• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3158-3163
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• 2021

Chromium material is commonly used for fusion plasma facing applications because of the low neutron activation property. The Monte Carlo method is one of the useful ways to investigate the ion-target interactions. In this study, Chromium target irradiated by protons was investigated using Monte Carlo based simulation tools. In this context, the calculations of radiation damage on Chromium material irradiated with protons at 17.9 and 22.3 MeV energies were carried out using GEANT4 and SRIM codes. Besides, the cross sections for proton interaction with Chromium target were calculated by the TALYS 1.9 code using CTM + FGM, BSFGM, and GSFM level densities. As a result, GEANT4, SRIM and TALYS 1.9 codes provide a suitable tool for the predictions of radiation damage and cross cross section with proton irradiation.

• 대한방사선기술학회지:방사선기술과학
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• 제47권3호
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• pp.213-218
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• 2024

The myocardial nuclear medicine examination is widely performed to diagnose myocardium disease using various radionuclides. Although image quality according to radionuclides has improved, the radiation exposure for target organ as well as peripheral organs should be considered. Here, the aim of this study was to evaluate absorbed dose (Gy) for peripheral organs in myocardial nuclear medicine scan from myocardium according to various scan environments based on Monte Carlo simulation. The simulation environment was modeled 5 cases, which were considered by radionuclides, number of injections, and radiodosage. In addition, the each radionuclide simulation such as distribution fraction was considered by recommended standard protocol, and the mesh computational female phantom, which is provided by International Commission on Radiological Protection (ICRP) 145, was used using the particle and heavy ion transport code system (PHITS) version 3.33. Based on the results, the closer to the myocardium, the higher the absorbed dose values. In addition, application for dual injection for radionuclides leaded to high absorbed dose compared with single injection for radionuclide. Consequently, there is difference for absorbed dose according to radionuclides, number of injections, and radiodosage. To detect the accurate diseased area, acquisition for improved image quality is crucial process by injecting radionuclides, however, we need to consider absorbed dose both target and peripheral inner organs from radionuclides in terms radiation protection for patient.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1593-1603
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• 2023

In deterministic and Monte Carlo transport codes, b-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available b-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code - named Time Dependent OpenMC or OpenMC(TD)- time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.

• 한국방사선학회논문지
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• 제5권6호
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• pp.383-389
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• 2011

Geant4와 DICOM 파일의 연동을 이용한 몬테카를로 시뮬레이션을 통해 실제 환자의 흡수선량을 산출하는 새로운 방법을 제시하고, Geant4 계산코드의 검증을 위해 MOSFET 선량계를 이용하여 PMMA 모의 팬텀 깊이에 따른 중심에서의 흡수선량 실측값과 Geant4 시뮬레이션 결과값을 비교하였다. PMMA slab의 불완전한 압착으로 인해 발생한 불균등한 간격의 공기층이 존재하지 않은 부분에서는 X선 조사야 $15{\times}15cm^2$와 $20{\times}20cm^2$에서 각각 $0.46{\pm}4.69%$와 $-0.75{\pm}5.19%$로 나타났다. PMMA 모의 팬텀의 불완전한 압착에 의해 나타난 오차를 제외하면 Geant4와 DICOM 파일의 연동을 통한 몬테카를로 시뮬레이션에 의한 계산값이 잘 일치함을 알 수 있다.

• 한국정밀공학회지
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• 제18권10호
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• pp.154-160
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• 2001

For major structural components periodic inspections and integrity assessments are needed for the safety. However, many flaws are undetectable because sampling inspection is carried out during in-service inspection. Probabilistic integrity assessment is applied to take into consideration of uncertainty and variance of input parameters arise due to material properties and undetectable cracks. This paper describes a Probabilistic Fracture Mechanics(PFM) analysis based on the Monte Carlo(MC) algorithms. Taking a number of sampling data of probabilistic variables such as fracture toughness value, crack depth and aspect ratio of an initial surface crack, a MC simulation of failure judgement of samples is performed. for the verification of this analysis, a comparison study of the PFM analysis using a commercial code, mathematical method is carried out and a good agreement was observed between those results.

• Nuclear Engineering and Technology
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• 제29권6호
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• pp.459-467
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• 1997

The steady-state behavior of recycling neutral atoms in a tokamak edge region has been analyzed through a two-dimensional Monte Carlo simulation. A particle tracking algorithm used in earlier research on the neutral particle transport is applied to this Monte Carlo simulation in order to perform more accurate calculations with the EDGETRAN code which was previously developed for a two-dimensional edge plasma transport in the authors' laboratory. The physical model of neutral recycling includes charge-exchange and ionization interactions between plasmas and neutral atoms. The reflection processes of incident particles on the device wall are described by empirical formulas. Calculations for density, energy, and velocity distributions of neutral deuterium-tritium atoms have been carried out for a medium-sized tokamak with a double-null configuration based on the KT-2 conceptual design. The input plasma parameters such as plasma density, ion and electron temperatures, and ion fluid velocity are provided from the EDGETRAN calculations. As a result of the present numerical analysis, it is noticed that a significant drop of the neutral atom density appears in the region of high plasma density and that the similar distribution of neutral energy to that of plasma ions is present as frequently reported in other studies. Relations between edge plasma conditions and the neutral recycling behavior are discussed from the numerical results obtained herein.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4678-4684
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• 2023

Using the Monte Carlo method, the impact of the angular distribution of the electron source on the dose distribution for the 2.5 MeV ELV electron accelerator was explored. The experiment measured the 3-D dose distribution in the irradiation chamber for electron energies of 1.0 MeV and 2.5 MeV. The simulation used the MCNP6.2 code to evaluate three angular distribution models of the source: a mono-directional beam, a cone shape, and a triangular shape. Of the three models, the triangular shape with angles θ = 30°, φ = 0° best represents the angle of the scan hood through which the electron beam exits. The MCNP6.2 simulation results demonstrated that the triangular model is the most accurate representation of the angular distribution of the electron source for the 2.5 MeV ELV electron accelerator.

• 한국진공학회지
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• 제12권3호
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• pp.162-167
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• 2003

시료에 주입된 이온의 깊이방향에 따른 농도분포를 알아보기 위하여 시료표면을 sputtering 하면서 튀어나온 주입된 이온을 depth profiling한다. Depth profiling 측정 시에 깊이방향에 영향을 주는 sputtering rate가 변화하는 효과를 SRIM simulation을 이용하여 계산하였다. 시료에 이온이 주입하게 되면 시료의 원자밀도는 약간 증가하게 되는데, 그 결과로 sputtering yield가 변화하게 된다. 이러한 변화가 결과적으로 depth profile 측정시에 깊이방향에 영향을 줄 수 있는 sputtering rate를 변화시키는 원인이 된다. SRIM(Stopping and Range of Ions in Matter) Monte Carlo simulation code를 사용하여 이온주입에 의한 시료의 원자밀도의 변화에 따른 sputtering yield를 구하여 sputtering rate를 계산하고, 그 차이가 depth profiling 측정에서 깊이방향 분포에 영향을 줄 수 있다는 것을 확인하였다.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.801-809
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• 2017

The purpose of this study is to assess the additional neutron effective dose during passive scattering proton therapy. Monte Carlo code (Monte Carlo N-Particle 6) simulation was conducted based on a precise modeling of the National Cancer Center's proton therapy facility. A three-dimensional neutron effective dose profile of the interior of the treatment room was acquired via a computer simulation of the 217.8-MeV proton beam. Measurements were taken with a $^3He$ neutron detector to support the simulation results, which were lower than the simulation results by 16% on average. The secondary photon dose was about 0.8% of the neutron dose. The dominant neutron source was deduced based on flux calculation. The secondary neutron effective dose per proton absorbed dose ranged from $4.942{\pm}0.031mSv/Gy$ at the end of the field to $0.324{\pm}0.006mSv/Gy$ at 150 cm in axial distance.

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.338-345
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• 2004

This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the $UO_2$ fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the $UO_2$ fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the $UO_2$ fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the $UO_2$ fuel.