• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2918-2927
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• 2020

This paper proposes a simplified elastic-plastic analysis procedure using the penalty factors presented in the Code Case N-779 for strain-based fatigue assessment of nuclear safety class 1 components under severe seismic loads such as safety shutdown earthquake and beyond design-basis earthquake. First, a simplified elastic-plastic analysis procedure for strain-based fatigue assessment of nuclear safety class 1 components under the severe seismic loads was proposed based on the analysis result for the simplified elastic-plastic analysis procedure in the Code Case N-779 and the stress categories corresponding to normal operation and seismic loads. Second, total strain amplitude was calculated directly by performing finite element cyclic elastic-plastic seismic analysis for a hot leg nozzle in pressurizer surge line subject to combined loading including deadweight, pressure, seismic inertia load, and seismic anchor motion, as well as was derived indirectly by applying the proposed analysis procedure to the finite element elastic stress analysis result for each load. Third, strain-based fatigue assessment was implemented by applying the strain-based fatigue acceptance criteria in the ASME B&PV Code, Sec. III, Subsec. NB, Article NB-3200 and by using the total strain amplitude values calculated. Last, the total strain amplitude and the fatigue assessment result corresponding to the simplified elastic-plastic analysis were compared with those using the finite element elastic-plastic seismic analysis results. As a result of the comparison, it was identified that the proposed analysis procedure can derive reasonable and conservative results.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.313-326
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• 2017

The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

• Journal of the Korea Safety Management & Science
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• v.4 no.1
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• pp.49-56
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• 2002

In this case study, results of the explosion accident at MEK-PO factory were analysed by using the consequence analysis of quantitative hazard assessment and the explosion energy, the burst pressure of vessel, and overpressures at the explosion center and at 300m distance from the explosion center were estimated, respectively. As a result, we found that a cause of accident was the runaway reaction of product(MEK-PO) because of the molecular expansion in vessel and that the possibility of the runaway reaction was classified the mechanical failure(the obstacle of refrigerator or the shutdown valve), design error, and operating error by lack of thermochemical knowledge. Also, the evasive action to prevent accident was suggested.

• International Journal of Reliability and Applications
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• v.3 no.1
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• pp.25-35
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• 2002

This paper presents a dynamic reliability assessment methodology for use in the safety assessment of a complex system such as a nuclear power plant. The method is applied to a dynamic analysis of the potential accident sequences that may occur during mid-loop operation in a nuclear power plant. The idea behind this approach consists of both the use of the concept of the performance achievement/requirement correlation and of a dynamic event tree generation method. The assessment of the system reliability depends on the determination of both the required performance distribution and the achieved performance distribution. The quantified correlation between requirement and achievement represents a comparison between two competing variables. It is demonstrated that this method is easily applicable and flexible in that it can be applied to any kind of dynamic reliability problem.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.331-343
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• 2002

We performed an importance analysis of In-Service Testing (157) components for Ulchin Unit 3 using the integrated evaluation method for categorizing component safety significance developed in this study. The developed method is basically aimed at having a PSA expert perform an importance analysis using PSA and its related information. The importance analysis using the developed method is initiated by ranking the component importance using quantitative PSA information. The importance analysis of the IST components not modeled in the PSA is performed through the engineering judgment, based on the expertise of PSA, and the quantitative and qualitative information for the 157 components. The PSA scope for importance analysis includes not only Level 1 and 2 internal PSA but also Level 1 external and shutdown/low power operation PSA. The importance analysis results of valves show that 167 (26.55%) of the 629 IST valves are HSSCs and 462 (73.45%) are LSSCs. Those of pumps also show that 28 (70%)of the 40157 pumps are HSSCs and 12 (30%) are LSSCs.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.191-196
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• 1995

The intelligent human-machine interface (HMI) has been developed to enhance the safety and availability of a nuclear power plant by improving operational reliability The key elements of the HMI are the large display panels which present synopsis of the plant status and the compact, digital work stations for the primary operator control and monitoring functions. The work station consists of four consoles such as a dynamic alarm console (DAC), a system information console (SIC), a computerized operating-procedure console (COC), and a safety related information console (SRIC). The DAC provides clean alarm pictures, in which information overlapping is excluded and alarm impacts are discriminated, for quick situation awareness. The SIC covers a normal operation by offering all necessary plant information and control functions. In addition, it is closely linked with the DAC and the COC to automatically display related system information under the request of these consoles. The COC aids the operator with proper emergency operation guidelines so as to shutdown the plant safely, and it also reduces his physical/mental burden by automating the operating procedures. The SRIC continuously displays safety related information to allow the operator to assess the plant status focusing on plant safety. The proposed HMI has been validated and demonstrated with on-line data obtained from the full-scope simulator for Yonggwang Units 1,2.

• Journal of the Korean Society of Safety
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• v.32 no.4
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• pp.7-15
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• 2017

International standards such as IEC-61508 and IEC-61511 require Safety Integrity Levels (SILs) for Safety Instrumented Functions (SIFs) in process industries. SIL verification is one of the methods for process safety description. Results of the SIL verification in some cases indicated that several Safety Instrumented Functions (SIFs) do not satisfy the required SIL. This results in some problems in terms of cost and risks to the industries. This study has been performed to improve the reliability of a safety instrumented function (SIF) installed in hydrodesulfurization reactor heater using Partial Stroke Testing (PST). Emergency shutdown system was chosen as an SIF in this study. SIL verification has been performed for cases chosen through the layer of protection analysis method. The probability of failure on demands (PFDs) for SIFs in fault tree analysis was $4.82{\times}10^{-3}$. As a result, the SIFs were unsuitable for the needed RRF, although they were capable of satisfying their target SIL 2. So, different PST intervals from 1 to 4 years were applied to the SIFs. It was found that the PFD of SIFs was $2.13{\times}10^{-3}$ and the RRF was 469 at the PST interval of one year, and this satisfies the RRF requirements in this case. It was also found that shorter interval of PST caused higher reliability of the SIF.

• Journal of the Korean Society of Safety
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• v.39 no.2
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• pp.75-86
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• 2024

Hydrogen has been selected as one of the key technologies for reducing CO₂ emissions to achieve carbon neutrality by 2050. However, hydrogen safety issues should be fully guaranteed before the commercial and widespread utilization of hydrogen. Here, a bow-tie risk assessment is conducted for the hydrogen fuel supply system in a gas turbine power plant, which can be a mass consumption application of hydrogen. The bow-tie program is utilized for a qualitative risk assessment, allowing the analysis of the causes and consequences according to the stages of accidents. This study proposed an advanced bow-tie method, which includes the barrier criticality matrix and visualized maps of quantitative risk reduction. It is based on evaluating the importance of numerous barriers for the extent of their impact. In addition, it emphasizes the prioritization and concentrated management of high-importance barriers. The radar chart of a bow tie allows the visual comparison of risk levels before/after the application of barriers (safety measures). The risk reduction methods are semi-quantitatively analyzed utilizing the criticality matrix and radar chart, and risk factors from multiple aspects are derived. For establishing a secure hydrogen fuel storage system, the improvements suggested by the bow-tie risk assessment results, such as 'Ergonomic equipment design to prevent human error' and 'Emergency shutdown system,' will enhance the safety level. It attempts to contribute to the development and enhancement of an efficient safety management system by suggesting a method of calculating the importance of barriers based on the bow-tie risk assessment.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.21 no.1
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• pp.760-767
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• 2020

Sustainable growth of hydroelectric power plants is expected in consideration of climate change and energy security. However, hydroelectric power plants always have a risk of water hammer damage, and safety assurance is very important. The water hammer phenomenon commonly occurs during operations such as rapid opening and closing of the valves and pump/turbine shutdown in pipe systems, which is more common in cases of emergency shutdown. In this study, a computational numerical model was developed using the MOC-FDM scheme to reflect the mechanism of water hammer occurrence. The proposed model was implemented in boundary conditions such as reservoir, pipeline, valve, and pump/turbine conditions and then applied to simulate hypothetical case studies. The analysis results of the model were verified using the analysis results at the main points of the pipe systems. The model produced reasonably good performance and was validated by comparison with the results of the SIMSEN package model. The model could be used as an efficient tool for the safety assessment of hydroelectric power plants based on accurate prediction of transient behavior in the operation of hydropower facilities.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.37 no.3
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• pp.205-216
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• 2024

Recently, seismic Probabilistic Safety Assessment (PSA) methods have been developed for process plants, such as gas plants, oil refineries, and chemical plants. The framework originated from the PSA of nuclear power plants, which aims to assess the risk of reactor core damage. The original PSA method was modified to adopt the characteristics of a process plant whose purpose is continuous operation without shutdown. Therefore, a fault tree, whose top event is shut down, was constructed and transformed into a Bayesian Network (BN), a probabilistic graph model, for efficient risk-informed decision-making. In this research, the fault tree-based BN from the previous research is further developed to consider the multi-hazard of earthquake-induced fire and explosion (EQ-induced F&E). For this purpose, an event tree describing the occurrence of fire and explosion from a release is first constructed and transformed into a BN. And then, this BN is connected to the previous BN model developed for seismic PSA. A virtual plot plan of a gas plant is introduced as a basis for the construction of the specific EQ-induced F&E BN to test the proposed BN framework. The paper demonstrates the method through two examples of risk-informed decision-making. In particular, the second example verifies how the proposed method can establish a repair and retrofit strategy when a shutdown occurs in a process plant.