• 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.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1994-1998
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• 2023

This study quantitatively evaluated the source term of a linear accelerator according to target thickness for a 6-20 MeV electron beam using MCNP6. The elements of the target were tungsten and copper, and a composite target and single target were simulated by setting different thickness parameters depending on energy. The accumulation of energy generated through interaction with the collided target was evaluated at 0.1-mm intervals, and F6 tally was used. The results indicated that less than 3% reference error was maintained according to the MCNP recommendations. At 6, 8, 10, 15, 18, and 20 MeV, the energy accumulation peaks identified for each target were 0.3 mm in tungsten, 1.3 mm in copper, 1.5 mm in copper, 0.5 mm in tungsten, 0.5 mm in tungsten, and 0.5 mm in tungsten. For 8 and 10 MeV in a single target consisting only of copper, the movement of electrons was confirmed at the end of the target, and the proportion of escaped electrons was 0.00011% and 0.00181%, respectively.

• Journal of Radiation Protection and Research
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• v.19 no.1
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• pp.13-22
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• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• 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.

• Journal of Electrical Engineering and Technology
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• v.6 no.1
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• pp.139-146
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• 2011

This paper presents a new modified Contract Net Protocol (CNP) for microgrid operation based on multiagent system. The CNP is a widely used protocol for interactions among distributed problem solving. The Contract Net Interaction Protocol of the Foundation for Intelligent Physical Agents (FIPA-CNIP) is a minor modification of the original CNP for multiagent system applications. In this paper, a modified CNP (MCNP) based on the FIPA-CNIP is proposed for more specialized interactions among agents for microgrid operation. A multiagent system is designed and constructed for microgrid operation. A microgrid operation based on the multiagent system is tested to check the functionality of the proposed MCNP.

• 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.