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

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2092-2098
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• 2024

Probabilistic safety assessment (PSA) plays a critical role in ensuring the safe operation of nuclear power plants. In PSA, event trees are developed to identify accident sequences that could lead to core damage. These event trees are then transformed into a core-damage fault tree, wherein the accident sequences are represented by usual and complemented logic gates representing failed and successful operations of safety systems, respectively. The core damage frequency (CDF) is estimated by calculating the minimal cut sets (MCSs) of the core-damage fault tree. Delete-term approximation (DTA) is commonly employed to approximately solve MCSs representing accident sequence logics from noncoherent core-damage fault trees. However, DTA can lead to an overestimation of CDF, particularly when fault trees contain many nonrare events. To address this issue, the present study introduces a new zero-suppressed ternary decision diagram (ZTDD) algorithm that averts the CDF overestimation caused by DTA. This ZTDD algorithm can optionally calculate MCSs with DTA or prime implicants (PIs) without any approximation from the core-damage fault tree. By calculating PIs, accurate CDF can be calculated. The present study provides a comprehensive explanation of the ZTDD structure, formula of the ZTDD algorithm, ZTDD minimization, probability calculation from ZTDD, strength of the ZTDD algorithm, and ZTDD application results. Results reveal that the ZTDD algorithm is a powerful tool that can quickly and accurately calculate CDF and drastically improve the safety of nuclear power plants.

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

• Journal of the Korean Society of Safety
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• v.38 no.3
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• pp.110-120
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• 2023

Following the Fukushima accident, portable equipment employed as accident mitigating systems have been installed and operated to reduce core damage and large early release frequencies. In addition, the establishment of an accident management strategy has gained importance. This study investigated the current status of portable equipment including the international portable equipment FLEX (diverse and flexible coping strategies), and domestic portable equipment multi-barrier accident coping strategy (MACST). Research on optimal utilization of MACST remains insufficient. As a preliminary study for establishing an optimal strategy, sensitivity studies were conducted to facilitate the priority of use on portable equipment, number of portable equipment, and dependency of operator actions based on a multi-unit probabilistic safety assessment model. The results revealed the conditions that reduced the multi-unit and site conditional core damage probabilities, indicating the optimal strategy of MACST. The results of this study can be used as a reference for establishing an optimal strategy that utilizes domestic safety equipment in the future.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 1998.11a
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• pp.219-224
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• 1998

확률론적 안전성 평가(Probabilistic Safety Assessment PSA)나 정량적인 위험도 평가(Quantitative Risk Assessment: QRA)에서 인간신뢰도분석(human reliability analysis)은 인간행위를 기기처럼 생각하여 전체 시스템의 안전성에 중요한 초기사건(initiating event) 이전이나 초기사건 이후 또는 초기사건을 유발하는 인간행위를 파악하고 정량화하여, 확률론적 평가의 논리구조인 사건 및 고장수목(event tree 및 fault tree)이나 사고경위 단절집합 (accident sequence outsets)에 포함시키는 것이다. (중략)

• Journal of the Korean Society of Safety
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• v.15 no.2
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• pp.126-135
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• 2000

Probabilistic Safety Assessment (PSA) is an engineering analysis method to identify possible contributors to the risk from a nuclear power plant and now it has become a standard tool in safety evaluation of nuclear power plants. PSA consists of three phases named as Level 1, 2 and 3. Level 2 PSA, mainly focused in this paper, uses a step-wise approach. At first, plant damage states (PDSs) are defined from the Level 1 PSA results and they are quantified. Containment event tree (CET) is then constructed considering the physico-chemical phenomena in the containment. The quantification of CET can be assisted by a decomposition event tree (DET). Finally, source terms are quantitatively characterized by the containment failure mode. As the main benefit of PSA is to provide insights into plant design, performance and environmental impacts, including the identification of the dominant risk contributors and the comparison of options for reducing risk, this technique is expected to be applied to the industrial safety area.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.426-433
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• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.692-701
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• 2019

Recently, instrumentation and control (I&C) systems in nuclear power plants have undergone digitalization. Owing to the unique characteristics of digital I&C systems, the reliability analysis of digital systems has become an important element of probabilistic safety assessment (PSA). In a reliability analysis of digital systems, fault-tolerant techniques and their effectiveness must be considered. A fault injection experiment was performed on a safety-critical digital I&C system developed for nuclear power plants to evaluate the effectiveness of fault-tolerant techniques implemented in the target system. A software-implemented fault injection in which faults were injected into the memory area was used based on the assumption that all faults in the target system will be reflected in the faults in the memory. To reduce the number of required fault injection experiments, the memory assigned to the target software was analyzed. In addition, to observe the effect of the fault detection coverage of fault-tolerant techniques, a PSA model was developed. The analysis of the experimental result also can be used to identify weak points of fault-tolerant techniques for capability improvement of fault-tolerant techniques

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2047-2052
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• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Journal of the Korean Society of Safety
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• v.36 no.5
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• pp.86-91
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• 2021

Calculating the scrutable core damage frequency (CDF) of nuclear power plants is an important component of the seismic probabilistic safety assessment (SPSA). In this work, a simple approach is developed to calculate CDF from minimal cut sets (MCSs) with non-rare events. When conventional calculation methods based on rare event approximations are employed, the CDF of industry SPSA models is significantly overestimated by non-rare events in the MCSs. Recently, quantification algorithms using binary decision diagrams (BDDs) have been introduced to prevent CDF overestimation in the SPSA. However, BDD structures are generated from a small part of whole MCSs due to limited computational memory, and they cannot be reviewed due to their complicated logic structure. This study suggests a simple approach for scrutinizing the CDF calculation based on whole MCSs in the SPSA system analysis model. The proposed approach compares the new results to outputs from existing algorithms, which helps in avoiding CDF overestimation.