• Nuclear Engineering and Technology
• /
• v.56 no.11
• /
• pp.4587-4600
• /
• 2024

Based on the lessons from the Fukushima accident, new strategies to mitigate the spread of accidents in nuclear power plants have been established that consider portable equipment. The diverse and flexible coping strategies (FLEX) and multi-barrier accident coping strategy (MACST) based on the utilization of portable equipment are being adopted in case of an extended loss of all AC power and loss of ultimate heat sink. In this context, the establishment of specific strategies for utilizing portable equipment has emerged as an important issue. This study proposes a multi-unit probabilistic safety assessment (MUPSA)-based evaluation framework consisting of several steps to quantitatively assess the reduction in risk from the use of portable equipment at a multi-unit site, which can provide a basis for establishing optimal portable equipment strategies. According to the framework, after first defining the scope of the evaluation, risk-significant types of portable equipment are selected and the scenarios are analyzed. Risk reduction is then evaluated using modified MUPSA models according to the deployment of the portable equipment. A case study was conducted for an internal loss of off-site power (LOOP) and seismic LOOP focusing on how to evaluate the effectiveness of portable equipment through the developed framework.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.180.1-180
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.50 no.8
• /
• pp.1234-1245
• /
• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.464-470
• /
• 1997

Severe accident analysis has been performed for the Wolsung nuclear power plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks(LOAH) which is caused by loss of off-site power, diesel generators, and DC power. ISAAC (Integrated Severe Accident Analysis Code)computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanism of calandria vessel and containment. In addition, some insights for accident management program(AMP) are also given.

• KSCE Journal of Civil and Environmental Engineering Research
• /
• v.44 no.3
• /
• pp.277-282
• /
• 2024

Recently, the intensity and frequency of typhoons have been increasing due to climate change, and typhoons can cause a loss of offsite power (LOOP) at nuclear power plants (NPPs). Therefore, it is necessary to prepare for typhoon-induced high winds through the probabilistic safety assessment (PSA) of offsite power systems. However, research on PSA for offsite power system in NPPs under typhoon-induced high winds is still lacking. In this study, PSA was performed for offsite power systems subjected to typhoon-induced high winds at the Kori NPP site, which has experienced frequent damages to its offsite power system among NPP sites in Korea. In order to perform PSA for typhoon-induced high winds in offsite power systems, the typhoon hazard at Kori NPP site was derived using logic tree and Monte Carlo simulation. Utilizing the fragility of components constituting the power system, performed a fragility analysis of the power system. Lastly, the probability that offsite power system will not be able to supply power to the NPP was derived.

• Proceedings of the KIEE Conference
• /
• 1998.07c
• /
• pp.1170-1172
• /
• 1998

The electrical power of nuclear power plant consists of Safety related power systems and Non-Safety related power systems. The safety related power systems are designed to have sufficient capacity to safely shut down the unit and to mitigate the effects of an accident assuming loss of off-site power. This paper presents the operation scheme of the safety related power system for several plant conditions in Korean Next Generation Reactor and reviews the diversity of power supply to the safety related bus.

• Journal of Power System Engineering
• /
• v.17 no.4
• /
• pp.17-22
• /
• 2013

Diesel generator of nuclear power plant has a role for supply of emergency electric power to protect reactor core system in event of loss of off-site power supply. Therefore diesel generator should be tested periodically to verify the function that can supply specified frequency and voltage at design power level within limited time. For this purpose, appropriate maintenances in case that abnormal conditions were found are required in allowed time. In this paper, results of development of engine condition diagnosis technology and study on power plant of its technology for diesel generator are described.

• Proceedings of the KIEE Conference
• /
• 2000.07a
• /
• pp.175-181
• /
• 2000

월성원자력4호기가 '99.06.09일 11:53분 경 전출력 소외전원상실시험시 주발전기 병입차단기 및 기동용변압기 차단기의 트립과 동시에 예상 밖의 13.8Kv 원자로 냉각재펌프모터#3(9.000HP)이 순시과전류 보호계전기(50Y) 동작으로 트립되어 이로 인한 냉각재 저유량으로 원자로 제1정지계통이 동작되고 원자로가 비상정지 되어 동 시험이 실패되었음. 이 비정상적인 고장은 예비디젤발전기의 수동 기동 병입과 터빈 수동 정지 및 주발전기 트립후 적절한 조치로 소내전원은 정상적으로 복구되었음. 이에 대해 냉각재펌프모터#3의 순시과전류 동작 원인을 유도전동기의 전원상실 후 발생되는 잔류전압(Residual Voltage)과 공급 모선전압(Bus Voltage) 측면에서 분석하며, 모터의 회전속도, 위상각, 잔류전압크기 변화 및 신속개방 절체시 냉각재펌프모터의 돌입 기동전류를 계산하고, PSS/E 프로그램을 사용한 간략한 모의 사례로 검증하였으며 이에 대한 재발방지를 위한 대책을 제시함.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.458-463
• /
• 1997

A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA, at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation.

• Journal of the Korean Society of Systems Engineering
• /
• v.12 no.2
• /
• pp.91-99
• /
• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.