• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.333-335
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• 2001

The 2 GeV electron linac, the injector of the Pohang Light Source, was used as a photo-neutron source for radiation shielding research. The operational beam parameters are the nominal electron intensity of $0.5\;{\sim}5\;nC/sec$, the repetition rate of 10 Hz, and the beam pulse length of 1.0 nsec. One electron beam line was modified in order to install the target systems for producing pulsed photo-neutrons. The neutron spectrum and intensity were investigated by the time-of-flight technique. The reliable maximum energy of the measured neutrons was about 500 MeV. The number of neutrons above 20 MeV produced by one 1 GeV electron in a thick Pb target was about $6.45{\times}10^{-4}/sr$ at 90 degrees to the beam axis. The status of the photo-neutron source and the application research are presented.

• 한국표면공학회:학술대회논문집
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• 한국표면공학회 2011년도 춘계학술대회 및 Fine pattern PCB 표면 처리 기술 워크샵
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• pp.205-206
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• 2011

Cu thin films for electromagnetic wave shielding were prepared on PET film and Ni-coated PET film by using Dry and Wet coating method, such as evaporation method, DC sputtering method and copper sulfate($CuSO_4$). After that, Zn thin film and Ni thin film were prepared onto the Cu thin films by using evaporation dry process and Ni electro plating wet process as a finishing treatment, respectively. The result of conductivity test and corrosion resistance test revealed Cu thin films which were formed with bigger grain size and high Cu composition rate have superior properties. Zn thin film by dry evaporation process and Ni thin film by wet electro plating process on Cu thin films were largely contributed to corrosion resistance. However, Ni thin film by wet process made conductivity of all specimen worse, the other hand, Zn thin film by dry process made it better to improve condictivity of specimens just prepared by dry process.

• 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.

• 한국재료학회지
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• 제28권12호
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• pp.685-690
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• 2018

In this study, a multifunctional ophthalmic lens material with an electromagnetic shielding effect, high oxygen permeability, and high water content is tested, and its applicability is evaluated. Metal oxide nanoparticles are applied to the ophthalmic lens material for vision correction to shield harmful electromagnetic waves; the pyridine group is used to improve the antibacterial effect; and silicone substituted with urethane and acrylate is employed to increase the oxygen permeability and water content. In addition, multifunctional tinted ophthalmic lens materials are studied using lens materials with an excellent antibacterial effect (2,6-difluoropyridine, 2-fluoro-4-pyridinecarboxylic acid) and functional (UV protection, high wettability) lens materials (2,4-dihydroxy benzophenone, 2-hydroxy-4-(methacryloyloxy)benzophenone). To solve problems such as air bubbles generated during the polymerization process for the manufacturing and turbidity of the lens surface, polymerization conditions in which the defect rate is minimized are determined. The results show that the polymerization temperature and time are most appropriate when they are $110^{\circ}C$ and 40 minutes, respectively. The optimum injection amount of the polymerization solution is 350 ms. The turbid phenomenon that appears in lens processing is improved by 10 to 95 % according to the test time and conditions.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.1966-1973
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• 2023

Virtual reality technology has been widely used in the field of nuclear and radiation safety, dose rate computing in virtual environment is essential for optimizing radiation protection and planning the work in radioactive-controlled area. Because the CPU-based gamma dose rate computing takes up a large amount of time and computing power for voxelization of volumetric radioactive source, it is inefficient and limited in its applied scope. This study is to develop an efficient gamma dose rate computing code and apply into fast virtual simulation. To improve the computing efficiency of the point kernel algorithm in the reference (Li et al., 2020), we design a GPU-based computing framework for taking full advantage of computing power of virtual engine, propose a novel voxelization algorithm of volumetric radioactive source. According to the framework, we develop the GPPK(GPU-based point kernel gamma dose rate computing) code using GPU programming, to realize the fast dose rate computing in virtual world. The test results show that the GPPK code is play and plug for different scenarios of virtual simulation, has a better performance than CPU-based gamma dose rate computing code, especially on the voxelization of three-dimensional (3D) model. The accuracy of dose rates from the proposed method is in the acceptable range.

• Journal of Radiation Protection and Research
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• 제6권1호
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2335-2347
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• 2023

As medical facilities are usually built at urban areas, special concrete aggregates and evaluation methods are needed to optimize the design of concrete walls by balancing density, thickness, material composition, cost, and other factors. Carbon treatment rooms require a high radiation shielding requirement, as the neutron yield from carbon therapy is much higher than the neutron yield of protons. In this case study, the maximum carbon energy is 430 MeV/u and the maximum current is 0.27 nA from a hybrid particle therapy system. Hospital or facility construction should consider this requirement to design a special heavy concrete. In this work, magnetite is adopted as the major aggregate. Density is determined mainly by the major aggregate content of magnetite, and a heavy concrete test block was constructed for structural tests. The compressive strength is 35.7 MPa. The density ranges from 3.65 g/cm³ to 4.14 g/cm³, and the iron mass content ranges from 53.78% to 60.38% from the 12 cored sample measurements. It was found that there is a linear relationship between density and iron content, and mixing impurities should be the major reason leading to the nonuniform element and density distribution. The effect of this nonuniformity on radiation shielding properties for a carbon treatment room is investigated by three groups of Monte Carlo simulations. Higher density dominates to reduce shielding thickness. However, a higher content of high-Z elements will weaken the shielding strength, especially at a lower dose rate threshold and vice versa. The weakened side effect of a high iron content on the shielding property is obvious at 2.5 µSv=h. Therefore, we should not blindly pursue high Z content in engineering. If the thickness is constrained to 2 m, then the density can be reduced to 3.3 g/cm³, which will save cost by reducing the magnetite composition with 50.44% iron content. If a higher density of 3.9 g/cm³ with 57.65% iron content is selected for construction, then the thickness of the wall can be reduced to 174.2 cm, which will save space for equipment installation.

• Journal of Radiation Protection and Research
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• 제39권4호
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• pp.176-181
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• 2014

The object of this paper is to evaluate the fission product inventories and radiological doses in a non-LOCA event, based on the U.S. NRC's regulatory methodologies recommended by the TID-14844 and the RG 1.195. For choosing a non-LOCA event, one fuel assembly was assumed to be melted by a channel blockage accident. The Hanul nuclear power reactor unit 6 and the CE $16{\times}16$ fuel assembly were selected as the computational models. The burnup cross section library for depletion calculations was produced using the TRITON module in the SCALE6.1 computer code system. Based on the recently licensed values for fuel enrichment and burnup, the source term calculation was performed using the ORIGEN-ARP module. The fission product inventories released into the environment were obtained with the assumptions of the TID-14844 and the RG 1.195. With two kinds of source terms, the radiological doses of public in normal environment reflecting realistic circumstances were evaluated by applying the average condition of meteorology, inhalation rate, and shielding factor. The statistical analysis was first carried out using consecutive three year-meteorological data measured at the Hanul site. The annual-averaged atmospheric dispersion factors were evaluated at the shortest representative distance of 1,000 m, where the residents are actually able to live from the reactor core, according to the methodology recommended by the RG 1.111. The Korean characteristic-inhalation rate and shielding factor of a building were considered for a series of dose calculations.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• 한국추진공학회지
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• 제25권2호
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• pp.42-54
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• 2021

Aft-deck은 가스터빈 배기 플룸을 차폐하거나, 제트를 확산하여 외부 공기와 혼합율을 증가시켜 배기가스의 온도를 저감하기 위한 용도로 최신 무인항공기에 적용되고 있다. 본 연구에서는 Aft-deck의 설계 변수에 따라 노즐의 성능이 어떻게 영향을 받는지 알아보고자 하였다. Aft-deck의 설계 변수로는 길이, 확산각, 상부 덱 형상으로 선정하였으며, 길이 변수에 따른 추력과 차폐율 간의 상관관계를 제시하였다. 그리고 확산 각도의 범위에 따라 노즐 추력과 제트 확산에 대한 상관관계를 제시하였다. 또한, 상부 덱 제거를 통해 추력 향상 효과를 확인하였으며, 횡 방향 속도 벡터의 특성이 외부 유동과의 혼합 성능을 결정하는 것을 알게 되었다.