• 한국방사선학회논문지
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• 제12권4호
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• pp.467-474
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• 2018

CBCT는 치료부위의 정확도 향상에 유용하지만, 반복적인 사용으로 피폭선량이 높아지는 단점이 있다. 이에 본 연구에서는 차폐체를 사용한 모의실험과 선량감소 효과를 데이터화하여 CBCT 시행 시 선량 저감화를 위한 기초자료를 제공하고자 한다. 본 연구에서는 MCNPX를 통해 CBCT를 모사하여 광자선을 분석한 후, UF-revised 인체 모의 피폭체를 대상으로 흉복부 촬영 시 장기의 흡수선량을 계산하였다. 이 때, 차폐체(납, 안티몬, 황산바륨, 텅스텐, 비스무스) 유무와 차폐 재질에 따른 장기선량을 평가하였다. 차폐를 하지 않고 CBCT 촬영을 하였을 경우 유방 과 척추에서 선량이 높게 계산되었으며, 식도와 폐에서 선량이 낮게 계산되었다. 차폐체 재질에 따른 선량 은 황산바륨, 안티몬, 비스무스, 납, 텅스텐 순으로 선량이 높게 계산되었다. 차폐체 유무에 따른 선량 감소율을 평가해 보면 흉선(73.6%), 유방(59.9%)에서 가장 차폐율이 높고, 폐(2.1%), 척추(12.6%)에서 가장 낮은 차폐율을 보였다.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.674-680
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• 2022

In the present study, radiation shielding and protection ability of prepared Flyash-lime-Gypsum (FaLG) bricks has been studied in terms of energy exposure build up factors and dose parameters. The energy exposure build up factors of Flyash-lime-Gypsum (FaLG) bricks have been calculated for the energy range of 0.015 MeV-15 MeV and for penetration depth upto 40 mfp directly using a new and simplified Piecewise Linear Spline Interpolation Method (PLSIM). In this new method, the calculations of G.P fitting parameters are not required. The verification and accuracy of this new method has been checked by comparing the results of exposure build up factor for NBS concrete calculated using present method with the results obtained by using G.P fitting method. Further, the relative dose distribution and reduced exposure dose rate for various radioactive isotopes without any shielding material and with Flyash-lime-Gypsum (FaLG) bricks have been calculated in the energy range of 59.59-1332 keV. On the basis of the obtained results, it has been reported that the prepared Flyash-lime-Gypsum (FaLG) bricks possess satisfactory radiation shielding properties and can be used as environmentally safe storage facilities for low level nuclear waste.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2215-2221
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• 2023

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

• 대한방사선치료학회지
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• 제2권1호
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• pp.31-40
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• 1987

In the case of designing a high dose rate remote controlled afterloading treatment room with existing hospital facilities. We must construct the effective protective barriers so as to reduce the primary and scattered radiation up to the maximum permissible dose level. It is difficult to reinforce the barrier thickness of the shielding requirements because of the limited space and the problem of the existing building structure at the surrounding area. Therefore we can reduce the intensity of primary radiation to the required degree at the location of interest with installing the appropriate I shaped Pb barriers between the radiation source and the shielding wall of the concrete. As a result, it was possible to reduce the intensity of the primary radiation below the M.P.D level by using additional Pb barriers instead of increasing thickness of concrete wall.

• Journal of Welding and Joining
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• 제29권5호
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• pp.58-62
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• 2011

Stainless steels are alloy steels with a nominal chromium content of at least 11 percent, with other alloy additions. The stainlessness and corrosion resistance of these alloy steels are attributed to the presence of a passive oxide film on the surface. When exposed to conditions like Resistance Spot Welding (RSW) process that remove the passive oxide film, stainless steels are subject to corrosive attack. And exposure to elevated temperatures causes oxidation (discoloration) of areas around indentation in Spot welding. In this paper, deal with the effect of shielding gas (Ar) preventing the corrosion, oxidation of stainless steel. And find the optimal shielding gas flow rate. In addition, suggest effective purging method for direct/indirect spot welding process.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2015년도 춘계 학술논문 발표대회
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• pp.160-161
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• 2015

Borosilicate glass can be used for improving neutron shielding of concrete. The well known expansion of borosilicate glass caused by expansion of mortar bar was can cause serious damage to the concrete. In this research, borosilicate glass was powdered to reduce the particle size similar to that of cement, and 20% cement replacement set was reduced expansion rate about 30%. But aggregate replacement set was damaged because of Alkali-Silica Reaction expansion.

• International Journal of Contents
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• 제14권4호
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• pp.65-69
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• 2018

The objective of this study was to conduct a radiation shielding analysis for the facility equipped with radiation generator. The analysis was carried out in two aspects. First, from the aspect of the effect caused by primary and leakage radiation. Second, effect of scattered radiation was evaluated by applying a simple calculation method based on a scattering rate concept since effect of scattered radiation is significantly important at maze entrance of the radiation facility. The calculated results obtained using the simple method were compared to the results calculated using Geant4 code and the measured values. The results calculated by the suggested method indicate that slight error exists in a radiation shielding analysis done at the maze entrance comparing to other two results, while the results evaluated at the outside of the maze entrance door are relatively consistent with other values.

• Journal of Advanced Marine Engineering and Technology
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• 제29권4호
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• pp.399-406
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• 2005

Alloy 625 is used widely in industrial applications such as aeronautical aerospace, chemical, petrochemical and marine applications. Because of a good combination of yield strength. tensile strength, creep strength, excellent fabricability, weldability and good resistance to high temperature corrosion on prolonged exposure to aggressive environments. High qualify weldments for this material are readily produced by commonly used processes. But all of processes are not applicable to this material by reason of unavailability of matching, position or suitable welding filler metals and fluxes may limit the choice of welding processes. Recently, the flux cored wire is developed and applied for the better productivity in several welding position including the vortical position. In this study. the weldability and weldment characteristics of Alloy 625 are evaluated in FCAW weld associated with the several shielding gases($80\%Ar+20\%\;CO_2,\;50\%Ar+50\%\;CO_2.\;100\%\;CO_2$) in viewpoint of welding productivity. The results of the experimental study on corrosive characteristics of Alloy 625 are as follows; There is no remarkable difference among shielding gases. however they has a striking difference among corrosive solutions by results of distinguished density and time of corrosive solution. Generally, the shielding gases($80\%Ar+20\%\;CO_2$) was superior to the other gases on high temperature tensile and a low temperature impact. but all of the shield gases were making satisfactory results on corrosion test.

• 대한방사선기술학회지:방사선기술과학
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• 제42권2호
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• pp.99-104
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• 2019

Mammography, conducted every two years, causes cancer due to regular exposure to radiation while reducing rate of death caused by breast cancer. The study evaluates the effect of breast shielding apron made to shield off scattered radiation that occurs to the breast when the opposite side breast is mammogramed. AGD was measured using ACR phantom, composed of 50% mammary glands and 50% fat, and radiation was measured before and after wearing the apron on the breast when the opposite side of the breast is mammogramed. When CC direction mammography was conducted to a breast, the AGD was 1.84 mGy. When CC direction and MLO direction mammography were done to a breast, the average dose detected from the opposite side breast from four directions(top to bottom and medial to lateral) was $140{\mu}Gy$ with maximum dose of $256{\mu}Gy$ at medial side. After putting on the apron, the dose, caused by scattered radiation, was not detected from any of the four directions. Using of breast shielding apron is expected to minimize the radiation exposure by blocking scattered radiation to the breast shielded, when mammography is done to the opposite side breast.