• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2013.05a
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• pp.1017-1019
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• 2013

The radiation sources leaked from large-scale radiation leak accident like the Hukushima nuclear power plant accident or nuclear explosions can cause to the very large damage for us. So that the damage can be minimized, we have being developed a detector that can providing information about the location of the source to remove dangerous substances quickly than the conventional single detector. In this paper, we designed the shielding and collimator for the development of the Stereo Radiation detector in order to detect contaminants using MCNP Code. And we analyzed the results that is detected from the discretionary position of the radiation source. The results of this paper will be used as the basis for designing efficient structure for the radiation detectors.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2013.05a
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• pp.1031-1033
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• 2013

Radiation protection, as well as finding the location of the radiation source, such as the Fukushima radiation leak accident, it is important for the early and safe disposal of nuclear accident. The three-dimensional position of the radiation source detection distance of the radiation source can provide additional information to the existing radiation detectors radiation of a two-dimensional position detection function and then it can play a decisive role in the radiation contaminant removal and decontamination work. In this research, three-dimensional semiconductor sensor based on dual radiation detectors radiation source device visible part of the research and development of efficient radiation sensor unit on the design of the shielding structure.The lightweight, high-efficiency radiation source locator implementation was attempted for the structure and thickness of the shielding and collimator to perform the simulation of the radiation shielding for the various parameters of the shape model through design the optimal structure of the MCNP-based heavy-duty tungsten shielding, lead shielding The results of this study, is a compact, lightweight three-dimensional radiation source detection and future of silicon - based sensors will be used in the study.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Geophysics and Geophysical Exploration
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• v.15 no.4
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• pp.219-226
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• 2012

This study provides an analysis on results of neutron logging for various borehole environments through numerical simulation based on a Monte Carlo N-Particle (MCNP) code developed and maintained by Los Alamos National Laboratory. MCNP is suitable for the simulation of neutron logging since the algorithm can simulate transport of nuclear particles in three-dimensional geometry. Rather than simulating a specific tool of a particular service company between many commercial neutron tools, we have constructed a generic thermal neutron tool characterizing commercial tools. This study makes calibration chart of the neutron logging tool for materials (e.g., limestone, sandstone and dolomite) with various porosities. Further, we provides correction charts for the generic neutron logging tool to analyze responses of the tool under various borehole conditions by considering brine-filled borehole fluid and void water, and presence of borehole fluid.

• Journal of Radiation Protection and Research
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• v.19 no.1
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• pp.13-22
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• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• Journal of Electrical Engineering and Technology
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• v.6 no.1
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• pp.139-146
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• 2011

This paper presents a new modified Contract Net Protocol (CNP) for microgrid operation based on multiagent system. The CNP is a widely used protocol for interactions among distributed problem solving. The Contract Net Interaction Protocol of the Foundation for Intelligent Physical Agents (FIPA-CNIP) is a minor modification of the original CNP for multiagent system applications. In this paper, a modified CNP (MCNP) based on the FIPA-CNIP is proposed for more specialized interactions among agents for microgrid operation. A multiagent system is designed and constructed for microgrid operation. A microgrid operation based on the multiagent system is tested to check the functionality of the proposed MCNP.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.606-611
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• 1998

330MWt 출력의 신형 원자로인 SMART(System integrated Mod씰w Advanced ReacTor)가 전기 생산뿐만 아니라 해수의 담수화를 위한 에너지 공급을 위해 한국원자력연구소에 의해 개발되고 있다. SMART의 원자로 압력용기에서의 중성자 조사량을 기존의 각분할법 코드 대신에 몬데칼로 수송 코드인 MCNP-4A를 이용하여 평가하였다. MCNP-4A에 의한 몬데 칼로 모사는 각분할법에 비해 핵 단면적 자료, 선원항, 그리고 기하학적 모델링의 문제로부터 야기되는 불확실성을 감소시킬 수 있을 뿐만 아니라 초기 개념 설계 단계에서 상세 노심 출력 분포 자료에 의존하지 않고 선원항을 평가할 수 있는 장점이 있다. 본 연구에서는 원자로 압력 용기 내부의 원자로 노심 및 다른 구조물을 포함하는 전체 원자로 구조에 대하여 몬테 칼로 모사를 적용하였다. 1단계에서는 임계도 계산에 의해 선원항으로 이용되는 원자로 노심내의 열 출력 분포를 평가하고, 2단계에서는 노심내의 열 출력 분포를 고정 선원으로 이용하여 압력 용기에서의 중성자 조사량을평가하였다. 그 결과 SMART 압력용기의 중성자 조사량은 규제 요건을 만족하는 것으로 나타났다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.618-623
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• 1998

A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between AMISN and MCNP estimates, which may require a consistent library generation for both codes.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.115-120
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• 1997

영광3호기 방사선관리구역에 대한 중성자선량률을 정확히 평가하기 위하여 MCNP4A 전산코드를 이용, 방사선관리구역에서의 중성자 스펙트럼 계산을 수행하였다. 영광3호기에 대한 보다 정확하고 정밀한 3차원 몬테칼로 모델을 구축하기 위하여 핵연료집합체 구성요소 및 원자로심을 둘러싸고 있는 baffle, barrel,압력용기 등을 정확하게 묘사하였으며, 특히 방사선관리구역 주위의 구조물에 대해서도 3자원 MCNP 모델을 구축함으로써 원자로심부터 방사선관리구역까지 완전한 몬테칼로 모사(full-scope Monte Carlo simulation)를 이용한 계산을 수행하였다. 계산결과는 에너지 구간에 따른 중성자속 스펙트럼으로 나타내었으며 이 결과를 바탕으로 중성자속에 대한 선량률 환산인자를 고려하여 중성자선량률을 계산할 수 있다.

• 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.