Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “SHIELDING”

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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Investigating the effect of different shielding gas mixtures on microstructure and mechanical properties of 410 stainless steel fabricated via large scale additive manufacturing

Metal Big Area Additive Manufacturing (mBAAM) offers the potential to fabricate large scale tools at high deposition rates (15 lb/h+). 410 martensitic steel is a potential tooling material, owing to its low cost, good machinability and reasonable printability. During the mBAAM process, the shielding gas can have a significant impact on the material properties as well as the process cost. Therefore, the current study aims to understand the effect of different shielding gas mixtures on large-scale additive manufacturing of 410 martensitic stainless steel. We show that an argon mixture with 3% nitrogen gas produced the best performance in terms of maximum hardness and tensile strength, with much less scatter in tensile strength. He-Ar-CO 2 or tri-mix shielded samples showed a low tensile strength with wide scatter, due to stabilized delta ferrite in microstructure during printing. Both tri-mix and Ar-CO 2 shielded samples showed slightly higher porosity. Thus, we recommend the use of argon-3% nitrogen as a shielding gas mixture for processing 410 steel for tool applications, based on the relatively low cost of this gas mixture and the resulting higher hardness, higher dimensional stability, and lower porosity.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Copper-nickel electroplating of 3D-printed acrylonitrile butadiene styrene for interference and radiation shielding applications

Here, in this study, electromagnetic and radiation shielding structures were fabricated by 3D printing acrylonitrile butadiene styrene (ABS) structures via fused filament fabrication and electroplating with copper (Cu) and nickel (Ni) metals. Prior to electroplating, the printed ABS materials were made conductive by surface electrodeposition of polypyrrole (PPy). ABS adsorbs the pyrrole monomer, which subsequently polymerizes via oxidation forming conductive PPy layer on the printed structures. The successful anchoring of PPy layer can be attributed to its hydrogen bonding and electrostatic interactions with ABS. Thermal analysis, optical, and electron microscopy equipped with energy dispersive X-ray were employed to characterize the printed shielding structures, while the interference and radiation shielding effectiveness were evaluated using a spectrum analyzer and dosimeter, respectively. The interference shielding was found to be effective at 60 ± 5 dB over relatively small frequency range. Shielding was also apparent against gamma and X-ray ionizing rays, but more significantly effective against beta rays. Furthermore, the Cu/Ni coating was able to reduce any heat-induced dimensional changes for the 3D printed ABS substrate without compromising the mechanical integrity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Printed graphene and its composite with copper for electromagnetic interference shielding applications

Abstract Advances in mobile electronics and telecommunication systems along with 5G technologies have been escalating the electromagnetic interference (EMI) problem in recent years. Graphene-based material systems such as pristine graphene, graphene-polymer composites and other graphene-containing candidates have been shown to provide adequate EMI shielding performance. Besides achieving the needed shielding effectiveness (SE), the method of applying the candidate shielding material onto the object in need of protection is of enormous importance due to considerations of ease of application, reduced logistics and infrastructure, rapid prototyping and throughput, versatility to handle both rigid and flexible substrates and cost. Printing readily meets all these criteria and here we demonstrate plasma jet printing of thin films of graphene and its composite with copper to meet the EMI shielding needs. SE over 30 dB is achieved, which represents blocking over 99.9% of the incoming radiation. Graphene and its composite with copper yield higher green index compared to pure copper shields, implying reduced reflection of incoming electromagnetic waves to help reduce secondary pollution.

Materials Science↗
DOE OSTI2023

Overview of Electrically Conductive and Active Shielding for Wireless Power Transfer with a Polyphase Wireless Electric Vehicle Charging Study

The development and deployment of wireless power transfer (WPT) is increasing across industries including applications in healthcare, sensing and telecommunications, and electric vehicle charging due to its increased mobility and convenience. An important aspect of wireless power transfer is shielding to limit the intensity of electromagnetic fields (EMF) within external regions to protect users and nearby technology. Within this paper, the classifications and recent developments in shielding technologies are reviewed with discussion on potential benefits and drawbacks towards application for wireless charging of electric vehicles (EV). Additionally, a design study is proposed and simulated in 3D FEA for a novel active shield to substantially reduce magnetic field emissions of a high-power, high-frequency rotating-field 3-phase electromagnetic coupler for quickly charging electric vehicles. The active shield, designed to reduce Z-axis emissions, was found to be highly effective, reducing maximum EMF emissions by 83% in aligned static operation and by 50% with lateral misalignment. Active shielding was also found to strongly mitigate the impact of lateral misalignment at each studied partial alignment.

Lewis, Donovin D.↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

E-PiPEline: Quick to Market Face Coverings and Face Shields Using Commonly Available Materials

The Center for Disease Control has recommended that the public should wear cloth face coverings in public settings. Face coverings and face shields can be made by using Commonly Available Materials (CAMs). As part of the Sandia COVID-19 LDRD effort (funded under the Materials Science Investment Area), the Sandia E-PiPEline task evaluated design options for face coverings and face shields considering their effectiveness, durability, build difficulty, build cost, and comfort. Observations from this investigation are presented here to provide guidelines for home construction of face coverings and face shields. This executive summary includes a brief roadmap of the analysis methodology, two one-page handouts geared to be distributed to the public at large (one for face coverings and one for face shields), and additional observations regarding potential solutions for face coverings and face shields included to further support the one-page handouts.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Second Target Station (STS) Project: Analysis of Bunker Shielding–Preliminary Design

At the Department of Energy’s Oak Ridge National Laboratory, the Second Target Station (STS) beamline sources for preliminary design have been used to perform a shielding analysis of the bunker. The simulations in this report used the ST11 beamline source, primarily the ST11 neutron source, because that source provides conservative prompt total dose rate estimates for all beamlines. Prompt total dose rates (i.e., neutron plus photon dose rates when the beam is on) were calculated on top of the bunker roof, outside the bunker wall, and in the basement below the bunker and instrument floor. Design changes were made to ensure the prompt total dose rate in these areas does not exceed 0.25 mrem/h. This report documents the required shielding thicknesses to meet this dose rate limit. There is one exception to this: for a combination of open and closed beamlines, the prompt total dose rate outside the bunker across from the closed beamline, where there is less shielding due to the lack of beamline shielding, slightly exceeds 0.25 mrem/h. A future analysis will document the shielding modifications required to reduce the calculated prompt total dose rates for this configuration to less than 0.25 mrem/h once more details regarding the STS high density concrete density and composition are known. Finally, prompt total dose rates inside the bunker were simulated. The results of this simulation show that the majority of prompt total dose rate in the bunker is due to neutrons, and a few percent of the total is caused by secondary photons created by the neutrons. The contribution to prompt total dose rate inside the bunker by the primary photons that enter each beamline from the target/moderator area is at least two orders of magnitude less than the neutron contribution. This means prompt total dose rate contributions due to primary photons inside and outside the bunker can be neglected.

42 ENGINEERING↗
DOE OSTI2022

SCALE 6.2.4 Validation: Radiation Shielding

For safe and reliable use of computer codes by the community, accuracy must be clearly evaluated. In particular, the nuclear reactor engineering and licensing field needs accurate tools for radiation shielding modeling. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is one such tool, with built-in variance reduction methods distributed within the SCALE code, and its validity is demonstrated in this report for the released version 6.2.4. Representative benchmarks corresponding to shielding analysis are selected for the validation study. Typical experimental results analyzed from those benchmarks include neutron fluxes, detector count rates, detector energy response functions, neutron and gamma doses, foil neutron activation rates and activities, neutron leakage fluxes, and skyshine dose rates. Thousands of points of comparison between experiment and calculation are presented in this work. Other than rare outliers typically explained by either a lack of information or large uncertainties in the experiment conditions, material, or dimensions, MAVRIC agrees well with the experiment results. MAVRIC is also compared to Monte Carlo N-Particle (MCNP) calculations when available, and both codes generally produce good agreements within estimated uncertainties. The selected benchmarks are obtained from reliable sources such as the International Criticality Safety Benchmark Evaluation Project Handbook (ICSBEP Handbook), the Shielding Integral Benchmark Archive & Database (SINBAD), and other shielding validation work found in the literature. Additional datapoints and benchmarks will be added to future versions of this report to incrementally expand the shielding validation suite incrementally.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2026

Towards Commercialization of Hydrophone Flow Shields to Improve Passive Acoustic Data Quality (CRADA 678) (Final Report)

This project performed testing, market research, and manufacturing refinement to ready patent-pending hydrophone flow shields developed at PNNL for commercialization. The flow shields reduce the impact of flow noise on underwater sound recordings in areas of strong flow, like tidal channels or rivers. Flow noise is a challenge in many applications where it is necessary to measure underwater sound, including measuring the sound produced by marine energy converters and monitoring marine mammals. This project addressed several key questions about flow shield performance that are high-priority for potential customers and refined existing manufacturing processes to be scalable for market-level production. Specifically, this project conducted both laboratory and field evaluations of the flow shields to comprehensively evaluate their performance. Overall, the results indicate that the flow shields reduce flow noise by over 20 dB at frequencies below 50 Hz at flow speeds greater than 0.5 m/s and are proven to survive deployments more than 6 months long.

42 ENGINEERING↗
DOE OSTI2023

Overview on Shielding Analyses for the VENUS Instrument at SNS

VENUS, a world‐class versatile neutron imaging instrument, is under construction and is expected to be completed and ready to start commissioning in 2023. The range of cold to epithermal neutrons at SNS will give users of VENUS access to novel imaging methods, as well as to significantly improved existing methods. The instrument is being built on beam line 10 at Spallation Neutron Source (SNS) First Target Station (FTS) facing a decoupled poisoned hydrogen moderator. Instrument design, which includes optics, Front-End components and instrument enclosure components and shape started over 10 years ago and during this time had significant changes. All changes were supported by neutronics analyses to provide adequate shielding for both Front-End and instrument enclosure.Final optics design will provide the field of view (FOV) at the detector position (where the image is formed) to be as high as 0.20 by 0.20 m. Both the beam, which contains a large fraction of high-energy neutrons, and the desire for a large footprint on the detector are challenges for shielding design and budget, because the driving cost for the instrument is the beam line and enclosure shielding. For cost reason VENUS baseline design is optimized with respect to both the instrument cave footprint and its wall thickness, and Front-End shielding is tailored along the beam. Analyses are performed with the Monte Carlo particle transport code MCNPX version 2.7.0 to make choices on materials, thicknesses and configurations of the shielding. Numerous calculations are performed to meet space and coast constrain and to satisfy instrument physics needs and to comply with radiation protection requirements.

Gallmeier, Franz X.↗
DOE OSTI2023

Thermal Analysis of Thermal Shield for PIP-II SSR1 Cryomodule

Fermilab's PIP-II project plans to enhance the onsite accelerator into the world's most intense neutrino beam, bringing Fermilab into the next generation of particle accelerators. The five different types of cryomodules are vital parts of the PIP-II project and each requires numerous engineering analyses to prove operation. For example, the thermal shield of the Single Spoke Resonator-1 (SSR1) cryomodule required a thermal analysis to ensure the design was operating as intended before fabrication commenced. The thermal shield is cooled by the supercritical helium inside of the extrusion and must remain between 45 and 80 Kelvin. The main purpose of the thermal shield is to intercept different high heat loads and thermal radiation and prevent them from reaching the more sensitive internal components. Finite Element Analysis (FEM) was used to determine temperature estimates of the thermal shield based on the location of a variety of heat sources. Results determined temperatures exceeding the design limit around the crucial component of the current leads. Due to superconducting properties, the areas around the current leads cannot surpass 68 Kelvin. Different solutions to the critical high-temperature areas will be proposed. These solutions range from changing the properties of the helium flow to designing an attachment for the thermal shield. Each proposition works as a viable solution, however, each proposition also has its own set of advantages and disadvantages that are described for further design consideration.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Residual Dose Rates Analyses from Vent Lines Shield Block on the Target Cart at Spallation Neutron Source

The target carriage supports the mercury target vessel in the Spallation Neutron Source (SNS) in a cantilevered fashion and supplies the liquid mercury flowing to and from the target vessel. The target carriage is designed to move the target vessel into its target monolith position to receive the proton beam and then remove it horizontally back into an adjacent service cell, where the target vessel can be exchanged about 2-3 times per calendar year. At the time of the performed studies, the SNS mercury target can accept a proton beam at 1 GeV energy and 1.4 MW power; however, it will be upgraded to receive 1.3 GeV energy at 2 MW by the end of 2024 year [1]. The cart is practically a shield block on wheels to minimize fluxes and dose rates in the service cell downstream of the carriage. There are vent lines located inside a passage on top of the target carriage, and they are covered by a stepping-up steel shield block (vent lines shield block). Vent lines are permanently attached to the shield block. New vent lines with a different design are going to be installed. These lines are shielded by the new steel block to which the vent lines are attached. The estimate of the residual dose rates from the block after it is extracted from the target carriage is performed to estimate the radiation fields and to decide on the type of storage container to be used.

Popova, Irina I↗
DOE OSTI2022

Particulate Fuel Modeling of MC 2 -3 using Iterative Local Spatial Self-shielding Method

We report a new spatial self-shielding method for particulate fuels has been developed based on disadvantage factors and implemented in the MC 2 -3 code. This method named the iterative local spatial self-shielding (ILSS) method considers the shadowing effect of randomly distributed particles on spatial self-shielding in particles through a homogenized composition region added outside the particle of interest at the center. The self-shielded cross sections of the central particle are determined iteratively since they are used in determining the cross sections of the homogenized composition region. The ILSS method was verified for infinite stochastic medium problems of single and multiple types of particles, VHTR unit cell problems, and HTTR assembly problems. The verification test results show that the ILSS method accurately predicts the stochastic particle shadowing effect and reaction rates in particles, whereas the regular array model and the stochastic collision probability method underpredict the particle shadowing effect and overestimate reaction rates in particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Modeling a generic TRISO-fueled heat pipe microreactor using SCALE: Depletion, transportation criticality, and shielding

This paper demonstrates the applicability of the SCALE code system to tristructural-isotropic (TRISO)-fueled heat pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study conducted supports US Nuclear Regulatory Commission code readiness efforts for advanced non–light-water reactor technologies and is intended as a code capability demonstration rather than as an optimization of a specific microreactor design. The modeled reactor employs high-assay low-enriched uranium (HALEU) uranium oxycarbide (UCO) TRISO fuel and beryllium oxide (BeO) reflectors and operates at 7.5 MWth with a nominal lifetime of about 3 effective full power years. Representative cases for fresh and irradiated cores were selected to exercise SCALE methods relevant to reactor operation and post-irradiation transport. The discharged-core decay heat is approximately 6% of operating power immediately after shutdown. Transportation criticality calculations show that internal water ingress is the dominant reactivity effect, with fully flooded fresh core and irradiated core configurations remain above the subcriticality criterion, even with the available control mechanisms. Shielding calculations for a simplified transportation package indicate that normal-condition dose rates are governed mainly by shielding thickness and cooling time, whereas the breached hypothetical accident case is governed primarily by cooling time. Overall, the study shows that SCALE supports depletion, transportation criticality, and shielding evaluations efficiently for TRISO-fueled heat pipe microreactors within a single code system.

Criticality↗
DOE OSTI2025

Effects of shielding gas and spot reduction on humping suppression in high-speed laser welding

High-speed laser welding is a crucial process to enhance the production rate of fuel cell fabrication. However, exceeding a critical welding speed leads to defect formation such as humps and root concavities, which degrade weld quality and mechanical performance. This study employed shielding gas and spot reduction approaches to suppress these defects. The relationships between individual defect types and their influence on weld strength were systematically examined. Moreover, in situ high-speed synchrotron X-ray imaging was employed to comprehend the mechanisms of humping mitigation by shielding gas through analysis of instantaneous keyhole and melt pool dynamics. The observations reveal that shielding gas mitigates humping by shortening the molten pool through enhanced surface heat convection and by altering surface tension to prevent melt flow collisions. The effects of shielding gas and spot reduction were further incorporated into the dimensionless humping index (π h ) that describes the humping formation tendency. These findings provide mechanistic insights and quantitative guidance for defect control strategies in high-speed laser welding.

08 HYDROGEN↗
DOE OSTI2020

Radiation shielding analysis of the concrete hutch of a long beamline of the Advanced Photon Source Upgrade project

The radiation shielding analysis of the external concrete enclosure of a feature long beamline of the Advanced Photon Source Upgrade project is described in this work. The beamline has 2 undulators that can operate in two different configurations and the source terms are estimated for the worst configuration using the SRW code in the OASYS package as well as the STAC8 code. The shielding analysis is carried out using the STAC8 code, and for the case of pink beam, with the FLUKA code also. Effective dose rates on contact with the outside surface of the concrete walls for pink, reflected full spectrum and monochromatic beams are calculated. For the monochromatic beams, 2 sets of discrete bandwidths (BW) from the Double Crystal Monochromator (DCM) and the full reflectance of the Double Multilayer Monochromator (DMM) are used. Further, due to the size of the direct beam and the energy domain of interest in the direct and scattered fields, the importance of the equivalent dose to the lens of the eye is discussed with respect to the shielding calculations of such facilities. The dose rates outside the concrete hutch in the lateral directions calculated using STAC8 is about 50% higher than the FLUKA results for the case of pink beam but may be considered as a reasonable agreement due to the differences in the calculational methodologies. Those dose rates in the lateral and forward directions are higher for the pink beam to allow full occupancy from the radiation protection point of view. For the full spectrum after 3 reflections, the effective dose rates are less than 0.5 Sv/h in the lateral direction but is higher in the forward region below about 22° with respect to the beam direction. For the DCM beams, except for the scenario with one mirror and 0.1%BW, all dose rates are less than 0.5 Sv/h outside the concrete hutch in the lateral directions. Here, the higher harmonics are what contributes to the dose outside the shield and the bandwidths assumed to derive the photon flux thus becomes important. The dose rates with the DMM are generally lower compared to the 0.1% BW DCM beam but are higher than the results obtained with XOP calculated bandwidths.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

An in-silico NMR laboratory for nuclear magnetic shieldings computed via finite fields: Exploring nucleus-specific renormalizations of MP2 and MP3

Here we developed and implemented a method-independent, fully numerical, finite difference approach to calculating nuclear magnetic resonance shieldings, using gauge-including atomic orbitals. The resulting capability can be used to explore non-standard methods, given only the energy as a function of finite-applied magnetic fields and nuclear spins. For example, standard second-order Møller-Plesset theory (MP2) has well-known efficacy for 1H and 13C shieldings and known limitations for other nuclei such as 15N and 17O. It is, therefore, interesting to seek methods that offer good accuracy for 15N and 17O shieldings without greatly increased compute costs, as well as exploring whether such methods can further improve 1H and 13C shieldings. Using a small molecule test set of 28 species, we assessed two alternatives: κ regularized MP2 (κ-MP2), which provides energy-dependent damping of large amplitudes, and MP2.X, which includes a variable fraction, X, of third-order correlation (MP3). The aug-cc-pVTZ basis was used, and coupled cluster with singles and doubles and perturbative triples [CCSD(T)] results were taken as reference values. Our κ-MP2 results reveal significant improvements over MP2 for 13C and 15N, with the optimal κ value being element-specific. κ-MP2 with κ = 2 offers a 30% rms error reduction over MP2. For 15N, κ-MP2 with κ = 1.1 provides a 90% error reduction vs MP2 and a 60% error reduction vs CCSD. On the other hand, MP2.X with a scaling factor of 0.6 outperformed CCSD for all heavy nuclei. These results can be understood as providing renormalization of doubles amplitudes to partially account for neglected triple and higher substitutions and offer promising opportunities for future applications.

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