• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.179-184
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• 1996

Power density distribution and criticality of a pressurized water reactor are calculated with a Monte Carlo calculation using the MCNP code. The MCNP model is based on one-eighth core symmetry. Individual fuel assemblies are modeled with fullscope three dimensional description except grid spacer. The fuel rod is divided into eight axial segments. Core internals above and below the active fuel region is represented as coolant. After 400 cycle calculations, the system converges to a k value of 1.09151$\pm$0.00066. Fission reaction rate in each rod is also calculated to use as the source term in pressure vessel fluence calculation.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.19-29
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• 2021

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.

• Geophysics and Geophysical Exploration
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• v.15 no.4
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• pp.227-234
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• 2012

In this study, we make simulation of density log using a Monte Carlo N-Particle (MCNP) algorithm to make an analysis on density logging under different borehole environments, since density logging is affected by various borehole conditions like borehole size, density of borehole fluid, thickness and type of casing, and so on. MCNP algorithm has been widely used for simulation of problems of nuclear particle transportation. In the simulation, we consider the specific configuration of a tool (Robertson Geologging Co. Ltd) that Korea institute of geoscience and mineral resources (KIGAM) has used. In order to measure accurate bulk density of a formation, it is essential to make a calibration and correction chart for the tool under considerations. Through numerical simulation, this study makes calibration plot of the density tool in material with several known bulk densities and with boreholes of several different diameters. In order to make correction charts for the density logging, we simulate and analyze measurements of density logging under different borehole conditions by considering borehole size, density of borehole fluid, and presence of casing.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.735-740
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• 1998

방사성동위원소를 이용한 아스팔트함량 측정장비의 실험적인 방법에 의한 설계는 많은 시간과 비용이 소요되므로, 코드모사를 통해 설계할 경우 이러한 노력을 줄일 수 있다. 본 연구에서는 장비의 활용성을 증대시키기 위해 법적 규제 면제치인 100 $\mu$Ci이하의 방사성동위원소를 이용하며, 6%의 아스팔트함량을 갖는 혼합물을 5분간 측정하였을 경우 0.2%이내의 함량측정오차를 갖는 장비를 MCNP 코드를 이용하여 설계하였다 또한 코드 모사를 통한 설계를 바탕으로 장비를 제작한 후 5개의 시료에 대한 함량을 측정하고 그 결과를 비교하여 코드의 적용가능성을 검증하였다 실험결과 6.03% 아스팔트 함량을 가진 시료를 5분간 측정하여 5.85%의 함량을 얻을 수 있었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.139-146
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• 1996

도로건설시 다짐조절은 안정성과 내구성 향상에 중요한 의미를 가지며 이러한 다짐조절에 있어서 수분함량의 측정은 매우 중요하다. 이전에는 흙의 수분함량을 측정하기 위한 계기를 설계하기 위하여 주로 실험에 의한 방법을 사용하였으나 본 연구에서는 3차원 모델링이 가능한 MCNP코드$^{(1)}$ 를 이용하여 계측기 설계에 있어서 중요한 설계변수인 방사선원의 위치와 측정계기 사이의 거리 그리고 계기구성요소인 검출기의 위치, 개수, 흡수재, 감속재의 기하학적 구조 등을 계산하여 설정하였다.

• Progress in Medical Physics
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• v.19 no.1
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• pp.21-34
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• 2008

The parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2006)] for calculating electron and photon beam doses has been developed based on the three dimensional geometry defined by computed tomography (CT) images and implemented on the Beowulf PC cluster. Understanding the limitations of Monte Carlo codes is useful in order to avoid systematic errors in simulations and to suggest further improvement of the codes. We evaluated the PMCEPT code by comparing its normalized depth doses for electron and photon beams with those of MCNP5, EGS4, DPM, and GEANT4 codes, and with measurements. The PMCEPT results agreed well with others in homogeneous and heterogeneous media within an error of $1{\sim}3%$ of the dose maximum. The computing time benchmark has also been performed for two cases, showing that the PMCEPT code was approximately twenty times faster than the MCNP5 for 20-MeV electron beams irradiated on the water phantom. For the 18-MV photon beams irradiated on the water phantom, the PMCEPT was three times faster than the GEANT4. Thus, the results suggest that the PMCEPT code is indeed appropriate for both fast and accurate simulations.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.61-67
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• 2005

The criticality analysis method using HELIOS-MASTER code system, which is the nuclear core analysis code system, was developed for the spent fuel storage of SMART-P reactor. We generated the macroscopic cross section of the geometric model with HELIOS and estimated the criticality of the 3-dimensional model with MASTER for SMART-P spent fuel storage. The validity of criticality analysis method for SMART-P spent fuel storage with the HELIOS-MASTER code system by 3-D MCNP calculation was also verified. The result of the criticality analysis with the HELIOS-MASTER code system is more conservative than that with the MCNP and the accuracy of this result is within the range of an allowable error. Because HELIOS-MASTER can perform the 3-D depletion calculation lot a spent fuel storage, it will be useful to perform the criticality analysis including a burnup credit in future.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1610-1616
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• 2017

To validate the new Evaluated Nuclear Data File $(ENDF)/B-VIII.0{\beta}4$ library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the $ENDF/B-VIII.0{\beta}4$ library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.