• 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.

• 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%.

• 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.

• 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.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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.79-84
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• 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
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• v.41 no.3
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• pp.191-195
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• 2016

Background: The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Materials and Methods: Validation calculations with recent nuclear data evaluations ENDF/BV-II. 0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. Results and Discussion: It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and ${chi}^2$ values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the $k_{eff}$ values. Conclusion: It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3528-3542
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• 2021

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were <3%. When studying axial power density profiles the differences in axial offset were less than 2.3% for hot full power condition. To further confirm the applicability of the developed model, the measurements with in-core neutron detectors were compared to the calculations, where differences of 5% were observed.

• 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.

• Journal of Radiation Protection and Research
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• v.32 no.1
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• pp.21-26
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• 2007

This study is performed to evaluate the dose rate and to analyze the dose distribution of the gamma irradiation facility (IR-221) by using a Monte Calro simulation, which is helpful of upgrading the radiation processing qualification. Monte Cairo simulation is performed by MCNP4B code. Dose rates were measured at total 369 points with alanine dosimeters to compare the calculation results and the measurements data. The results have shown that the MCNP4B code is very useful to determine the dose distribution of the IR-221 gamma irradiation facility, as the calculation dose rate is within about ${\pm}5%$ of the measurement data. Dosimetry about the gamma irradiation facility usually needs enormous manpower and time. However Monte Cairo calculation method can reduce the tedious dosimetry jobs and improve the irradiation processing qualification, which will probably contribute to obtain the reliability of the irradiation products.