• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.363.1-363.1
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• 2003

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.216-223
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• 2017

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and $10^{th}$ value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.49.2-49
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• 2002

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.11a
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• pp.379-382
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• 2002

Lately, intensifying screen of the $CaWO_4$ is used to medical treatment and diagnosis of the image. In this paper, we investigated transmission fraction and mass attenuation coefficient of $CaWO_4$ screen about diagnostic x-ray of low energy using MCNP 4C code. Experimentally, for 0.9 mm-$CaWO_4$ screen, the absorbable rate of diagnostic x-ray is more than 95%. according to kVp, the experimental value of mass attenuation coefficient is in a1most agreement with an corrected estimate value of MCNP and the deviation of experimental values is less than ${\pm}7%$. Using the MCNP code through this paper, we can make an estimate of signal and design for construction of the CaWO4/a-Se based digital x-ray image detector and make a good use of the foundation data for development of other materials.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, 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 peformed 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 $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ 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. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.389-394
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• 2016

Background: Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. Materials and Methods: TET2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. Results and Discussion: To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. Conclusion: In the present study, we have developed a computer program, TET2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.

• Advances in materials Research
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• v.2 no.2
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• pp.93-98
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• 2013

Objective of this study is to calculate gamma buildup factors for pointed and isotropic gamma sources in depleted uranium, uranium dioxide, natural uranium, tin, water and concrete using MCNP4C code. The thickness of the media ranges from 0.5 to 10 mean-free-path (mfp) and gamma energy ranges from 0.5 to 10 MeV. Owing to the outstanding accuracy of MCNP in calculation involving gamma interaction, results fairly match those reported previously. The maximum relative error is 2%.

• 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.

• 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.