• Nuclear Engineering and Technology
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• 제52권8호
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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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• 제54권9호
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• pp.3429-3439
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• 2022

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this paper, the radiation shielding capability of the MCS code is validated by simulating some selected SINBAD (Shielding Integral Benchmark Archive and Database) experiments. The whole validation was performed in two ways. Firstly, the functionality and computational rationality of the MCS code was verified by comparing the simulation results with those of MCNP code. Secondly, the validity and computational accuracy of the MCS code was confirmed by comparing the simulation results with the experimental results of SINBAD. The simulation results of the MCS code are highly consistent with the those of the MCNP code, and they are within the 2σ error bound of the experiment results. It shows that the calculation results of the MCS code are reliable when simulating the radiation shielding problems.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1265-1268
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• 2023

One important issue in using radiopharmaceuticals as therapeutic and imaging agents is predicting different organ absorbed dose following their injection. The present study aims at extrapolating dosimetry estimates to a female phantom from the animal data of ⁸⁹Zr radionuclide accumulation using the Sparks-Idogan relationship. The absorbed dose of ⁸⁹Zr radionuclide in different organs of the human body was calculated based on its distribution data in mice using both MIRD method and the MCNP simulation code. In this study, breasts, liver, heart wall, stomach, kidneys, lungs and spleen were considered as source and target organs. The highest and the lowest absorbed doses were respectively delivered to the liver (4.00E-02 and 3.43E-02 mGy/MBq) and the stomach (1.83E-03 and 1.66E-03 mGy/MBq). Moreover, there was a good agreement between the results obtained from both MIRD and MCNP methods. Therefore, according to the dosimetry results, [⁸⁹Zr] DFO-CR011-PET/CT seems to be a suitable for diagnostic imaging of the breast anomalies for CDX-011 targeting gpNMB in patients with TNBC in the future.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1828-1834
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• 2019

The performance of a plastic scintillator for use in an in situ measurement system was analyzed using simulation and experimental methods. The experimental results of four major pure beta-emitting radionuclides, namely $^3H$, $^{14}C$, $^{32}P$, and $^{90}Sr/^{90}Y$, were compared with those obtained using a Monte Carlo N-particle (MCNP) code simulation. The MCNP simulation and experimental results demonstrated good agreement for $^{32}P$ and $^{90}Sr/^{90}Y$, with a relative difference of 1.95% and 0.43% between experimental and simulation efficiencies for $^{32}P$ and $^{90}Sr/^{90}Y$, respectively. However, owing to the short range of beta particles in water, the efficiency for $^{14}C$ was extremely low, and $^3H$ could not be detected. To directly measure the low-energy beta radionuclides considering their short range, a system where the source could flow directly to the scintillator was developed. The optimal thickness of the plastic scintillator was determined based on the suggested diameter. Results showed that the detection efficiency decreases with an increase in the depth of the water. The detection efficiency decreased drastically to approximately 10 cm, and the tendency was gradually constant.

• Journal of Radiation Protection and Research
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• 제32권1호
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• pp.21-26
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• 2007

본 논문은 몬테칼로 시뮬레이션을 이용하여 대단위 감마선 조사시설 (IR-221)에 대한 선량률 평가 및 선량 분포를 해석하고, 이러한 방법을 통해 방사선 조사 품질을 향상시키는 것을 목적으로 하고 있다. 몬테칼로 시뮬레이션은 MCNP4B 코드를 이용하여 계산하였고, 이를 검증하기 위해 알라닌 선량계를 이용하여 전체 309개 지점에 대하여 흡수선량을 측정하였다. 계산 값과 측정치의 차이는 대략 ${\pm}5%$범위를 벗어나지 않음으로써 MCNP4B 코드가 IR-221 감사선 조사시설의 선량분포를 해석하는데 있어서 유효한 수단임을 알 수 있었다. 감마선 조사시설에 대한 도시메트리는 보통 많은 인력과 시간을 필요로 하지만, 몬테칼로 계산을 통해 이러한 손실을 줄일 수 있고, 무엇보다도 방사선 조사 품질을 향상시켜, 결국 방사선 조사 대상물에 대한 신뢰도를 확보하는 데에도 이바지 할 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2661-2668
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• 2021

Gamma radiation shielding features for three series of binary alloys identified as (Pb-Sn), (Pb-Zn), and (Zn-Sn) have been investigated. The mass attenuation coefficients (µ/ρ) for the selected alloys were simulated using the MCNP-5 code in the energy range between 0.01 and 15 MeV. Moreover, the (µ/ρ) values were computed using WinXCOM database in the same energy range to validate the simulation results. Results reveal a good agreement between the simulated and computed values. The half value layer (HVL), mean free path (MFP), effective atomic number (Z_eff) and exposure buildup factor (EBF) were evaluated for the selected binary alloys. Results showed that the PS1, PZ1, and ZS2 alloys have the best shielding parameters and better than the commercially standard and available radiation shielding materials. Therefore, the investigated alloys can be used as effective radiation shielding materials against gamma ray with energies between 0.01 and 15 MeV.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2006년도 제33회 추계학술대회 발표논문집
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• pp.67-67
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• 2006

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2002년도 Proceedings
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• pp.485-488
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• 2002

In order to understand and quantitatively analyze the physical phenomena and behavior of each component of mammography system during the breast imaging, we simulated mammography imaging using Monte Carlo simulation codes. MCNP4B code was used for our simulation purpose, and we investigated the effect of target material, anode angle, filtration, peak voltage and exposure on the image quality of mammograms. From the simulation results we expect that optimized operation condition of mammography system can be found.

• 지구물리와물리탐사
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• 제15권4호
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• pp.219-226
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• 2012

이 연구에서는 국내에서 연구가 미약했었던 중성자검층 수치모델링을 이용하여 다양한 시추공 환경에서의 검출기 반응을 분석하였다. 이를 위해 중성자검층 환경을 MCNP 알고리듬으로 구현하여 시뮬레이션을 수행하였다. MCNP 알고리듬은 방사선 수송 시뮬레이션이 및 3차원 기하구조 표현이 가능하여 다양한 분야에서 전세계적으로 많이 이용되고 있다. 먼저 시뮬레이션 결과를 검증하기 위해, 기존 연구의 검출기반응 결과 그래프를 이용하여 비교 분석하였다. 중성자 검층 반응 분석이 가능한 중성자 검층기의 일반적인 특징에 기초하여 수학적으로 중성자검층기 모형을 구성하여 반응을 계산하였다. 먼저, 석회암, 사암, 돌로마이트 등과 같은 매질에서 공극률을 다양하게 변화시켜 가며 수치 계산함으로써, 이 연구에서 고려하고 있는 중성자검층기의 교정곡선(calibration chart)을 도출하였다. 이에 기초하여, 실제 중성자검층 시 고려해야 할 공내수 유무에 의한 반응 변화, 염수가 중성자검층에 미치는 영향 등을 분석함으로써 시추공 환경 변화에 따라 보다 정확하게 공극률을 해석할 수 있는 기틀을 마련하고자 한다.

• 방사성폐기물학회지
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• 제13권1호
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• pp.21-34
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• 2015

원자력발전소 해체를 준비하기 위해서는 해체대상 발전소에 대한 선원항 평가가 선행되어야 한다. 해체전략 수립단계에서 선원항 평가 결과를 토대로 해체 폐기물을 분류하고 비용평가를 수행한다. 본 연구에서는 월성 1호기의 예비 선원항 계산을 수행할 수 있도록 MCNP/ORIGEN-2 모델의 타당성 평가를 수행하였다. 연소도가 다른 핵연료 다발의 악티나이드 계열과 핵분열 생성물의 핵종 수밀도는 싱글 채널 모델을 이용하여 MCNPX 코드로 연소 계산하여 구하였다. 선원항의 정확도에 영향을 미치는 두가지 요인에 대해 조사하였다. 첫번째 요인으로 선원항 계산에 영향을 미치는 중성자 스펙트럼을 MCNP로 계산하여 해당 핵종의 1군 미시 핵단면적에 반영하였다. 중성자 스펙트럼이 반영된 라이브러리로 계산한 선원항과 ORIGEN-2 코드 package에 내장된 library (CANDUNAU.LIB)로 구한 선원항을 비교하였다. 두번째 요인으로 선원항에 대한 출력이력의 영향을 조사하였다. 해체 폐기물의 저준위 폐기물 처분 가능성을 살펴보기 위해, 2010년도 교체된 압력관, 칼란드리아관과 기존 칼란드리아 동체에 대하여 중성자 스펙트럼을 반영한 library를 적용하여 MCNP/ORIGEN-2로 선원항 평가 계산을 수행하였다.