• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

• Journal of the Korean Institute of Gas
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• v.26 no.3
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• pp.38-44
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• 2022

As the supply of hydrogen charging stations for hydrogen supply accelerates due to the hydrogen economy revitalization policy, the risk of accidents is also increasing. Since most hydrogen explosion accidents lead to major accidents, it is very important to secure safety when using hydrogen energy. In order to utilize hydrogen energy, it is essential to secure the safety of hydrogen storage containers used for production, storage, and transportation of liquid hydrogen. In this paper, in order to evaluate the structural safety of a hydrogen-filled pressure vessel, the behavioral characteristics of gas pressure were analyzed by finite element analysis. SA-372 Grade J / Class 70 was used for the material of the pressure vessel, and a hexahedral mesh was applied in the analysis model considering only the 1/4 shape because the pressure vessel is axisymmetric. A finite element analysis was performed at the maximum pressure using a hydrogen gas pressure vessel, and the von Mises stress, deformation, and strain energy density of the vessel were observed.

• The monthly packaging world
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• s.150
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• pp.140-143
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• 2005

식품의약품안전청은 기구 및 용기 포장의 재질로 사용되는 폴리부틸렌숙시네이트(PBSA) 재질을 신설하는 내용을 입안.예고 했다. PBSA 수지는 토양매립시 자연에서 분리되는 친환경 소재로 최근 일회용 식품용기 등에 사용이 가능하다. 본고에서는 개정(안)과 신.구조문을 비교해 살펴본다.

• LP가스
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• v.22 no.4
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• pp.40-41
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• 2010

지난 7.22~23 제주 풍림리조트에서 지식경제부, 한국가스안전공사, 한국LP가스공업협회, 한국LP가스판매협회, 수입사 관계자 등과 LP가스소형용기시범 사업자 16명 등 총 40여명이 참석한 가운데 'LP가스소형용기직판 시범사업 Work Shop'이 개최됐다. 관련 내용을 요약 정리하여 게재한다.

• Journal of the KSME
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• v.55 no.10
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• pp.31-35
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• 2015

이 글에서는 원자력발전소의 안전성 확보를 위한 원자로 압력용기 재료의 신뢰성 평가 및 향상기술에 대해 개괄적으로 기술하였다. 압력용기 소재의 중요성과 가동 중 건전성 평가기술에 대해 다뤘으며 특히, 조사취화특성 평가와 관련한 기술현황을 중점적으로 소개하였다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.14 no.2
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• pp.143-150
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• 2001

원자력발전소의 1차 계통에서 오염된 장비들을 취급이 용이하고 안전하게 운반하기 위한 운반용기는 내부의 방사성 물질에 대한 방사능 평가에 의하여 방사성물질 A형 운반용기로 분류된다. 방사성물질 A형 운반용기는 IAEA Safety Standard Series No. ST-1 및 국내 원자력법 등 관련규정의 기술기준을 만족하여야 하는데, 운반용기는 중량에 따라 0.3∼1.2m의 높이에서 소성이 일어나지 않는 단단한 바닥면으로 가장 심각한 손상을 주는 방향으로 낙하시키는 정상운반조건(normal transport conditions)에 대하여 구조적 건전성을 유지하여야 한다. 여기서는 ABAQUS/Explicit 코드를 이용하여 컨테이너형태의 A형 운반용기에 대하여 최대손상이 야기되는 0.9m 경사낙하조건에 대한 3차원 충격해석을 수행하고 구조적 건전성을 평가하였는데, 운반용기는 경사낙하시 코너피팅(corner fitting)의 분쇄(crush)에 의하여 대부분의 충격을 흡수하였으며 운반용기의 격납경계는 구조적 건전성을 유지하였다.

• Nuclear Engineering and Technology
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• v.19 no.4
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• pp.292-309
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• 1987

A large dry PWR containment response analysis for postulated severe accidents was performed as part of the Zion Risk Rebaselining study for input to the U.S. NRC's "Reactor Risk Reference Document," NUREG-1150. The Methodologies used in the present work were developed as part of the Severe Accident Risk Reduction Program (SARRP) at Sandia National Laboratory specifically for the Surry Plant, but they were extrapolated to Zion. Major steps of the quantification of risk from a nuclear power plant are first outlined. Then, the methodologies of containment response analysis for severe accidents used for Zion are described in detail: major features of the containment event tree (CET) analysis codes and CET quantification procedures are summarized. In addition, plant specific features important to containment response analysis are presented along with the containment loading and performance issues included in the present uncertainty analysis. Finally, a brief summary of the results of deterministic and statistical containment event tree analysis is presented to provide a perspective on the large dry PWR containment response for postulated severe accidents.accidents.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2005.04a
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• pp.229-234
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• 2005

• The Safety technology
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• no.100
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• pp.60-63
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• 2006

• Journal of Radiation Protection and Research
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• v.31 no.1
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• pp.37-46
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• 2006

For the first step to develop a reference disposal container of spent fuel to be used in a deep geological repository, this paper examined safe dimensions of the disposal container on the points of nuclear criticality and radiation safety and mechanical structural safety and provided basic information for dimensioning the container and configuration of the container components, and establishing the favorable and safe disposal conditions. When the safety factor for stress due to the external loads (hydrostatic and swelling pressure) is taken as 2.0, the safe diameter of the filler material to provide enough container strength under the assumed external loads is found to be 112cm with 13cm spacing between inner baskets in PWR container. Also the thickness of the thinner section between the fuel basket and the surface of the cast insert is determined to be 150 mm. Regarding these dimensions of the container, the PWR fuel container is sketched to accommodate 4 square assemblies or 297 CANDU fuel 297 bundles (33 circle tubes x 9 stacks). However the top and bottom parts need to be checked again through the detail radiation shielding analysis with respects to the emplacement position and handling processes of the disposal container.