• Nuclear Engineering and Technology
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• 제53권1호
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• pp.282-293
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• 2021

This paper examines gamma-ray shielding properties of SBC-Bx glass system with the chemical composition of 40SiO₂-10B₂O₃-xBaO-(45-x)CaO- yZnO- zMgO (where x = 0, 10, 20, 30, and 35 mol% and y = z = 6 mol%). Mass attenuation coefficient (µ/ρ) which is an essential parameter to study gamma-ray shielding properties was obtained in the photon energy range of 0.015-15 MeV using PHITS Monte Carlo code for the proposed glasses. The obtained results were compared with those calculated by WinXCOM program. Both the values of PHITS code and WinXCOM program were observed in very good agreement. The (µ/ρ values were then used to derive mean free path (MFP), electron density (N_eff), effective atomic number (Z_eff), and half value layer (HVL) for all the glasses involved. Additionally, G-P method was employed to estimate exposure buildup factor (EBF) for each glass in the energy range of 0.015-15 MeV up to penetration depths of 40 mfp. The results reveal that gamma-ray shielding effectiveness of the SBC-Bx glasses evolves with increasing BaO content in the glass sample. Such that SBC-B35 glass has superior shielding capacity against gamma-rays among the studied glasses. Gamma-ray shielding properties of SBC-B35 glass were compared with different conventional shielding materials, commercial glasses, and newly developed HMO glasse. Therefore, the investigated glasses have potential uses in gamma shielding applications.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.344-349
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• 2016

Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient's body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.

• Journal of Radiation Protection and Research
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• 제42권1호
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• pp.16-20
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• 2017

Background: The design of deuteron accelerator neutron source facilities requires reliable yield estimation of gamma-rays as well as neutrons from deuteron-induced reactions. We have so foar measured systematically double-differential thick target neutron yields (DDTTNYs) for carbon, aluminum, titanium, copper, niobium, and SUS304 targets. In the neutron data analysis, the events of gamma-rays taken simultaneously were treated as backgrounds. In the present work, we have re-analyzed the experimental data for a thick carbon target with particular attention to gamma-ray events. Materials and Methods: Double-differential thick target gamma-ray yields from carbon irradiated by 5 and 9 MeV deuterons were measured using an NE213 liquid organic scintillator at the Kyushu University Tandem accelerator Laboratory. The gamma-ray energy spectra were obtained by an unfolding method using FORIST code. The response functions of the NE213 detector were calculated by EGS5 incorporated in PHITS code. Results and Discussion: The measured gamma-ray spectra show some pronounced peaks corresponding to gamma-ray transitions between discrete levels in residual nuclei, and the measured angular distributions are almost isotropic for both the incident energies. Conclusion: PHITS calculations using INCL, GEM, and EBITEM models reproduce the spectral shapes and the angular distributions generally well, although they underestimate the absolute gamma-ray yields by about 20%.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.301-308
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• 2024

Ceramic materials are being explored as alternatives to toxic lead sheets for radiation shielding due to their favorable properties like durability, thermal stability, and aesthetic appeal. However, crafting effective ceramics for radiation shielding entails complex processes, raising production costs. To investigate local viability, this study evaluated Malaysian ceramic tiles for shielding in diagnostic X-ray rooms. Different ceramics in terms of density and thickness were selected from local manufacturers. Energy Dispersive X-ray Fluorescence (EDXRF) and X-ray Fluorescence (XRF) characterized ceramic compositions, while Monte Carlo Particle and Heavy Ion Transport code System (MC PHITS) simulations determined Linear Attenuation Coefficient (LAC), Half-value Layer (HVL), Mass Attenuation Coefficient (MAC), and Mean Free Path (MFP) within the 40-150 kV energy range. Comparative analysis between MC PHITS simulations and real setups was conducted. The C3-S9 ceramic sample, known for homogeneous full-color structure, showcased superior shielding attributes, attributed to its high density and iron content. Notably, energy levels considerably impacted radiation penetration. Overall, C3-S9 demonstrated strong shielding performance, underlining Malaysia's potential ceramic tile resources for X-ray room radiation shielding.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4220-4225
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• 2022

Monte Carlo (MC) simulations are increasingly being used as an alternative or supplement to the gamma spectrometric method in determining the full energy peak efficiency (FEPE) necessary for radionuclide identification and quantification. The MC method is more advantageous than the experimental method in terms of both cost and time. Experimental calibration with standard sources is difficult, especially for specimens with unusually shaped geometries. However, with MC, efficiency values can be obtained by modeling the geometry as desired without using any calibration source. Modeling the detector with the correct parameters is critical in the MC method. These parameters given to the user by the manufacturer are especially the dimensions of the crystal and its front edge, the thickness of the dead layer, dimensions, and materials of the detector components. This study aimed to investigate the effect of the front edge geometry of the detector crystal on efficiency, so the effect of rounded and sharp modeled front edges on the FEPE was investigated for <300 keV with three different HPGe detectors in point and volume source geometries using PHITS MC code. All results showed that the crystal should be modeled as a rounded edge, especially for gamma-ray energies below 100 keV.

• Journal of Radiation Protection and Research
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• 제45권3호
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• pp.130-141
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• 2020

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd₂O₃, B₄C, polyethylene (PE), and borated (5 wt% ^natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B₄C and Gd₂O₃ were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B₄C or Gd₂O₃ behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd₂O₃+10 cm-thick PE or 0.1 cm-thick B₄C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B₄C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.

• Journal of Radiation Protection and Research
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• 제38권1호
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• pp.22-28
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• 2013

Skyshine은 고에너지 가속기의 차폐 설계 시 반드시 고려되는 중요한 현상이다. 본 연구에서는 고에너지 전자 가속기에 대한 중성자 skyshine 평가 방법을 새롭게 제시하였으며 기존의 방법들과 비교하여 타당성을 검증하였다. 고에너지 전자가속기로부터 원거리 지역의 방사선량을 계산하기 위해 몬테카를로 코드, FLUKA와 PHITS를 이용하였다. 가속기 건물로부터 원거리 지역까지 도달하는 방사선장의 경로를 skyshine, direct, groundshine, multiple-shine으로 분류하였다. 분류된 각각의 성분이 총 유효선량에 기여하는 정도를 평가하였다. 방사선원 계산에는 10 GeV 전자가 두꺼운 표적에 입사하여 생성되는 중성자를 고려하였다. Groundshine 효과를 평가하기위해 PAL-XFEL의 건설부지토양에 대한 성분을 고려하였다. 가속기 건물로부터 비교적 가까운 50 m 미만에서는 direct와 groundshine 성분들이 총 유효선량에 대부분 기여하였다. 가속기 터널로부터 거리가 멀어질수록 skyshine 성분의 기여도가 증가하였다. 평가된 skyshine 성분에 대한 유효선량은 기존에 skyshine 선량을 계산할 때 사용되었던 반실험식 중 전자가속기를 이용한 실험결과에 기반한 Rindi의 식과 가장 잘 일치하였다. 간이계산코드 SHINE3의 결과와 20% 이내로 일치하였다. 모든 성분이 포함된 총 유효선량은 기존의 평가방식에 비해 10배 정도 크게 평가되었다. 가속기 터널로부터 원거리 지역에 대한 선량평가계산에서 mutiple-shine 성분의 영향은 skyshine 성분보다 더 크다는 것을 확인하였다.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3448-3451
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• 2022

Biochar is a novel carbon based material derived from waste that shows promising properties for several applications. In this paper we investigate its potential use as a low cost, greener alternative to commonly used aggregates employed to enhance the neutron shielding performance of concrete. Monte Carlo simulations are performed with the PHITS code to estimate the neutron attenuation of blank and biochar-reinforced concrete exposed to high energy neutrons. We find that the shielding performance of concrete with 15% biochar is comparable with commonly used materials such as Boron Carbide at 20% and exceeds that of Basalt fibers with the same concentration, making these composites an interesting greener alternative to current solutions. A combination of biochar and heavier fillers also show extremely promising performance.

• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.123-131
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• 2016

Background: With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. Materials and Methods: The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and $290MeV{\cdot}n^{-1}$ oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. Results and Discussion: It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. Conclusion: The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.389-394
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• 2016

Background: Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. Materials and Methods: TET2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. Results and Discussion: To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. Conclusion: In the present study, we have developed a computer program, TET2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.