• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• 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.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.430-438
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• 2021

The spatial distribution of neutron flux or reaction rate was calculated by cell or mesh tally in traditional Monte Carlo simulation. However, either cell or mesh tally leads to the increase of memory consumption and simulation time. In this paper, the function expansion tally (FET) method was developed in Reactor Monte Carlo code RMC to solve this problem. The FET method was applied to the tallies of neutron flux distributions of uranium block and PWR fuel rod models. Legendre polynomials were used in the axial direction, while Zernike polynomials were used in the radial direction. The results of flux, calculation time and memory consumption of different expansion orders were investigated, and compared with the mesh tally. Results showed that the continuous distribution of flux can be obtained by FET method. The flux distributions were consistent with that of mesh tally, while the memory consumption and simulation time can be effectively reduced. Finally, the convergence analysis of coefficients of polynomials were performed, and the selection strategy of FET order was proposed based on the statistics uncertainty of the coefficients. The proposed method can help to determine the order of FET, which was meaningful for the efficiency and accuracy of FET method.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.790-802
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• 2022

Large-scale reactor simulation often requires the use of Monte Carlo calculation techniques to estimate important reactor parameters. One drawback of these Monte Carlo calculation techniques is they inevitably result in some uncertainty in calculated quantities. The present study includes parametric uncertainty quantification (UQ) and sensitivity analysis (SA) on the Advanced Test Reactor Critical (ATRC) facility housed at Idaho National Laboratory (INL) and addresses some complications due to Monte Carlo uncertainty when performing these analyses. This approach for UQ/SA includes consideration of Monte Carlo code uncertainty in computed sensitivities, consideration of uncertainty from directly measured parameters and a comparison of results obtained from brute-force Monte Carlo UQ versus UQ obtained from a surrogate model. These methodologies are applied to the uncertainty and sensitivity of k_eff for two sets of uncertain parameters involving fuel plate geometry and fuel plate composition. Results indicate that the less computationally-expensive method for uncertainty quantification involving a linear surrogate model provides accurate estimations for k_eff uncertainty and the Monte Carlo uncertainty in calculated k_eff values can have a large effect on computed linear model parameters for parameters with low influence on k_eff.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2002년도 Proceedings
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• pp.282-284
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• 2002

We have developed and used BISS as a radiation detector to verify patient dose and determine the physical characteristics of beams used in Stereotatic Radio Surgery(SRS) and Intensity Modulated Radiation Therapy(IMRT). In order to confirm the function and accuracy of our BISS, we simulate our measurements by BISS under the radiation of 6MV photons from a Varian Clinac 21EX equipped with a 60 leaf pairs MLC. For the simulation based on the Monte Carlo algorithm, which remains the most comprehensive and accurate theoretical method to verify beam profiles, we use the BEAM code. Compared with the measurements by BISS, our simulation of variously shaped phantom measurements show good agreements. Our simulation results can be used as a theoretical standard to compare and confirm measurements by BISS and other dosimeters such as ultramicro cylindrical ionization chamber(UCIC) and radiographic film.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2002년도 Proceedings
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• pp.116-118
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• 2002

It has been noted that Monte Carlo simulations are the most accurate method to calculate dose distributions in any material and geometry. Monte Carlo transport algorithms determine the absorbed dose by following the path of representative particles as they travel through the medium. Accurate Monte Carlo dose calculations rely on detailed modeling of the radiation source. We modeled the effects of beam modifiers such as collimators, blocks, wedges, etc. of our accelerator, Varian Clinac 600C/D to ensure accurate representation of the radiation source using the EGSnrc based BEAM code. These were used in the EGSnrc based DOSXYZ code for the simulation of particles transport through a voxel based Cartesian coordinate system. Because Monte Carlo methods use particle-by-particle methods to simulate a radiation transport, more particle histories yield the better representation of the actual dose. But the prohibitively long time required to get high resolution and accuracy calculations has prevented the use of Monte Carlo methods in the actual clinical spots. Our ultimate aim is to develop a Monte Carlo dose calculation system designed specifically for radiation therapy planning, which is distinguished from current dose calculation methods. The purpose of this study in the present phase was to get dose calculation results corresponding to measurements within practical time limit. We used parallel processing and some variance reduction techniques, therefore reduced the computational time, preserving a good agreement between calculations of depth dose distributions and measurements within 5% deviations.

• 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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• 제52권5호
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• 한국의학물리학회지:의학물리
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• 제18권4호
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• pp.226-232
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• 2007

국립암센터에 설치된 양성자 치료기의 빔 전달 시스템에 대하여 Geant4 코드를 이용하여 몬테카를로 전산모사를 수행하였고, 선량검증 도구로써의 이용 가능성에 대하여 연구하였다. 몬테카를로 기술을 기반으로 하는 선량계산은 물질내의 선량분포를 이해하는 데 가장 정확한 방법으로 알려져 있다 외부조사 방사선치료에 있어서 이 방법의 장점을 극대화 하기 위해서는, 빔이 지나가는 곳에 놓여진 노즐 구성요소들의 정확한 모델링과 더불어 초기빔 특성파악은 무엇보다 중요하다. 국립암센터에 설치된 양성자 치료기는 총 3가지 형태-double/single scattering, uniform scanning and pencil-beam scanning-로 치료빔을 조사할 수 있으며, 본 연구진은 Geant4.8.2 코드를 기반으로 double/single scattering 모드를 구성하는 모든 노즐구성요소들에 대하여 모델링 하였다. 특정 치료감이에 대하여 실험치와 일치하는 전산모사의 결과를 얻었다 본 기관에 설치된 양성자치료기에 대한 몬테카를로 전산모사에 대한 기반을 성공적으로 구축하였고, 치료빔에 대하여 정밀한 선량측정에 이용할 수 있다. 치료빔의 전 에너지 영역에 걸쳐 추가적인 커미셔닝을 수행할 것이다.