• Nuclear Engineering and Technology
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• 제48권3호
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• pp.642-649
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• 2016

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.151-158
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• 2018

The neutron powder diffractometer (NPD) is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo-based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo-based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than $10^6n/s/cm^2$ at sample position.

• 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.

• 한국산학기술학회논문지
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• 제11권3호
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• pp.1035-1039
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• 2010

고에너지 X-ray를 투시 방사선원으로 사용한 컨테이너 화물검색시스템에서 생성되는 광중성자에 의해 주변 콘크리트 차폐벽에서 발생되는 방사화생성물을 평가하였다. 몬테카를로 전산해석 코드인 MCNPX2.5.0을 사용하였으며, 참조시스템은 국내 주요 항만에 설치된 9MeV X-ray 고정식 양방향 컨테이너 화물검색시스템이다. 9MeV X-ray 조사에 따라 생성되는 광중성자의 (n,$\gamma$) 반응에 의한 방사화생성물 재고량을 계산하고 이에 따라 야기되는 방사선 피폭선량을 계산하였다.

• 한국안전학회지
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• 제32권6호
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• pp.29-33
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• 2017

In order to solve this sort of electrostatic failure in Display and Semiconductor process, Soft X-ray ionizer is mainly used. Soft X-ray Ionizer does not only generate electrical noise and minute particle but also is efficient to remove electrostatic as it has a wide range of ionization. There exist variable factors such as type of tungsten thickness deposited on target, Anode voltage etc., and it takes a lot of time and financial resource to find optimal performance by manufacturing with actual X-ray tube source. Here, MCNPX (Monte Carlo N-Particle Extended) is used for simulation to solve this kind of problem, and optimum efficiency of X-ray generation is anticipated. In this study, X-ray generation efficiency was compared according to target material thickness using MCNPX and actual X-ray tube source under the conditions that tube voltage is 5 keV, 10 keV, 15 keV and the target Material is Tungsten(W). At the result, In Tube voltage 5 keV and distance 100 mm, optimal target thickness is $0.05{\mu}m$ and fastest decay time appears + decay time 0.28 sec. - deacy time 0.30 sec. In Tube voltage 10keV and distance 100 mm, optimal target Thickness is $0.16{\mu}m$ and fastest decay time appears + decay time 0.13 sec. - deacy time 0.12 sec. In the tube voltage 15 keV and distance 100 mm, optimal target Thickness is $0.28{\mu}m$ and fastest decay time appears + decay time 0.04 sec. - deacy time 0.05 sec.

• Journal of Radiation Protection and Research
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• 제37권2호
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.55-63
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• 2016

Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy ($E_e$) and source multiplication coefficient ($k_s$), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such as effective multiplication coefficient ($k_{eff}$), net neutron multiplication (M), neutron yield ($Y_{n/e}$), energy constant gain ($G_0$), energy gain (G), importance of neutron source (${\varphi}^*$), axial and radial distributions of neutron flux, and power peaking factor ($P_{max}/P_{ave}$) in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current ($I_e$) have been decreased in the highest case of $k_s$, but G and ${\varphi}^*$ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing $E_e$ from 100 MeV up to 1 GeV, $Y_{n/e}$ and G improved by 91.09% and 10.21%, and $I_e$ and $P_{acc}$ decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np-Pu assemblies on the periphery allows for a consistent $k_{eff}$ because the Np-Pu assemblies experience less burn-up.

• 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.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• 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.