• 대한의용생체공학회:의공학회지
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• 제19권3호
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• pp.225-232
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• 1998

무정형 실리콘을 기반으로 한 X-선 영상기기에 대한 Monte Carlo 시뮬레이션 결과를 기술하였다. 무정형 실리콘 X-선 영상기기의 특성을 조사하고 최적의 설계변수들을 제공하기 위하여 Monte Carlo 시뮬레이션을 수행하였다. 본 연구의 목적에 맞도록 Monte Carlo simulation codes를 개발하였고, X-선 최대전압, 알루미늄 필터 두께, Cal(T1)두께, 그리고 무정형 실리콘 광다이오우드 픽셀 크기들을 변화시키면서 무정형 실리콘 X-선 영상기기의 계측 효율과 해상도의 변화를 연구하였다. 60kVP-120kVp의 X-선에 대하여, CsI(TI)의 두께가 300um-500um일 때 계측효율은 70%-95% 였고 에너지 흡수효율은 40%-70%였다. 시뮬레이션 결과로부터, 무정형 실리콘 픽셀크기와 Csl(TI) 두께가 해상도를 결정하는 가장 주된 요소임이 밝혀졌다. 본 연구에서 개발한 시뮬레이션을 사용하여 감도와 해상도를 최적화할 수 있는 적절한 픽셀 크기와 CsI(TI) 두께를 찾아낼 수 있었다.

• 한국의학물리학회지:의학물리
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• 제19권1호
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• pp.21-34
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• 2008

환자의 CT자료를 기반으로 만들어진 3차원상의 표적물질에 전자 및 광자의 전달 현상을 계산하는 몬테카를로(MC) 도즈계산용 병렬프로그램 (PMCEPT 코드)을 개발하여 베어울프 PC 클러스터에 탑제하였다. 시뮬레이션에서 오차를 최소화하고 코드를 더욱 발전시키기 위해서는 현재의 MC 코드의 한계를 아는 것이 매우 유익하다. 이러한 관점에서 저자는 PMCEPT코드를 이용하여 이질 혹은 동질의 표적물질에서 표준화된 깊이 도즈를 계산하여 잘 알려진 다른 코드들, MCNP5, EGS4, DPM, GEANT4 및 실험결과와 비교를 하였다. PMCEPT결과는 이질 혹은 동질의 표적에서 다른 코드들과 $1{\sim}3%$ 오차 범위 안에서 잘 일치하였다. 계산시간 비교에 있어서도 PMCEPT 코드가 MCNP5 보다는 약 20배, GEANT4코드보다는 약 3배정도 빨랐다. 이러한 결과를 종합하면, PMCEPT코드는 의학물리분야의 시뮬레이션 코드로 사용하기에 매우 좋은 것으로 사료된다.

• Nuclear Engineering and Technology
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• 제51권8호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.46-51
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• 1996

In this study we calculate the infite multiplication factors ($k_{\infty}$) and the Doppler temperature coefficients (DTC) of two mixed-oxide (MOX) fuel rods with different plutonium contents by using PHOENIX-P, HELIOS and CASMO-3 codes. The results were compared against the reference values obtained by MCNP-3A continuous-energy Monte Carlo code. The purpose of this study is to benchmark the accuracy of these lattice codes. The PHOENIX-P's Doppler coefficients calculated were in good agreement with the MCNP results within the Monte-Carlo uncertainty band which is in the order of $\pm$ 10% for the Doppler coefficients..

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1998-2004
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• 2019

The Multi-Physics (MP) instrument is one of 20 neutron spectrometers planned in the China Spallation Neutron Source (CSNS). This paper presents a shielding calculation for the MP instrument using Monte Carlo codes MCNPX and FLUKA. First, the neutrons that escape from the CSNS decoupled water moderator and are delivered to the beam line of the MP instrument are calculated to use as the source term of the shielding calculation. Then, to validate the calculation method based on multiple variance reduction techniques, a cross check between MCNPX and FLUKA codes is performed by comparing the calculation results of the dose rate distribution on a simplified beam line model. Finally, a complete geometry model of the MP instrument is set up, and the primary parameters for the shielding design are obtained according to the calculated dose rate map considering different worst-case scenarios.

• 한국통신학회논문지
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• 제38C권10호
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• pp.852-857
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• 2013

본 논문은 오류 마루 영역에서 우수한 성능을 가지는 비이진 LDPC (low-density parity-check) 부호의 설계 방법을 제안하고 성능을 검증한다. 제안된 설계 방법은 비이진 LDPC 부호의 이진 최소 거리(binary minimum distance)를 최대화하도록 패리티 검사 행렬의 비이진 원소 값들을 결정한다. BPSK (binary phase-shift keying) 변조 방식 하에서 제안된 방법으로 설계된 비이진 LDPC 부호가 오류 마루(error floor) 영역에서 우수한 성능을 가지는 것을 Monte Carlo 시뮬레이션과 중요도 표본 추출(importance sampling) 기법을 사용하여 검증한다.

• 한국정보통신학회:학술대회논문집
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• 한국해양정보통신학회 2000년도 추계종합학술대회
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• pp.235-238
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• 2000

본 논문에서는 BS-CDMA 이동통신 시스템에서 다중 사용자 간섭과 Rayleigh 페이딩의 영향을 효과적으로 제거하기 위하여 터보 부호화 다중 사용자 검출 방식을 사용하였으며, Monte Carlo 시뮬레이션을 이용하여 다양한 채널 조건에서 제안된 방식의 성능을 분석하였다.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2002년도 Proceedings
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• pp.485-488
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• 2002

In order to understand and quantitatively analyze the physical phenomena and behavior of each component of mammography system during the breast imaging, we simulated mammography imaging using Monte Carlo simulation codes. MCNP4B code was used for our simulation purpose, and we investigated the effect of target material, anode angle, filtration, peak voltage and exposure on the image quality of mammograms. From the simulation results we expect that optimized operation condition of mammography system can be found.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.811-816
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• 2022

This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction β_eff and neutron generation time L. The modified code, OpenMC(Time-Dependent) or OpenMC(TD), was then used to calculate the effective delayed neutron fraction by using the prompt method, while the neutron generation time was estimated using the pulsed method, fitting Λ to the decay of the neutron population. OpenMC(TD) is intended to serve as an alternative for the estimation of kinetic parameters when licensed codes are not available. The results obtained are compared to experimental data and MCNP calculated values for 18 benchmark configurations.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.151-158
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• 2018

The neutron powder diffractometer (NPD) is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo-based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo-based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than $10^6n/s/cm^2$ at sample position.