• Nuclear Engineering and Technology
• /
• v.26 no.1
• /
• pp.1-8
• /
• 1994

The flame speed correlation considering thermal-hydraulic phenomena under severe accidents is proposed and correction coefficients are defined. This correlation modifies the pressure dependency in Iijima-Takeno correlation and adds the steam suppression effects to it in the anticipated hydrogen and steam concentration ranges under severe accidents. The existing models of flame speed due to hydrogen combustion under severe accidents are based on the experiments which were performed merely at room temperature and atmospheric pressure. They have difficulty in predicting a accurate flame speed in a case of high temperature and pressure during severe accidents. Thus the flame structure is assumed as a prerequisite to the reliable determination of flame speed and theoretical model is developed. To examine the validity, flame speeds in various conditions calculated by this model are compared with those obtained by the calculation of the existing correlations of the codes such as improved HECTR and MAAP. Also the steam suppression ratio is quantified and the steam suppression coefficient is defined as a composition of mixture. Initial temperature and pressure dependencies are investigated and correction coefficents are determined. More experimental studies can be recommended to improve this correlation to its further works.

• Nuclear Engineering and Technology
• /
• v.27 no.3
• /
• pp.292-300
• /
• 1995

This paper present a new dynamic HRA (Human Reliability Analysis) method and its application for Quantifying the human error probabilities in implementing an accident management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequently used methods in PSAs, are discussed. The action associated with the implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic distributions obtained, human error probabilities are calculated with respect to the various means and variances of the timings. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.

• Journal of Korea Society of Industrial Information Systems
• /
• v.27 no.6
• /
• pp.95-103
• /
• 2022

Quick and accurate understanding of the situation in a severe accident is essential for conducting the appropriate accident management and response using the accident diagnosis information. This study employed deep learning technology to diagnose severe accidents through the major safety parameters transferred from a nuclear power plant (NPP) to AtomCARE. After selecting the major accident scenarios to consider, a learning database was established for particular scenarios affiliated with major scenarios by performing a large number of severe accident analyses using MAAP5 code. The severe accident diagnosis technology, which classifies detailed accident scenarios using the major safety parameters from NPPs, was developed by training it with the established database . Verification and validation were conducted by blind test and principal component analysis. The technology developed in this study is expected to be extended and applied to all severe accident scenarios and be utilized as a base technology for quick and accurate severe accident diagnosis.

• Nuclear Engineering and Technology
• /
• v.44 no.3
• /
• pp.311-322
• /
• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• Proceedings of the Korean Institute of Navigation and Port Research Conference
• /
• 2016.05a
• /
• pp.10-12
• /
• 2016

이 연구는 케냐 해양대학 항해학과 교과과정의 개발을 위해 다음과 같은 다양한 연구들이 수행되었다. 먼저 2장에서 케냐의 교육 및 환경 여건을 살펴보기 위해 케냐의 전반적인 상황과 JKUAT 및 동 대학 기관공학과의 교과과정을 살펴보았다. 이어 3장에서는 항해학과 교과과정의 표준으로 삼아야 할 IMO STCW 협약 해기교육요건과 IMO 모델 코스를 살펴보았다. 이어 4장에서는 항해학과 교과과정의 개발에 참고하기 위해 모범적인 해기교육을 실시하고 있는 미국의 USMMA와 CMA, 필리핀의 PMMA와 MAAP, 한국의 한국해양대학교, 중국의 대련해사대학교, 그리고 베트남의 VIMARU와 HCMUT의 교과과정을 각각 살펴보았다. 마지막으로 5장에서는 이상의 검토 결과를 토대로 케냐 해양대학의 항해학과 교과과정을 제시하였다.

• Journal of Radiation Protection and Research
• /
• v.26 no.3
• /
• pp.197-201
• /
• 2001

A risk analysis consists of a triplet, , where Si is the scenario identification; Pi is the probability of each scenario; and Xi is the consequences of each scenario. A new computing framework, OMAM (ORIGEN-MAAP4-MMCS), has been developed and applied for assessing the risk of a reference plant as well as radiation source terms using the concept of risk triplet. The result of this study using the OMAM framework presented in this paper, can contribute to producing domestic nuclear power plant's risk data base as well as to establishing severe accident management plans.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.795-800
• /
• 1998

월성 발전소에서의 중대사고 진행을 예측하고 분석하기 위하여 CANDU 발전소의 중대사고 해석코드인 ISAAC을 이용하여 능동 안전 장치들이 동작하지 않는 발전소 정전사고에 대한 분석을 수행하였다. ISAAC은 사고 관리용 전산코드로 개발된 MAAP4/PWR를 근간으로 월성 발전소의 고유 특성에 관한 모델들을 추가하여 개발되었다 월성 발전소의 경우 칼랜드리아 안의 감속재와 칼랜드리아 볼트 안의 냉각수는 피동 열침원으로, 능동 안전 장치들이 동작하지 않는 중대사고시 노심으로부터의 붕괴열을 제거하여 노심 손상을 지연시킬 수 있다. 발전소 정전사고에 대한 ISAAC 계산 결과 노심 손상후 칼랜드리아 파손까지 약 40 시간이 예상되며, 이 동안에 운전자는 사고를 완화시킬 수 있는 방안을 모색할 수 있다. 따라서, 월성발전소의 피동안전장치는 사고 관리 전략 수립에 중요한 기능을 담당한다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05d
• /
• pp.95-100
• /
• 1996

현재 방사선 비상훈련에 사용하는 TID-14844 의 방사선원항은 너무 보수적이고, 각 사고경로별로 방사선원항의 특성을 나타낼 수 없으므로, 원전의 비상사고 발생시 주민의 피폭선량을 최소화하기 위한 발전소 요원의 신속, 정확한 대처능력을 배양하기 위하여 현실적인 방사선원항 평가자료의 필요성이 대두되어 왔다. 본 연구에서는 보수성을 배제한 최적 분석기법을 이용하여 선정된 사고경로에 대해 MAAP 전산코드로 사고진행 및 방사선원항을 분석하였고, 격납건물내 방사선계측기의 예측치를 평가할 수 있는 방법론을 개발하였으며, 이를 통해 사고경로별 안전변수 및 방사선 계측기 등에서의 사고 진행에 따른 예측치 등을 계산함으로써 효과적인 비상대책 수립을 위한 실질적인 방사선원항 데이타 베이스를 구축하였다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.771-776
• /
• 1998

울진 원전 3,4 호기를 대상으로 MAAP4.0.2 코드를 이용하여 발전소 정전사고를 모의/분석했다. 본 분석에서는 사고진행에 따른 일.이차계통의 상태변화를 원자로용기 파손때까지 상세 파악하였다. 사고관리 관점에서, 발전소 정전사고는 이차측의 대기방출밸브를 통한 강제감압에 의해 사고진행을 완화할 수 있으며 이러한 운전원 조치에 의한 완화효과를 검토하였다 그 결과 감압시작 2시간후에 일차측은 약 24$^{\circ}C$ 의 과냉각도를 보이며 안정되었고. 사고시작 1시간 후부터 3시간 동안의 강제감압이 성공한 경우, 노심노출시간 기준으로 약 2시간의 지면효과가 있었다.

• Nuclear Engineering and Technology
• /
• v.47 no.6
• /
• pp.719-728
• /
• 2015

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.