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

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.65-70
• /
• 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%.

• Journal of Radiation Protection and Research
• /
• v.41 no.4
• /
• pp.389-394
• /
• 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
• /
• v.22 no.4
• /
• pp.289-298
• /
• 1997

Time-consuming experiments have been required in the development of neutron moisture gauge to induce a relation between the water content in soil and the neutron counts. Applying a monte carlo computer code to simulate the experiments of neutron moisture gauging may contribute to reduce time and efforts for experiments and produce a calibration equation which is more applicable to soil in general. In this study MCNP4A, a monte carlo computer code, was employed to simulate soil experiments and the simulated results were compared with experimental ones. The comparative study showed that MCNP4A is applicable to simulate the experiments and calibration equation can be obtained through simulations. Effects of dry density changes were also studied.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1603-1610
• /
• 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
• /
• 1995.10a
• /
• pp.79-84
• /
• 1995

핵연료집합체 연소도 계산용 전산코드인 CASMO-3를 도입하여 한국고유핵설계체계를 개발하기 위해서는 CE형 핵연료집합체의 핵적특성을 파악하는 것은 필수적이다. 따라서, CASMO-3와 몬테칼로 전산코드인 MCNP-4A를 이용하여 CE형 16$\times$16 핵연료집합체에 대한 $K_{inf}$ 및 봉출력 분포를 비교 분석하였다. $K_{inf}$ 의 경우는 CASMO-3에 의한 계산 결과가 0.5% 이내에서 MCNP-4A의 계산 결과와 일치하였으며, 봉출력분포의 경우도 제어봉 주변이나 Gd$_2$O$_3$ 독봉을 제외하고는 CASMO-3에 의한 계산 결과가 MCNP-4A의 계산 결과와 거의 일치하는 것으로 나타났다.

• Journal of Radiation Protection and Research
• /
• v.21 no.2
• /
• pp.125-129
• /
• 1996

The energy response of HPGe detector for low energy Photons was determined by using three Monte Carlo codes. MCNP4A. EGS4, and CYLTRAN in ITS3. In this study. bare HPGe detector$(100 mm^2{\times}10mm)$ was used and a pencil beam was incident perpendicularly on the center of the detector surface. The photopeak efficiency, $K_{\alpha}$ and $K_{\beta}$ escape fractions were calculated as a function of incident X-ray energies ranging from 12 to 60 keV in 2-keV increments. Since the Compton. elastic. ana penetration fraction were negligible in this energy range. they were ignored in the calculation. Although MCNP. EGS, and CYLTRAN codes calculated slightly different energy response of HPGe detector for low energy Photons, it appears that the three Monte Carlo codes can Predict the low energy Photon scattering Processes accurately. The MCNP results, which are generally known as to be less accurate at low energy ranges than the EGS and ITS results. are comparable to the results of EGS and ITS and are applicable to the calculation of the low energy response data of a detector.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.397-397
• /
• 1999

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

• Korean Journal of Environmental Biology
• /
• v.24 no.4
• /
• pp.387-391
• /
• 2006

Computational and experimental dosimetry of Henschke applicator with respect to high dose rate brachytherapy using the MIRD phantom and a remote control afterloader were performed. A comparison of computational dosimetry was made between the simulated Monte Carlo dosimetry and GAMMADOT brachytherapy Planning system's dosimetry. Dose measurements was performed using ion chamber in a water phantom. Dose rates are calculated using Monte Carlo code MCNP4B and the GAMMADOT. Thecomputational models include the detailed geometry of Ir-192 source, tandem tube, and shielded ovoids for accurate estimation. And transit dose delivered during source extension to and retraction from a given dwell position was estimated by Monte Carlo simulations. Point doses at ICRU bladder/rectal pointswhich have been recommened by ICRU 38 was assessed. Calculated and measured dose distribution data agreed within 4% each other. The shielding effect of ovoids leads to 19% and 20% dose reduction at bladder surface and rectal points.