• Nuclear Engineering and Technology
• /
• 제52권2호
• /
• pp.397-416
• /
• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part II, based on the verified finite element (FE) models of aircrafts Airbus A320 and A380, as well as the NPP containment and auxiliary buildings in Part I of this paper, the whole collision process is reproduced numerically by adopting the coupled missile-target interaction approach with the finite element code LS-DYNA. The impact induced damage of NPP plant under four impact locations of containment (cylinder, air intake, conical roof and PCS water tank) and two impact locations of auxiliary buildings (exterior wall and roof of spent fuel pool room) are evaluated. Furthermore, by considering the inner structures in the containment and raft foundation of NPP, the structural vibration analyses are conducted under two impact locations (middle height of cylinder, main control room in the auxiliary buildings). It indicates that, within the discussed scenarios, NPP structures can withstand the impact of both two aircrafts, while the functionality of internal equipment on higher floors will be affected to some extent under impact induced vibrations, and A380 aircraft will cause more serious structural damage and vibrations than A320 aircraft. The present work can provide helpful references to assess the safety of the structures and inner equipment of NPP plant under commercial aircraft impact.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.581-582
• /
• 2005

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2004년도 추계학술발표회 발표논문집
• /
• pp.1327-1328
• /
• 2004

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.727-728
• /
• 2005

• 콘크리트학회논문집
• /
• 제22권6호
• /
• pp.817-824
• /
• 2010

약 콘크리트 격납구조물은 구조적 안전성이 뛰어나고 경제적이므로 원자력발전소 격납건물, LNG 저장탱크 등에 널리 사용되고 있다. 격납구조물 중 지붕 돔의 형태는 구조적 안전성, 물량 및 시공 난이도에 큰 영향을 미치므로 최적의 두께와 곡률을 도출하고자 하는 노력이 필요하다. 한편 일반적으로 PSC 구조로 설계되는 링빔은 이러한 돔을 지지하여 벽체의 변형을 최소화시키는 역할을 하며, 단면 크기와 더불어 프리스트레스 수준을 적절히 결정하는 것이 설계의 핵심이 된다. 이 연구에서는 축대칭 회전쉘의 막이론을 적용하여 본설계 시의 유한요소해석에 앞서 돔과 링빔의 초기 형상이나 프리스트레스 수준을 효율적으로 결정할 수 있는 기법을 제안하였다. 이러한 기법을 국내에서 시공된 격납구조물의 돔과 링빔에 적용하여 분석하고 단면 형상이나 프리스트레싱 설계에 대한 개선 방안을 고찰하였다.

• Nuclear Engineering and Technology
• /
• 제47권1호
• /
• pp.11-25
• /
• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• 항공우주시스템공학회지
• /
• 제14권3호
• /
• pp.60-68
• /
• 2020

항공기 엔진 내의 파손된 블레이드가 케이스를 관통하여 외부로 나가게 되면, 동체에 부딪혀 승객의 안전을 위협하게 된다. 따라서 엔진 케이스 설계 시 항공청의 안전성 평가 인증을 받아야한다. 본 논문에서는 국내의 독자적인 엔진 개발을 위해 필요한 엔진 케이스의 컨테인먼트 인증 요구조건과 개발 기술에 대하여 조사하였다. 미국과 유럽의 항공청에서 제시하는 컨테인먼트 안전성 요구 조건을 정리하고 이 인증에 부합하는 케이스의 파편 봉쇄를 검증하는 실험적/해석적 방법에 대하여 확인하였다. 컨테인먼트 케이스에 대한 최근의 연구를 분석하여 케이스 개발 시 검증 방법을 제시하고자 한다.

• 한국가스학회지
• /
• 제12권1호
• /
• pp.36-41
• /
• 2008

본 논문에서는 완전밀폐식 LNG 저장탱크의 외부탱크 측벽면과 지붕 구조물에 대한 강도안전성을 유한요소법으로 해석하였다. 예응력 콘크리트 구조물로 건설된 외부탱크는 내부탱크의 붕괴로 인해 발생하는 LNG 유체정압과 유체동압, 그리고 태풍을 포함한 외부의 풍압하중을 받는다. FEM 해석결과에 의하면, 외부탱크의 측벽면과 지붕 구조물이 서로 연결되는 링빔 구조물은 저장탱크에 작용하는 대부분의 내 외부 하중을 담당하고 있다. 이러한 해석결과는 외부탱크의 설계 포인트를 링빔에 두고, 그 다음은 지붕구조물의 중심부에 대한 설계 안전성을 검토하는 것이다. 완전밀폐식 LNG 저장탱크 해석에서 사용한 해석모델은 LNG 누설에 의한 내부압력 및 태풍과 같은 외부압력이 결합된 복합하중에서도 안전한 강도안전성을 유지하고 있음을 알 수 있다.

• Nuclear Engineering and Technology
• /
• 제33권2호
• /
• pp.241-253
• /
• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Nuclear Engineering and Technology
• /
• 제29권3호
• /
• pp.229-239
• /
• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.