• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.21-34
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• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

• Journal of the Korean Society of Radiology
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• v.16 no.6
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• pp.687-695
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• 2022

To find a 3D printer material that can replace lead used as a shield for high-energy electron beam treatment, the shielding composites were simulated by using MCNP6 programs. The Percent Depth Dose (PDD), Flatness, and Symmetry of linear accelerators emitting high-energy electron beams were measured, and the linear accelerator was compared with MCNP6 after simulation, confirming that the source term between the actual measurement and simulation was consistent. By simulating the lead shield, the appropriate thickness of the lead shield capable of shielding 95% or more of the absorbed dose was selected. Based on the absorption dose data for lead shield with a thickness of 3 mm, the shielding performance was analyzed by simulating 1, 5, 10, and 15 mm thicknesses of ABS+W (10%), ABS+Bi (10%), and PLA+Fe (10%). Each prototype was manufactured with a 3D printer, measured and analyzed under the same conditions as in the simulation, and found that when ABS+W (10%) material was formed to have a thickness of at least 10mm, it had a shielding performance that could replace lead with a thickness of 3mm. The surface morphology and atomic composition of the ABS+W (10%) material were evaluated using a scanning electron microscope (SEM) and an energy dispersive X-ray spectrometer (EDS). From these results, it was confirmed that replacing the commercialized lead shield with ABS+W (10%) material not only produces a shielding effect such as lead, but also can be customized to patients using a 3D printer, which can be very useful for high-energy electron beam treatment.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.841-845
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• 2020

An airborne alpha beta detection system using passivated implanted planar silicon (PIPS) detector was modeled with the MCNP6 code and its resolution and detection efficiency were analyzed. Simulation of the resolution performed using the Gaussian energy broadening (GEB) function showed that the full width at half maximum (FWHM) of 35.214 keV for alpha particles was within 34-38 KeV, which is the FWHM range of the actual detector, and the FWHM of 15.1 keV for beta particles was constructed with a similar model to 17 keV, which is the FWHM range of an actual detector. In addition, the detection efficiency and the resolution were simulated according to the distance between the detector and the air filter. When the distance was decreased to 0.2 cm from 0.8 cm, the efficiency of the alpha and beta particles detection decreased from 5.33% to 4.89% and from 5.64% to 4.27%, respectively, and the FWHM of the alpha and beta particles improved from 40.9 KeV to 29.84 keV and 25.76 keV-13.27 keV, respectively.

• Nuclear Engineering and Technology
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• v.53 no.8
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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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.115-120
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• 1997

영광3호기 방사선관리구역에 대한 중성자선량률을 정확히 평가하기 위하여 MCNP4A 전산코드를 이용, 방사선관리구역에서의 중성자 스펙트럼 계산을 수행하였다. 영광3호기에 대한 보다 정확하고 정밀한 3차원 몬테칼로 모델을 구축하기 위하여 핵연료집합체 구성요소 및 원자로심을 둘러싸고 있는 baffle, barrel,압력용기 등을 정확하게 묘사하였으며, 특히 방사선관리구역 주위의 구조물에 대해서도 3자원 MCNP 모델을 구축함으로써 원자로심부터 방사선관리구역까지 완전한 몬테칼로 모사(full-scope Monte Carlo simulation)를 이용한 계산을 수행하였다. 계산결과는 에너지 구간에 따른 중성자속 스펙트럼으로 나타내었으며 이 결과를 바탕으로 중성자속에 대한 선량률 환산인자를 고려하여 중성자선량률을 계산할 수 있다.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3012-3017
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• 2021

The spatial distribution and the energy spectrum of the neutron yield were investigated with the neutron activation analysis and MCNP simulation was carried out to verify the analysis results and to extend the results to a 3D mapping of the neutron yield distribution in the KSTAR. High purity Ni specimen was selected in the neutron activation analysis. Total neutron yields turned out to be 3.76 × 10¹² n/s - 7.56 × 10¹² n/s at the outer vessel of the KSTAR, two orders of magnitude lower than those at the inner vessel of the KSTAR, which demonstrates the attenuation of neutron yield while passing through the different structural materials of the reactor. Based on the fully expanded 3D simulation results, 2D cross-sectional distributions of the neutron yield on XY and ZX planes of KSTAR were examined. The results reveal that the neutron yield has its maximum concentration near the center of blanket and decreases with increasing proximity to the vacuum vessel wall.

• Journal of radiological science and technology
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• v.43 no.6
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• pp.469-473
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• 2020

Breast cancer is growing rapidly year by year and has the highest incidence since 2001. As a result, the interest in mammography for early detection of breast cancer is increasing. However, mammography is accompanied by radiation exposure and therefore it is necessary to reduce exposure dose through appropriate test conditions. The significance of this study is that breast dose studies, which were limited to ordinary women, were applied to breast implant patient. Using MCNP simulation, the phantom with prosthesis inserted was developed to compare dose by tube voltage by pressure thickness. In addition phantom without prostheses has higher dose than phantom with prostheses. If these results were used as basic data, it would be possible to recommend test condition guideline only for breast implant patients.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2013.05a
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• pp.1031-1033
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• 2013

Radiation protection, as well as finding the location of the radiation source, such as the Fukushima radiation leak accident, it is important for the early and safe disposal of nuclear accident. The three-dimensional position of the radiation source detection distance of the radiation source can provide additional information to the existing radiation detectors radiation of a two-dimensional position detection function and then it can play a decisive role in the radiation contaminant removal and decontamination work. In this research, three-dimensional semiconductor sensor based on dual radiation detectors radiation source device visible part of the research and development of efficient radiation sensor unit on the design of the shielding structure.The lightweight, high-efficiency radiation source locator implementation was attempted for the structure and thickness of the shielding and collimator to perform the simulation of the radiation shielding for the various parameters of the shape model through design the optimal structure of the MCNP-based heavy-duty tungsten shielding, lead shielding The results of this study, is a compact, lightweight three-dimensional radiation source detection and future of silicon - based sensors will be used in the study.