers.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation

Quantification of activation in structure materials around the nuclear reactor through calculation is necessary before start of decommissioning. Then a neutron shielding calculation is carried out to estimate activation of structural materials. Neutron shielding calculation is performed out for spatially large regions around a reactor through deep penetration calculation, calculation result would have significant uncertainty. In order to minimize total amount of structure materials treated as radioactive waste, quantitative estimation on uncertainty of structural activation and its reduction is important. The uncertainties due to calculation arises from that of input data and that of calculation method. Examples of uncertainty of input data is those for nuclear reaction cross section and atom density of the nuclide in the structure. In this study, we focus on the uncertainty of cross section, which would be one of the dominant uncertainties, and try to estimate uncertainty of activation due to cross section data in a neutron shielding calculation. The random sampling method (RS) is used as the uncertainty quantification method. With the RS method, uncertainty estimation can be carried out without modification of computer codes for a neutron shielding calculation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

Thermal neutron shield and method of manufacture

A thermal neutron shield comprising boron shielding panels with a high percentage of the element Boron. The panel is least 46% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of boron shielding panels which includes enriching the pre-cursor mixture with varying grit sizes of Boron Carbide.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Thermal neutron shield and method of manufacture

A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Defect formation and transmutation behaviors in irradiated tungsten under thermal-neutron shielded and mixed spectrum conditions

Here, this study investigates the microstructural evolution of pure tungsten irradiated under thermal-neutron shielded and mixed spectrum conditions in the High Flux Isotope Reactor (HFIR). Four samples were irradiated at temperatures from 570 °C to 1130 °C up to 0.73 dpa. Neutron spectrum significantly influenced the accumulation of transmutation products, with Re+Os content estimated at ∼0.3–0.6% under thermal-neutron shielded conditions and ∼5.2% under the mixed spectrum condition. Irradiation temperature strongly influences tungsten’s microstructure, with dislocation loops and fine voids forming at lower temperatures and only larger voids and Re/Os segregation observed at higher temperature. Under thermal-neutron shielded conditions, dislocation loops and voids were observed at 570 °C and 790 °C. At the highest irradiation temperature (1130 °C), dislocation loops were no longer observed, while larger but less dense voids remained. Re and Os segregation to void surfaces was evident at 790 °C and 1130 °C, though no precipitation was observed. In contrast, under the mixed-spectrum condition, both spherical and needle-like Re/Os-rich precipitates were observed, frequently accompanied by large voids. Dislocation loops were not observed, but loop-like contrast within the precipitates suggests they may have nucleated on pre-existing loops. Irradiation-induced hardening was assessed for the shielded samples at 570 °C and 790 °C. Dispersed barrier hardening (DBH) analysis, based on TEM-resolved defects, revealed that voids were the dominant contributors to hardening, consistent with literature results. A schematic model is proposed to describe defect and precipitate evolution in tungsten under fusion-relevant transmutation-to-dpa conditions.

Dislocation loops↗
DOE OSTI2024

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

The effect of mixed spectra and thermal neutron shielded irradiation on tungsten was evaluated with plasma exposure in the tritium plasma experiment followed by thermal desorption spectroscopy, X-ray photoelectron spectroscopy, and transmission electron microscopy. The two different irradiation campaigns were performed at the High Flux Isotope Reactor to 0.39–0.74 displacement per atom (dpa) in the 894–1379 K temperature range. A neutron spectrum influence on the void size and void number density was not observed. However, a strong correlation was found between void size and void number density with temperature, but not with dpa in the limited dpa range of this study. Thermal neutron shielding significantly reduced the transmutation to Re+Os. Higher irradiation temperature will lead to larger voids with lower number density, which reduces deuterium retention. In conclusion, grain growth was also observed for high-temperature irradiation of over ~1300 K within the limited grains visible in the transmission electron microscopy specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2014

SU-D-BRE-07: Neutron Shielding Assessment for a Compact Proton Therapy Vault

Purpose: To perform a neutron shielding assessment of a commercially available compact proton therapy system. Methods: TOPAS (TOol for PArticle Simulation) beta release was used to model beam line components for Mevion S250 proton treatment system the design of which is that the cyclotron is present in the treatment room. Three neutron production sources were taken into account in the simulation. These are the cyclotron, the treatment nozzle and the patient itself, respectively. The cyclotron was modeled as a cylindrical iron target (r =5 cm, length = 8 cm). A water phantom (10 cm ×10 cm ×60 cm) was used to model the patient and various structures (scattering foils, range modulator wheel, applicator and compensator) defaulted in TOPAS were used to model the passive scattering treatment nozzle. Neutron fluences and energy spectra were counted in a spherical scoring geometry per incident proton in 18 angular bins (10 degree each). Fluence to dose conversion factors from ICRU publication 74 were used to acquire neutron ambient dose equivalent H*(10). A point source line of sight model was then used to calculate neutron dose at eight locations beyond shielding barriers. Results: The neutron ambient dose equivalent was calculated at the 8 points of interest around the proton treatment vault. The highest dose was found to be less than 0.781 mSv/year outside south barrier wall. However, the dose is less than 0.05 mSv/year at the control room area of the proton vault. Conclusion: All Points of interest were well under annual dose limits. This suggests that the shielding design of this compact proton therapy system is sufficient for radiation protection purpose. However, it is important to note that the workload and the occupancy factors are direct multipliers for dose calculations beyond the barrier and must be accurately estimated for validation of our results.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS2024

Wear and Neutron Shielding Resilience of Titanium-Hexagonal Boron Nitride Coatings Against Extreme Lunar Radiation and Thermal Cycles

Ti6Al4V and Al6061 are aerospace alloys used in lunar structural components and rovers owing to their high specific strength. However, they deteriorate rapidly when subjected to demanding conditions of the lunar environment, such as extreme thermal cycles, solar particle radiation, and abrasive regolith. Cryo-milled powders were employed to deposit hBN-reinforced protective titanium coatings through plasma spray (atmospheric APS and vacuum-VPS) with 2 and 10 vol% of hBN on Ti6Al4V and Al6061 substrates. These coatings were evaluated for durability under extreme lunar-like conditions, including thermal cycling, electron radiation, and synergistic (electron radiation-thermal cycle) exposure. The coatings exhibited radiation-induced hardening, resulting in a substantial increase in their microhardness ~20–40% compared to virgin condition. Bright-field transmission electron microscopy (TEM) analysis reveals the presence of high dislocation densities, black dot defects, and dislocation clusters, providing evidence of the severe impact of synergistic environmental stresses. Ball-on-disk tribological tests were conducted in the presence of JSC-1 A lunar regolith simulant to evaluate the wear performance of coatings. Compared to conventional Ti6Al4V substrate, 50%, 90%, and 70% reductions in wear volume were observed in the Ti/2 vol% hBN coatings in thermal-only, radiation-only, and synergistic environments. Neutron shielding tests indicate that the synergistic coatings offer higher neutron shielding performance by up to 28% compared to virgin coatings, attributed to improved neutron capture by 10B isotopes. Upon a comprehensive assessment and ranking of multiple coatings under varying conditions, the VPS Ti/2 vol% hBN coating emerges as the top choice for protection against extreme radiation and thermal cycles for future lunar explorations.

Boron nitride↗
DOE OSTI2012

Ageing of a neutron shielding used in transport/storage casks

In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

36 MATERIALS SCIENCE↗
DOE OSTI2019

PHENIX U.S.-Japan Collaboration Investigation of Thermal and Mechanical Properties of Thermal Neutron–Shielded Irradiated Tungsten

The United States and Japan have collaborated on fusion materials research in a series of agreements reaching back to 1981. The PHENIX collaboration is the latest U.S.-Japan project which spans 2013 to 2019 and has the goal of assessing technical feasibility of tungsten-based, helium-cooled plasma-facing component concepts for a demonstration fusion power reactor (DEMO). Task 2 within the PHENIX project is focused on evaluating the neutron irradiation effects in tungsten. For tungsten, the transmutation to Re and Os is at least as important to determining its properties after irradiation as the displacement damage, and the transmutation rate depends on the energy spectrum of the reactor. A large-scale, instrumented irradiation capsule with thermal neutron shielding to better mimic fusion conditions was irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. The tungsten specimens were irradiated in different temperature zones between 500°C and 1200°C to doses of ~0.2 to 0.7 displacements per atom. More than 20 varieties of pure tungsten and tungsten alloys were included in the irradiation, and they were evaluated in the 3025E hot-cell facility and at the Low Activation Materials Development and Analysis Laboratory. The elevated temperature tensile, fracture toughness, hardness, thermal conductivity, electrical resistivity, density, elemental composition, and microstructure properties of the irradiated materials are being collected. Furthermore, this paper overviews the experimental design, specimen matrix, and the initial results of postirradiation examinations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

SU-E-T-569: Neutron Shielding Calculation Using Analytical and Multi-Monte Carlo Method for Proton Therapy Facility

Purpose: To evaluate the shielding wall design to protect patients, staff and member of the general public for secondary neutron using a simply analytic solution, multi-Monte Carlo code MCNPX, ANISN and FLUKA. Methods: An analytical and multi-Monte Carlo method were calculated for proton facility (Sumitomo Heavy Industry Ltd.) at Samsung Medical Center in Korea. The NCRP-144 analytical evaluation methods, which produced conservative estimates on the dose equivalent values for the shielding, were used for analytical evaluations. Then, the radiation transport was simulated with the multi-Monte Carlo code. The neutron dose at evaluation point is got by the value using the production of the simulation value and the neutron dose coefficient introduced in ICRP-74. Results: The evaluation points of accelerator control room and control room entrance are mainly influenced by the point of the proton beam loss. So the neutron dose equivalent of accelerator control room for evaluation point is 0.651, 1.530, 0.912, 0.943 mSv/yr and the entrance of cyclotron room is 0.465, 0.790, 0.522, 0.453 mSv/yr with calculation by the method of NCRP-144 formalism, ANISN, FLUKA and MCNP, respectively. The most of Result of MCNPX and FLUKA using the complicated geometry showed smaller values than Result of ANISN. Conclusion: The neutron shielding for a proton therapy facility has been evaluated by the analytic model and multi-Monte Carlo methods. We confirmed that the setting of shielding was located in well accessible area to people when the proton facility is operated.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Lightweight concrete with enhanced neutron shielding

A lightweight concrete containing polyethylene terephthalate in an amount of 20% by total volume. The concrete is enriched with hydrogen and is therefore highly effective at thermalizing neutrons. The concrete can be used independently or as a component of an advanced neutron radiation shielding system.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Neutron shielding panels for reactor pressure vessels

In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Uncertainty Quantification for Neutron Shield Using Convolutional Neural Networks

Uncertainty quantification from radiation transport calculations was conducted using a Bayesian inference approach. A surrogate model, using a convolutional neural network, was employed to emulate the neutron fluence, which was simulated with a Monte Carlo radiation transport model. This allowed for a computationally cheap approach to evaluate input parameters and to sample their corresponding posterior probability distributions. Experimental data from the literature were employed to perform uncertainty quantification studies for concrete shields. As a result, the method is a nonintrusive approach that enables studies with multiple input parameters and can be applied to any radiation transport model.

Bayesian inference↗
DOE OSTI2017

Nuclear reactor neutron shielding

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗