• Title, Summary, Keyword: MCNP code

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TET2MCNP: A Conversion Program to Implement Tetrahedral-mesh Models in MCNP

  • Han, Min Cheol;Yeom, Yeon Soo;Nguyen, Thang Tat;Choi, Chansoo;Lee, Hyun Su;Kim, Chan Hyeong
    • 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.

Application of a new neutronics/thermal-hydraulics coupled code for steady state analysis of light water reactors

  • Safavi, Amir;Esteki, Mohammad Hossein;Mirvakili, Seyed Mohammad;Arani, Mehdi Khaki
    • 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.

COMPARISON OF CANDU DUPIC PHYSICS CODES WITH MCNP

  • Gyuhong Roh;Park, Hangbok
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.65-70
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    • 1997
  • Computational benchmark calculations have been performed for CANDU DUPIC fuel lattice and core using a Monte Carlo code MCNP-4B with ENDF/B-V library. The eigenvalues of the DUPIC fuel lattice have been predicted by an integral transport code WIMS-AECL using ENDF/B-V library for different burnup steps and lattice conditions. The comparison has shown that the eigenvalues match those of MCNP-4B within 0.20% $\Delta$k difference between WIMS-AECL and MCNP-4B results. The calculation of a 2-dimensional CANDU core loaded with DUPIC fuel has shown that the eigenvalue predicted by a diffusion code RFSP using lattice parameters generated by WIMS-AECL matches that of MCNP-4B within 0.12%Δk and the largest bundle power prediction error is around 7.2%.

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Neutron Shielding Analysis for a Spent Fuel Container Using XSDRN, ONEDANT and MCNP Codes (XSDRN, ONEDANT및 MCNP에 의한 사용후 핵연료 용기의 중성자 차폐 해석)

  • Kim, Kyo-Youn;Lee, Tae-Young;Ha, Chung-Woo;Kim, Jong-Kyung
    • Journal of Radiation Protection and Research
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    • v.14 no.1
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    • pp.46-55
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    • 1989
  • Neutron shielding for a spent fuel container was analized using the Monte Carlo code MCNP coupled with discrete ordinates codes, XSDRN and ONEDANT. The ORIGEN-S code was used to determine the fixed neutron source, and the spectral source information for MCNP were obtained from a 10 group XSDRN calculation and a 27 group ONEDANT calculation. When the depleted uranium shield was used, the results from ONEDANT and XSDRN calculations agreed with the MCNP results within 10% and 20%, respectively. Depleted uranium appears more effective than lead or steel as a neutron shield.

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Calculations of Radiation Measurement-Related Correction Factors (방사선 측정관련 보정인자 계산)

  • Shin, Hee-Sung;Ro, Seung-Gy;Kim, Ho-Dong
    • Journal of Radiation Protection and Research
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    • v.28 no.1
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    • pp.19-24
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    • 2003
  • The self-attenuation factor for an $^{198}Au$ sample and the 0.412 MeV gamma-ray penetration ratio in the circular Al-cover of the radiation detector have been determined using an analytical solution and MCNP code. The results show that the self-attenuation factors obtained from the analytical solution coincide with those of MCNP code for all but the Au sample with the relatively larger radius. Then the maximum difference between the two methods appears to be 9 % in the Au sample of 1.5 mm radius. It also is revealed that the analytical solutions of the 0.412 MeV gamma-ray penetration ratio in the Al-cover of 7.62 cm radius are consistent with those of the MCNP code within the standard deviation.

Gamma Ray Shielding Study of Barium-Bismuth-Borosilicate Glasses as Transparent Shielding Materials using MCNP-4C Code, XCOM Program, and Available Experimental Data

  • Bagheri, Reza;Moghaddam, Alireza Khorrami;Yousefnia, Hassan
    • 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.

Power Density Distribution Calculation of a Pressurized Water Reactor with Fullscope Explicit Modeling by MCNP Code

  • Kim, Jong-Oh;Kim, Jong-Kyung
    • Proceedings of the Korean Nuclear Society Conference
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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.

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Calculation of gamma buildup factors for point sources

  • Kiyani, Abouzar;Karami, Abbas Ali;Bahiraee, Marziye;Moghadamian, Hossein
    • 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%.

A Theoretical Consideration about Effects of Radiation on the Physical Properties of PP (PP 재질의 물성에 미치는 방사선의 영향에 대한 이론적 고찰)

  • 김문수;강덕원;엄희문
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.517-523
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    • 2003
  • The physical properties of polypropylene (PP) membranes under the radiation field were investigated. To calculate radiation flux affecting to PP, it was used MCNP4A Code. The PP membrane and deoxygenation equipment were standardized to bar structure in order to calculate the phonton flux with MCNP4A Code. The change in the properties of the PP membrane to be used in deoxygenation equipment was rarely occurred during the usage work because the radiation level of reactor coolant water was very low level and The doses of radiation workers are very low. From the results, it was found that the Physical properties of PP membranes which used for nuclear power plant reactor coolant water disposal were not rarely changed under the simulated radiation field.

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