• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.357-357
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.46-46
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• 1999

• 한국진공학회:학술대회논문집
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• 한국진공학회 2008년도 제35회 하계학술대회 초록집
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• pp.187-187
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• 2008

• 한국방사선학회논문지
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• 제3권3호
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• pp.23-25
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• 2009

Gd-157 material has very large neutron capture cross section in the thermal region. So it is very useful to shield material for thermal neutrons. Futhermore, in the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the case of very narrow resonance, the effect of scattering is to be more considerable factor. In the present study, we obtained energy dependent neutron absorption ratios of natural indium in energy region from 0.003 to 100 keV by MCNP-4B Code. The coefficients for neutron absorption was calculated for circular type and 1 mm thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1505-1512
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• 2017

Because accelerator-based boron neutron capture therapy (BNCT) systems are planned for use in hospitals, entry into the medical room should be controlled as hospitals are generally assumed to be public and safe places. In this paper, computational investigation of the medical staff effective dose during BNCT has been performed in different situations using Monte Carlo N-Particle (MCNP4C) code and two voxel based male phantoms. The results show that the medical staff effective dose is highly dependent on the position of the medical staff. The results also show that the maximum medical staff effective dose in an emergency situation in the presence of a patient is ${\sim}25.5{\mu}Sv/s$.

• 한국컴퓨터산업학회논문지
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• 제8권1호
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• pp.9-16
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• 2007

본 연구에서는 아스팔트 함량 변화에 따라서 중성자 계측수가 어떻게 변화되는가를 계산하여 법적 규제 면제치인 $100[{\mu}Ci]$이하의 방사성동위원소를 이용한 아스팔트 함량측정기의 기본 설계 자료로 활용하고자한다. 이를 위하여 1차 년도에서 실시했던 설계자료를 활용하여 아스팔트 함량의 변화에 따라 중성자 계측수가 어떻게 증감이 이루어지고, 또한 감속재인 폴리에틸렌 주변에 흡수체인 카드늄판을 설치했을 때의 계측수의 변화를 MCNP 코드를 이용하여 살펴보았다.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2004년도 제29회 추계학술대회 발표논문집
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• pp.33-35
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• 2004

방사선치료를 위한 고에너지 광자선의 품질관리를 위해 사용하는 TLD의 광자선 선질에 대한 에너지 의존도를 몬테카를로 모사법을 사용하여 평가하였다. IAEA 선량보증사업에 이용되는 LiF TLD 및 홀더를 EGS4기반의 사용자 코드인 DOSIMETER 와 MCNP4C 몬테카를로 코드를 사용하여 기하학구조를 구성하고, Co, 4, 6,10 밑 15 MV 광자선을 시뮬레이션하였다. DOSIMETER계산 결과를 통해 TLD의 에너지 보정인자가 실험 데이터와 일치함을 확인할 수 있었으며, 이와 별도로 캡슐에 의한 교란량도 무시할 수 없음을 발견하였다.

• Nuclear Engineering and Technology
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• 제35권6호
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• pp.581-595
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• 2003

Transport lattice codes require the resonance integral tables for the resonant nuclides where the resonance integral is a function of the background cross section and can be prepared through a special program solving the slowing down equation. In case the cross section libraries do not include the resonance integral table for the resonant nuclides, the computational prediction produces a large error. We devised a new method using a Monte Carlo calculation for the effective resonance cross sections to solve this problem provisionally. We extended this method to obtain the resonance integral table for general purpose. The MCNP code is used for the effective resonance integrals and the LIBERTE code for the effective background cross sections. We modified the HELIOS library with the effective cross sections and the resonance integral tables obtained by the newly developed Monte Carlo method, and performed sample calculations using HELIOS and LIBERTE. The results showed that this method is very effective for the resonance treatment.

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.225-230
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• 2006

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of variance reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered in the areas of source definition, skyshine, streaming, and transmission.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.477-484
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• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.