• 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.

• Journal of Radiation Protection and Research
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• 제41권3호
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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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• 제52권5호
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• pp.1002-1007
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• 2020

The purpose of the current study was to evaluate a spectrum formulation set employed to modify the neutron spectrum of D-D fusion neutrons in a IS plasma focus device using GEANT4, MCNPX2.6, and FLUKA codes. The set consists of a moderator, reflector, collimator and filters of fast neutron and gamma radiation, which placed on the path of 2.45 MeV neutron energy. The treated neutrons eliminate cancerous tissue with minimal damage to other healthy tissue in a method called neutron therapy. The system optimized for a total neutron yield of 10⁹ (n/s). The numerical results indicate that the GEANT4 code for the cubic geometry in the Beam Shaping Assembly 3 (BSA3) is the best choice for the energy of epithermal neutrons.

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.225-230
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• 2006

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of variance reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered in the areas of source definition, skyshine, streaming, and transmission.

• 한국방사선학회논문지
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• 제11권4호
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• pp.197-203
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• 2017

의료용 선형가속장치의 두부 구성요소 중 광자 발생의 원인이 되는 타깃에 대한 연구로써, 타깃의 재질에 따른 광자를 분석하여 타깃 재질 별 발생하는 광자특성에 대한 기초자료를 제시하고자 한다. 본 연구에서는 몬테카를로 방식을 바탕으로 한 MCNPX를 사용하여 타깃 재질에 따른 6, 15 MV의 광자 특성을 비교분석하였다. 타깃 재질 별 평균에너지는 6 MV에서 1.69 ~ 1.84 MeV, 15 MV에서는 3.38 ~ 3.56 MeV로 분석되었다. Flux는 6 MV에서 $1.64{\times}10^{-5}{\sim}1.80{\times}10^{-5}{\sharp}/cm^2/e$, 15 MV는 $1.76{\times}10^{-4}{\sim}1.85{\times}10^{-4}{\sharp}/cm^2/e$로 계산되었다. 결과를 분석하면, 타깃 재질이 고원자번호일수록 평균에너지와 Flux가 증가하는 것으로 평가다. 본 연구를 바탕으로 광자의 물리적 특성에 대한 기초적인 자료를 제시할 수 있었으며, 추후 타깃 선정 시 경제성, 효율성은 물론 물리적 측면을 고려할 수 있어 적절한 선택을 할 수 있을 것으로 판단된다.

• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.111-115
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• 2016

Background: The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. Materials and Methods: The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. Results and Discussion: The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination. Conclusion: It was shown that the grid of inclined type had better performance than that of parallel one.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4671-4678
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• 2022

The purpose of this comprehensive research is to observe the impact of scintillator crystal type on entire detection process. For this aim, MCNPX (version 2.6.0) is used for designing of a physical plastic scintillation detector available in our laboratory. The modelled detector structure is validated using previous studies in the literature. Next, different types of plastic scintillation crystals were assessed in the same geometry. Several fundamental detector properties are determined for six different plastic scintillation crystals. Additionally, the deposited energy quantities were computed using the MCNPX code. Although six scintillation crystals have comparable compositions, the findings clearly indicate that the crystal composed of PVT 80% + PPO 20% has superior counting and detecting characteristics when compared to the other crystals investigated. Moreover, it is observed that the highest deposited energy amount, which is a result of the highest collision number in the crystal volume, corresponds to a PVT 80% + PPO 20% crystal. Despite the fact that plastic detector crystals have similar chemical structures, this study found that performing advanced Monte Carlo simulations on the detection discrepancies within the structures can aid in the development of the most effective spectroscopy procedures by ensuring maximum efficiency prior to and during use.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1349-1354
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• 2018

Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES) is widely used for the determination of trace elements in geological samples in Iran. In this paper, we have calculated the detection limits of neutron activation analysis (NAA) for some of the common elements in such samples utilizing the ORIGEN and MCNP codes and verified the simulations using the experimental results of three soil standard reference materials, namely, G02.SRM, G18.SRM, and G28.SRM. The results show that while the detection limit of ICP-AES method is usually in the mg/kg range, it is represented to the ${\mu}g/kg$ range for most of the elements of interest using the NAA method, and the simulations can be verified in a tolerance range of 20%.

• Radiation Oncology Journal
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• 제28권4호
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• pp.224-230
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• 2010

목적: 본 연구에서는 콘빔 CT에서 산란선 제거를 위한 aft-multple-slit (AMS) 시스템을 설계하였다. 예비 연구로서 본 시스템의 효용성을 검증하기 위해 MC 시뮬레이션을 수행하였다. 대상 및 방법: 가상 시뮬레이션은 산란선과 산란선+일차선을 계산할 수 있는 MCNPX의 radiography tally 5를 이용하였다. AMS는 빔의 발산성을 고려한 각이 동일한 아크 형태이고, 길이 방향에서의 산란선을 막는다. AMS의 효용성을 위한 평가는 AMS를 사용하지 않았을 때의 일차선과 산란선을 비교함으로써 수행되었다. 2D projection 영상을 얻기 위해 전체의 AMS는 한번의 캔트리 회전 후 AMS에 의해 가려진 부분의 영상 획득을 위해 다시 한 번 회전하는 구조이다. 결과: 일차선의 2D projection 영상은 모든 AMS의 폭에서 그리고 AMS를 사용하지 않았을 때에도 동일하였으나 일차선+산란선의 2D projection 영상은 slit의 폭에 따라 결과가 변했다. Slit의 폭을 5 mm, 10 mm, 15 mm, 20 mm로 하였을 때 평균 산란성 제거율은 29%, 15%, 9%, 8%였다. 결론: 본 연구에서는 AMS를 이용한 콘빔 CT의 산란선 제거 효과를 평가하였다. MC 시뮬레이션을 이용한 본 시스템의 사전 연구에서는 상당한 산란선 제거 효과를 보여주었다.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.