• 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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.269-274
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• 1996

월성원자력발전소 2호기와 같은 CANDU 6형 원자로의 반응도제어기구 설치대에는 여러 반응도제어기구가 삽입되기때문에 원자로심으로부터의 방사선흐름현상으로 인한 방사선피폭이 예상될 수 있는 위치이다. 좁고 긴 반응도제어기구 도관에서의 방사선 흐름으로 인한 반응도제어기구 설치대에서의 방사선량을 예측하기 위해 몬테 칼로 MCNP 코드를 1차원 각분할법 코드인 ANISN과 연계하여 사용하였다. 월성원자력2호기의 상단차폐해석을 위한 ANISN 계산, 도관의 방사선흐름을 평가하기 위한 MCNP 계산, 그리고 반응도제어기구 설치대에서의 방사선량율 평가를 위한 MCNP 계산등 3단계 계산 기법의 적응이 시도되었다.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.4
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• pp.351-359
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• 2009

We have developed a computer simulator fur cone-beam computed tomography (CBCT) based on the commercial Monte Carlo code, MCNP. All the functions to generate input files, run MCNP, convert output files to image data, reconstruct tomographs were realized in graphical user-interface form. The performance of the simulator was demonstrated by comparing with the experimental data. Although some discrepancies were observed due to the ignorance of the detailed physics in the simulation, such as scattered X-rays and noise in image sensors, the overall tendency was well agreed between the measured and simulated data. The developed simulator will be very useful for understanding the operation and the better design of CT systems.

• 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

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.45-52
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• 1996

고리 1 호기 원자로 감시？슐에서의 고속중성자 플루언스를 계산하기 위하여 MCNP코드로 3차원 모델링을 하였다. 중성자선원은 MCNP에 의해 계산된 핵연료봉출력분포를 사용하였고 원자로심부터 원자로 압력용기 밖까지 1 MeV이상의 중성자를 수송시켰다. 감시？슐은 실제의 같은 크기로 모델링하였고 감시？슐안의 시편은 원자로 압력용기와 같은 재질의 직육면체로 가정하였다. 그리고 MCNP에 의해 감시시편내의 방사화 시료의 핵반응단면적을 계산하였다. 또한 MCNP에 의해 이론적으로 계산된 감시？슐에서 중성자 플루언스와 기존의 감시시험에서 측정된 포화방사능으로 부터 계산된 실험적 감시？슐 중성자 플루언스를 비교하였다. 이론적 ？슐플루언스와 실험적 ？슐플루언스의 비는 대체로 1.0에서 크게 벗어나지 않았으나 감시시험과 시편에 따라 크게 벗어나는 경우도 있었다. MCNP에 의한 유효반응단면적의 계산방법이 기존의 방법보다 모델링 및 계산의 불확실성을 최소화 할 수 있으므로 이번 연구에서 고려하지 못한 원자로심의 연소도를 고려한다면 매우 신뢰성이 높은 결과를 얻을 수 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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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%.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4426-4430
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• 2023

In this article, Exposure Buildup Factor(EBF) and the Energy Absorption Buildup Factor(EABF) have been determined for blood, brain, and muscle using the Monte Carlo method which is represented by MCNP5 codes and compared with geometric progression(G-P) fitting method which is represented by Phy-X/PSD online platform. The novelty of the present work is used an energy source of less than 0.1 MeV to determine buildup factors using MCNP5 and using Phy-X/PSD for some human tissues. thus, the energy range used in this case study was 0.06-3 MeV for penetration depths covered 0.5-3 MFP. Results of MCNP5 and Phy-X/PSD are validated against reference values of water that were reported at ANS-6.4.3. present results of EABFs and EBFs for the previously mentioned human tissues appeared good agreement between MCNP5 in comparison with Phy-X/PSD, whereas, the maximum average relative deviation did not exceed 2.37%. results of our article can be used in different medical applications, such as brachytherapy, radiotherapy, and diagnostics.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.21-34
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• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.