• Journal of Radiation Protection and Research
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• 제26권2호
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• pp.67-71
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• 2001

극저준위 방사능측정시스템의 백그라운드에 영향을 주는 중성자를 차폐하기 위한 차폐체를 설계하였다. 중성자 차폐방법은 고 밀도 폴리에틸렌을 이용하여 고속중성자를 감속한 후 $B_4C$를 이용하여 감속된 열중성자를 흡수하는 방법을 이용하였다. 몬테카를로 모사방법인 MCNP4B 코드를 이용하여 계산한 결과 고 밀도 폴리에틸렌의 두께가 10 cm 일 때 열중성자속이 최대가 되는 것으로 나타났으며 감속된 중성자의 흡수는 용제에 자연상태의 $B_4C$ 분말을 30 w% 섞을 경우 2 mm의 두께에서 94%의 중성자 흡수가 일어나는 것으로 나타났다. 또한 몬테카를로 모사를 통한 계산결과의 타당성 여부를 조사하기 위하여 중성자 차폐실험 장치를 제작하여 실험 결과와 비교하였으며, 비교 결과 실험값과 일치하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.674-680
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• 2020

We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO₂ and (Th+Pu)O₂ mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O₂, achieved about 24% higher fuel cycle length than the UO₂ case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found.

• 한국방사선학회논문지
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• 제11권4호
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• pp.197-203
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• 2017

의료용 선형가속장치의 두부 구성요소 중 광자 발생의 원인이 되는 타깃에 대한 연구로써, 타깃의 재질에 따른 광자를 분석하여 타깃 재질 별 발생하는 광자특성에 대한 기초자료를 제시하고자 한다. 본 연구에서는 몬테카를로 방식을 바탕으로 한 MCNPX를 사용하여 타깃 재질에 따른 6, 15 MV의 광자 특성을 비교분석하였다. 타깃 재질 별 평균에너지는 6 MV에서 1.69 ~ 1.84 MeV, 15 MV에서는 3.38 ~ 3.56 MeV로 분석되었다. Flux는 6 MV에서 $1.64{\times}10^{-5}{\sim}1.80{\times}10^{-5}{\sharp}/cm^2/e$, 15 MV는 $1.76{\times}10^{-4}{\sim}1.85{\times}10^{-4}{\sharp}/cm^2/e$로 계산되었다. 결과를 분석하면, 타깃 재질이 고원자번호일수록 평균에너지와 Flux가 증가하는 것으로 평가다. 본 연구를 바탕으로 광자의 물리적 특성에 대한 기초적인 자료를 제시할 수 있었으며, 추후 타깃 선정 시 경제성, 효율성은 물론 물리적 측면을 고려할 수 있어 적절한 선택을 할 수 있을 것으로 판단된다.

• Journal of Radiation Protection and Research
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• 제24권4호
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• pp.205-213
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• 1999

지표에 오염된 방사성핵종의 단위방사능당 유효선량환산계수를 남성과 여성 인형모의피폭체와 MCNP4A 코드를 이용하여 계산하였다. 모사실험은 40 keV에서 10 MeV 영역의 19개 단일 에너지에 대한 유효선량 계산을 수행하였다. 에너지에 따른 단위 선원강도에 대한 유효선량 E를 기존 연구자들의 결과물인 유효선량당량 $H_E$와 비교한 결과, 본 연구의 E값이 USEPA의 FGR에 주어진 $H_E$ 값에 비해 30%의 편차를 보였다. 에너지와 유효선량의 관계를 polynomial fitting을 통해 구한 유효선량 감응함수는 다음과 같다. $f({\varepsilon})[fSv\;m^2]=\;0.0634\;+\;0.727{\varepsilon}-0.0520{\varepsilon}^2+0.00247{\varepsilon}^3$ 여기서, ${\varepsilon}$는 감마선의 에너지(MeV)이다. 감응함수와 ICRP 38의 방사성핵종 붕괴 자료를 이용하여 지표면과 공기 오염의 단위 방사능농도에 대한 유효선량환산계수를 계산한 후 DOSEFACTOR코드를 사용하여 계산한 베타선에 의한 피부선량을 합하여 90개의 중요 핵종들에 대한 환산계수를 평가하여 도표로 제시하였다. 기존 자료들과 비교를 통해 기존 환산계수를 사용할 경우 특히 저에너지 감마선이나 고에너지 베타선을 방출하는 핵종에 대해서 상당한 과소평가가 이루어질 수 있음을 확인할 수 있었다.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.365-374
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• 1999

A nuclear design of fission moly production targets for a research reactor, HANARO was peformed. It was found that the use of MCNP-4A, ORIGEN-2 code was reliable for the analysis of production characteristics of $^{99}$ Mo in a target fuel at an irradiation holes. A parametric study was done for the optimization of target location, target dimension, target shape and fuel materials. It was shown that a fuel thickness was the most sensitive parameters and electro-deposited target gave the highest 99Mo yield ratio. A pellet target with vibro-compaction powder, however, showed the largest production capacity and better engineering feasibility even with less yield ratio. Ten kinds of optimized target design for both LEU and HEU satisfied all the given design constraints. The most favorable design was the HEU ring-shaped electro-deposited target, considered the safety limit, production yield, chemical process easiness, yield ratio, and amount of radioactive waste.

• Journal of Biosystems Engineering
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• 제38권2호
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• pp.87-94
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• 2013

Purpose: Pineapple is now the third most important tropical fruit in world production after banana and citrus. Phytosanitary irradiation is recognized as a promising alternative treatment to chemical fumigation. However, most of the phytosanitary irradiation studies have dealt with physiochemical properties and its efficacy. Accurate dose calculation is crucial for ensuring proper process control in phytosanitary irradiation. The objective of this study was to optimize phytosanitary irradiation treatment of pineapple in various radiation sources using Monte Carlo simulation. Methods: 3-D geometry and component densities of the pineapple, extracted from CT scan data, were entered into a radiation transport Monte Carlo code (MCNP5) to obtain simulated dose distribution. Radiation energy used for simulation were 2 MeV (low-energy) and 10 MeV (high-energy) for electron beams, 1.25 MeV for gamma-rays, and 5 MeV for X-rays. Results: For low-energy electron beam simulation, electrons penetrated up to 0.75 cm from the pineapple skin, which is good for controlling insect eggs laid just below the fruit surface. For high-energy electron beam simulation, electrons penetrated up to 4.5 cm and the irradiation area occupied 60.2% of the whole area at single-side irradiation and 90.6% at double-side irradiation. For a single-side only gamma- and X-ray source simulation, the entire pineapple was irradiated and dose uniformity ratios (Dmax/Dmin) were 2.23 and 2.19, respectively. Even though both sources had all greater penetrating capability, the X-ray treatment is safer and the gamma-ray treatment is more widely used due to their availability. Conclusions: These results are invaluable for optimizing phytosanitary irradiation treatment planning of pineapple.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.523-531
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• 2022

In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U₂₃₅(n, f), $ U_{235}(n\bar{\nu})$ and H₁(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.

• Journal of Radiation Protection and Research
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• 제48권4호
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• pp.184-196
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• 2023

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities. Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO₃), and tungsten disulfide (WS₂). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z_eff), and macroscopic effective removal cross-section for fast neutrons (Σ_R) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation. Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF-xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF-0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF-0.20WO₃ and PVDF-0.20WS₂. PVDF-0.20WC also exhibited the highest Z_eff values, while PVDF-0.20WS₂ showed a slightly higher increase in Z_eff at energies of 0.662 and 1.333 MeV. PVDF-0.20WC has demonstrated the highest Σ_R value, indicating effective shielding against fast neutrons, while PVDF-0.20WS₂ had the lowest Σ_R value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF-xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate. Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF-xWC showing the most promising results.

• Radiation Oncology Journal
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• 제19권1호
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• pp.66-73
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• 2001

목적 : 붕소-중성자 포획치료법(Boron Neutron Capture Therapy, BNCT)을 위해 원자력병원 싸이클로트론에서 발생되는 최대에너지 34.4 MeV의 속중성자(Fast neutron)를 70 cm 파라핀으로 감속시킨 후 선량 특성을 조사하였다. 그 결과를 토대로 열외중성자(Epithermal neutron) 선량 측정법에 대한 프로토콜을 확립하여 원자로에서 방출되는 열외 중성자 선량 특성 평가의 기초를 삼고, 가속기를 이용한 BNCT 연구에 대한 타당성 여부를 조사하고자 한다. 대상 및 방법 : 공기 중 선량 및 물질 내 선량 분포 측정을 위해 Unidos 10005 (PTW, Germany) 전기계와 조직 등가 물질인 A-150 플라스틱으로 제작된 IC-17 (Far West, USA) 및 IC-18, ElC-1 이온함을 사용하였고, 감마선의 측정을 위해서는 마그네슘으로 제작된 IC-l7M 이온함을 이용하였으며 조직등가 기체와 아르곤 기체를 분당 5cc 씩 주입하며 측정하였다. 중성자, 광자, 전자가 혼합된 장의 모의 수송 해석을 위해 이용되는 Monte Carlo N-Particle (MCNP) transport code를 사용하여 2차원적 선량 분포 및 에너지 분포를 계산하였으며 이 결과를 측정값과 비교하였다. 결과 : BNCT에서의 유효 치료 깊이인 물 팬텀 4 cm에서의 선량은 치료기 1 MU 당 $6.47\times10^{-3}\;cGy$로 미세하였으며, 이때 감마 오염도(contamination)는 $65.2{\pm}0.9\%$로 중성자보다는 감마선에 의한 선량 기여분이 우세하였다. 깊이에 따른 선량 분포 특성에서는 중성자 선량은 선형적으로 감쇠 되었고, 감마선량은 지수적으로 보다 급격히 감쇠되는 경향을 보였으며 전체 선량의 $D_{20}/D_{10}$은 0.718 이었다. MCNP에 의한 에너지 분포 전산 계산의 결과 2.87 MeV 이하에서 중성자 피크가 나타났으며, 저에너지 영역에서는 감마선이 연속적으로 분포되는 양상을 보였다. 결론 : 벽 물질이 서로 다른 두 개의 이온함을 사용한 직접 선량 측정과 MCNP 전산 시뮬레이션을 이용한 공간 선량분포 계산으로 미세 속중성자 빔에 대한 선량 특성을 파악할 수 있었으며, 원자로 열외중성자 주(Epithermal neutron column)에 대한 선량 평가 자료로 확보하였다. 아울러 가속기에 대한 연구가 진행되어 고전압, 고전류를 발생시키는 전원 공급장치와 표적핵(Target) 물질이 개발되고 비스무스나 납 등에 의해 감마 오염도를 줄일 경우, 싸이크로트론에 의한 보론-중성자 포획치료도 가능해질 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.