• Proceedings of the Korean Institute of Industrial Safety Conference
• /
• 1997.11a
• /
• pp.179-184
• /
• 1997

지금까지 수행되었던 원자력발전소의 확률론적 안전성 평가 (Probabilistic Safety Assessment; PSA) 결과, 노심손상 빈도의 30% - 70%가 인간행위와 관련이 있는 것으로 밝혀져 PSA에서 인간행위를 적절히 다루는 것은 매우 중요하다. 특히 원자력발전소의 정지운전인 경우에는 자동으로 작동하는 계통이 거의 없어 고장수목(fault tree)과 사건수목(event tree)의 모델링에 많은 운전인 행위가 포함되기 때문에 노심손상 빈도와 관련이 있는 인간행위는 전출력 운전(full power operation)에 대한 PSA 결과의 경우보다 많은 것으로 나타났다. PSA에서 인간신뢰도분석(human reliability analysis)은 PSA의 논리구조인 고장수목과 사건수목에 모델링될 인간행위를 파악하고 정량화하는 것이다. 현재 인간신뢰도분석은 인간행위에 대한 데이타의 부족과 인간행위 자체의 다변성(variability)으로 인해 분석에 어려움이 있고 분석자의 주관성이 개입될 여지가 많은 실정이며, 이에 따라 분석 결과에는 많은 불확실성을 내포하게 된다. (중략)

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4367-4381
• /
• 2023

Nuclear power plants are increasingly being equipped with digital I&C systems. Although some probabilistic safety assessment (PSA) models for the digital I&C of nuclear power plants have been constructed, there is currently no specific internationally agreed guidance for their modeling. This paper presents an initiative by the OECD Nuclear Energy Agency called "Digital I&C PSA - Comparative application of DIGital I&C Modelling Approaches for PSA (DIGMAP)", which aimed to advance the field towards practical and defendable modeling principles. The task, carried out in 2017-2021, used a simplified description of a plant focusing on the digital I&C systems important to safety, for which the participating organizations independently developed their own PSA models. Through comparison of the PSA models, sensitivity analyses as well as observations throughout the whole activity, both qualitative and quantitative lessons were learned. These include insights on failure behavior of digital I&C systems, experience from models with different levels of abstraction, benefits from benchmarking as well as major contributors to the core damage frequency and those with minor effect. The study also highlighted the challenges with modeling of large common cause component groups and the difficulties associated with estimation of key software and common cause failure parameters.

• Proceedings of the Korean Institute of Industrial Safety Conference
• /
• 2003.05a
• /
• pp.114-119
• /
• 2003

원자력발전소를 비롯한 위험 시설물의 확률론적 안전성 평가(Probabilistic Safety Assessment: PSA)는 고장수목(Fault Tree) 및 사건수목(Event Tree) 분석으로 이루어지며, 분석 결과로 그 시설물의 위험도(Risk)는 최소단절집합(Minimal Cutsets)들의 합으로 구성 된다.(중략)

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1225-1233
• /
• 2024

Level 3 probabilistic safety assessment (PSA) is performed to calculate radionuclide concentrations and exposure dose resulting from nuclear power plant accidents. To calculate the external exposure dose from the released radioactive materials, the radionuclide concentrations are multiplied by two factors of dose coefficient and a finite cloud dose correction factor (FCDCF), and the obtained values are summed. This indicates that a standard set of FCDCFs is required for external exposure dose calculations. To calculate a standard set of FCDCFs, the effective distance from the release point to the receptor along the wind direction should be predetermined. The TID-24190 document published in 1968 provides equations to calculate FCDCFs and the resultant standard set of FCDCFs. However, it does not provide any explanation on the effective distance required to calculate the standard set of FCDCFs. In 2021, Sandia National Laboratories (SNLs) proposed a method to predetermine finite effective distances depending on the atmospheric stability classes A to F, which results in six standard sets of FCDCFs. Meanwhile, independently of the SNLs, the authors of this paper discovered that an infinite effective distance assumption is a very reasonable approach to calculate one standard set of FCDCFs, and they implemented it into the multi-unit radiological consequence calculator (MURCC) code, which is a post-processor of the level 3 PSA codes. This paper calculates and compares short- and long-range FCDCFs calculated using the TID-24190, SNLs method, and MURCC method, and explains the strength of the MURCC method over the SNLs method. Although six standard sets of FCDCFs are required by the SNLs method, one standard sets of FCDCFs are sufficient by the MURCC method. Additionally, the use of the MURCC method and its resultant FCDCFs for level 3 PSA was strongly recommended.

• Nuclear Engineering and Technology
• /
• v.50 no.3
• /
• pp.319-326
• /
• 2018

Since the Fukushima Daiichi nuclear disaster, concern and worry about multiunit accidents have been increasing. Korea has a higher urgency to evaluate its site risk because its number of nuclear power plants (NPPs) and population density are higher than those in other countries. Since the 1980s, technical documents have been published on multiunit probabilistic safety assessment (PSA), but the Fukushima accident accelerated research on multiunit PSA. It is therefore necessary to summarize the present situation and draw implications for further research. This article reviews journal and conference papers on multiunit or site risk evaluation published between 2011 and 2016. The contents of the reviewed literature are classified as research status, initiators, and methodologies representing dependencies, and the insights and conclusions are consolidated. As of 2017, the regulatory authority and nuclear power utility have launched a full-scale project to assess multiunit risk in Korea. This article provides comprehensive reference materials on the necessary enabling technology for subsequent studies of multiunit or site risk assessment.

• Journal of the Korean Society of Safety
• /
• v.38 no.3
• /
• pp.102-109
• /
• 2023

Following the accident at Fukushima, the true impact of multi-unit accidents came to light. Accordingly, research related to multi-unit accident effect analysis, risk evaluation, and accident prevention/prevention technology has been conducted. Specific examples are mobile/fixed equipment such as multi-barrier accident coping strategy (MACST) and diverse and flexible coping strategies (FLEX), which have been introduced and installed in multi-units for preventing and mitigating multi-unit accidents. These strategies are useful for enhancing the safety of nuclear power plants (NPPs); however, a more efficient strategy is required in terms of the costs of physical and human resources. To effectively and efficiently mitigate an increase in multi-unit accidents, it is necessary to not only to utilize mobile/fixed equipment but to also use crosstie options with resources that already exist at NPPs. Therefore, we analyzed the current international and domestic status of crosstie systems technology and propose a method to evaluate feasibility alongside risk based on a multi-unit probabilistic safety assessment (PSA). To analyze the international and domestic status of crosstie systems technology, actual cases and related research were studied, and a list of potential crosstie safety resources was derived. Additionally, a case study was performed on crosstie cases of two systems within the assumed six units on-site under a multi-unit accident, and a multi-unit PSA-based risk evaluation method is proposed.

• Nuclear Engineering and Technology
• /
• v.52 no.7
• /
• pp.1462-1470
• /
• 2020

Reliability in safety-critical systems and equipment is of vital importance, so the probabilistic safety assessment (PSA) has been widely used for many years in the nuclear industry to address reliability in a quantitative manner. As many nuclear power plants (NPPs) become digitalized, evaluating the reliability of safety-critical software has become an emerging issue. Due to a lack of available methods, in many conventional PSA models only hardware reliability is addressed with the assumption that software reliability is perfect or very high compared to hardware reliability. This study focused on developing a new method of safety-critical software reliability quantification, derived from hardware-software integrated environment testing. Since the complexity of hardware and software interaction makes the possible number of test cases for exhaustive testing well beyond a practically achievable range, an importance-oriented testing method that assures the most efficient test coverage was developed. Application to the test of an actual NPP reactor protection system demonstrated the applicability of the developed method and provided insight into complex software-based system reliability.

• Journal of the Korean Society of Safety
• /
• v.32 no.3
• /
• pp.160-171
• /
• 2017

US national research laboratories developed the first Vital Area Identification (VAI) method for the physical protection of nuclear power plants that is based on Event Tree Analysis (ETA) and Fault Tree Analysis (FTA) techniques in 1970s. Then, Korea Atomic Energy Research Institute proposed advanced VAI method that takes advantage of fire and flooding Probabilistic Safety Assessment (PSA) results. In this study, in order to minimize the burden and difficulty of VAI, (1) a set of streamlined VAI rules were developed, and (2) this set of rules was applied to PSA fault tree and event tree at the initial stage of VAI process. This new rule-based VAI method is explained, and its efficiency and correctness are demonstrated throughout this paper. This new rule-based VAI method drastically reduces problem size by (1) performing PSA event tree simplification by applying VAI rules to the PSA event tree, (2) calculating preliminary prevention sets with event tree headings, (3) converting the shortest preliminary prevention set into a sabotage fault tree, and (4) performing usual VAI procedure. Since this new rule-based VAI method drastically reduces VAI problem size, it provides very quick and economical VAI procedure. In spite of an extremely reduced sabotage fault tree, this method generates identical vital areas to those by traditional VAI method. It is strongly recommended that this new rule-based VAI method be applied to the physical protection of nuclear power plants and other complex safety-critical systems such as chemical and military systems.

• Journal of the Korean Society of Safety
• /
• v.18 no.2
• /
• pp.104-118
• /
• 2003

It has been criticized that conventional human reliability analysis (HRA) methodologies for probabilistic safety assessment (PSA) have been focused on the quantification of human error probability (HEP) without detailed analysis of human cognitive processes such as situation assessment or decision-making which are crticial to successful response to emergency situations. This paper introduces a new human reliability analysis (HRA) methodology, AGAPE-ET (A guidance And Procedure for Human Error Analysis for Emergency Tasks), focused on the qualitative error analysis of emergency tasks from the viewpoint of the performance of human cognitive function. The AGAPE-ET method is based on the simplified cognitive model and a taxonomy of influencing factors. By each cognitive function, error causes or error-likely situations have been identified considering the characteristics of the performance of each cognitive function and influencing mechanism of PIFs on the cognitive function. Then, overall human error analysis process is designed considering the cognitive demand of the required task. The application to an emergency task shows that the proposed method is useful to identify task vulnerabilities associated with the performance of emergency tasks.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.15-19
• /
• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).