• Journal of Radiation Industry
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• v.17 no.1
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• pp.93-100
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• 2023

This study analyzes the neutron shielding performance of Soft Magnetic Material and proposes a military application. In general, the military protection facility has been constructed with thick concrete, so Soft Magnetic Material, consisting of boron, was used with concrete in this study. To do so, Monte-Carlo N-Particle (MCNP) was applied to simulate the Watt-fission neutron spectrum of ²³⁵U and ²³⁹Pu. As a result, a configuration of polyethylene and Soft Magnetic Material is evaluated about four times better than borated polyethylene concerning the atomic weight of boron inside each shielding material. Also, when a nuclear weapon explosion is simulated in MCNP, 1 mm of Soft Magnetic Material with 20 cm of concrete shows about 55% more additional neutron shielding performance compared to when Soft Magnetic Material is not used. In this work, the neutron shielding performance of Soft Magnetic Material could be identified and Soft Magnetic Material would be useful for neutron shielding if applicable to concrete structure.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2153-2162
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• 2024

The purpose of this study is to develop a high-energy neutron shielding material applied in proton therapy environment. Composite shielding material consisting of 10.00 wt% boron carbide particles (B₄C), 13.64 wt% surface-modified cross-linked polyethylene (PE), and 76.36 wt% tungsten particles were fabricated by hot-pressure sintering method, where the optimal ratio of the composite is determined by the shielding effect under the neutron field generated in typical proton therapy environment. The results of Differential Scanning Calorimetry measurements (DSC) and tensile experiment show that the composite has good thermal and mechanical properties. In addition, the high energy-neutron shielding performance of the developed material was evaluated using cyclotron proton accelerator with 100 MeV proton. The simulation shows a 99.99% decrease in fast neutron injection after 44 cm shielding, and the experiment result show a 99.70% decrease. Finally, the shielding effect of replacing part of the shielding material of the proton therapy hall with the developed material was simulated, and the results showed that the total neutron injection decreased to 0.99‰ and the neutron dose reduced to 1.10‰ before the enhanced shielding. In summary, the developed material is expected to serve as a shielding enhancement material in the proton therapy environment.

• Journal of Radiation Protection and Research
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• v.43 no.1
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• pp.20-28
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• 2018

Background: Heat energy produced in nuclear reactors and nuclear fuel cycle facilities interactions modifies the physical properties of the shielding materials containing water content. Therefore, in the present paper, effect of the heat on shielding effectiveness of the concretes is investigated for gamma and neutron. The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors. Materials and Methods: The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors of ordinary and heavy concretes were investigated using NIST data of XCOM program and Geometric Progression method. Results and Discussion: The improvement in shielding effectiveness for photon and reduction in fast neutron for ordinary concrete was observed. The change in the neutron shielding effectiveness was insignificant. Conclusion: The present investigation on interaction of gamma and neutron radiation would be very useful for assessment of shielding efficiency of the concrete used in high temperature applications such as reactors.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2017.11a
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• pp.155-156
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• 2017

Borosilicate glass was incorporated to improve the neutron shielding capability of concrete. Boron is a typical neutron shielding material, and it is contained in borosilicate glass. However, borosilicate glass causes alkali-silica reaction, which damages the concrete. Therefore, studied to reduce the expansion due to alkali-silica reaction and to improve the neuton shielding capability. The measurement of the expansion due to the alkali-silica reaction was based on ASTM C 1260. Experimental results show that the expansion due to alkali-silica reaction is reduced when borosilicate glass powder incorporated. In addition, the neutron shielding capability was significantly improved when the fine aggregate replaced with borosilicate glass.

• Journal of the Korean Recycled Construction Resources Institute
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• v.4 no.3
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• pp.269-275
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• 2016

In this study, the mechanical properties and the neutron shielding rate of concrete with the borosilicate glass and the amorphous boron steel fiber were investigated. The measures of this investigation includes air contents, slump loss, compressive strength, static modulus of elasticity, compressive toughness, flexural strength, flexure toughness and neutron shielding rate. As a result, the neutron shielding rate of the concrete with borosilicate glasses increased even though the compressive strength and flexural strength decreased in comparison with that of plain concrete. Also, the mechanical toughness and the neutron shielding rate of the concrete with amorphous boron steel fiber increased in comparison with that of plain concrete.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.647-653
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• 2020

Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

• Korean Journal of Materials Research
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• v.9 no.4
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• pp.404-408
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• 1999

Effects of heating time and radiation under high temperature on the properties of thermal, mechanical and shielding ability of modified (KNS-101), hydrogenated bisphenol-A(KNS-201) type epoxy resin and phenol-novolac(KNS-301) type epoxy resin based neutron shielding materials that are used for shipping casks for radioactive material have been investigated. At early stages, the offset temperatures of KNS-101, KNS-201 and KNS-301 increased with the heating time under high temperature, but it was rarely affected by the heating time in the later stages. In addition, the thermal conductivities of KNS-101 and KNS-201 decreased with heating time, but that of KNS-301 increased with the heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials decreased with heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials of KNS-101 and KNS-301 increased with heating time, but those of KNS-201 decreased with heating time. The shielding ability of neutron shielding materials slightly increased with the radiation dose, and shielding abilities of shielding materials of KNS-101 and KNS-201 were affected to a more extent than that of KNS-301 by radiation dose under high temperature.

• Journal of the Korean Society of Radiology
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• v.16 no.5
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• pp.527-536
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• 2022

The radiation shielding characteristic of neutron shielding material has been studied as the preliminary study in order to design cosmic-ray shielding material. Specially, Soft Magnetic Material, known to be effective in EMP and radiation shielding, has been investigated to check if the material would be applicable to cosmic-ray shielding. In this work, thermal neutron shielding experiment was conducted and the Monte Carlo N-Particle(MCNP) was applied to employ skymap.dat, which is cosmic-ray data embedded in MCNP. As a result, polyethylene, borated polyethylene, and carbon nano tube, containing carbon or hydrogen, have been found to be effective in reduction of neutron flux below 20 MeV (including thermal, epithermal, evaporation). In contrast, the materials composed of iron such as SS316 and Soft Magnetic Material show a good shielding performance in the cascade energy range (above 20 MeV). Since Soft Magnetic Material is consisting of 13% of boron, it can also decrease thermal neutron flux, so it is expected that it would show a significant reduction on the entire range of neutron energy if the Soft Magnetic Material is used with hydrogen and carbon, so called low Z material.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.