• Journal of the Computational Structural Engineering Institute of Korea
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• v.20 no.4
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• pp.463-468
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• 2007

Some network equipments made in Korea were exposed to severe earthquake in Japan several years ago. More than a hundred slim base transfer network stations had been seized with the severe earthquake at Nigata and it was reported that less than fifteen sets showed blackout by interruption of electricity, not by the structural failure. The purpose of this paper is to check the structural safety of the network equipments by performing table test, and the static and dynamic finite element analysis. For the dynamic test, the station weighing 200 kg was subjected to the Zone 3 earthquake loading of GR-63-CORE on the shaking table to obtain the dynamic responses to compare with the analysis results. It is shown that the FE analysis results are a little bit larger than that of the experimental values. And the sensitivity analysis and optimization for the natural frequency is performed and it is found that the first natural frequency is sensitive to small design change as shown in the results. And the dynamic response of optimized design is less than the original design.

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

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2221-2229
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• 2020

After the Fukushima Daiichi nuclear power plant (NPP) accident, level 3 probabilistic safety assessment (PSA) has emerged as an important task in order to assess the risk level of the multi-unit NPPs in a single nuclear site. Accurate calculation of the radionuclide concentrations and exposure doses to the public is required if a nuclear site has multi-unit NPPs and large number of people live near NPPs. So, there has been a great need to develop a new method or procedure for the fast and accurate offsite consequence calculation for the multi-unit NPP accident analysis. Since the multi-unit level 3 PSA is being currently performed assuming that all the NPPs are located at the same position such as a center of mass (COM) or base NPP position, radionuclide concentrations or exposure doses near NPPs can be drastically distorted depending on the locations, multi-unit NPP alignment, and the wind direction. In order to overcome this disadvantage of the COM method, the idea of a new multiple location (ML) method was proposed and implemented into a new tool MURCC (multi-unit radiological consequence calculator). Furthermore, the MURCC code was further improved for the multi-unit level 3 PSA that has the arbitrary number of multi-unit NPPs. The objectives of this study are to (1) qualitatively and quantitatively compare COM and ML methods, and (2) demonstrate the strength and efficiency of the ML method. The strength of the ML method was demonstrated by the applications to the multi-unit long-term station blackout (LTSBO) accidents at the four-unit Vogtle NPPs. Thus, it is strongly recommended that this ML method be employed for the offsite consequence analysis of the multi-unit NPP accidents.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.97-109
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• 2020

On March 11, 2011, an earthquake followed by a tsunami caused an extended station blackout (SBO) at the Fukushima Dai-ichi NPP Units. The accident was initiated by a total loss of both onsite and offsite electrical power resulting in the loss of the ultimate heat sink for several days, and a consequent core melt in some units where proper mitigation strategies could not be implemented in a timely fashion. To enhance the plant's coping capability, the Diverse and Flexible Strategies (FLEX) were proposed to append the Emergency Operation Procedures (EOPs) by relying on portable equipment as an additional line of defense. To assess the success window of FLEX strategies, all sources of uncertainties need to be considered, using a physics-based model or system code. This necessitates conducting a large number of simulations to reflect all potential variations in initial, boundary, and design conditions as well as thermophysical properties, empirical models, and scenario uncertainties. Alternatively, data-driven models may provide a fast tool to predict the success window of FLEX strategies given the underlying uncertainties. This paper explores the applicability of Artificial Intelligence (AI) to identify the success window of FLEX strategy for extended SBO. The developed model can be trained and validated using data produced by the lumped parameter thermal-hydraulic code, MARS-KS, as best estimate system code loosely coupled with Dakota for uncertainty quantification. A Systems Engineering (SE) approach is used to plan and manage the process of using AI to predict the success window of FLEX strategies under extended SBO conditions.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.27-37
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• 2020

The Fukushima Daiichi accident in 2011 revealed some vulnerabilities of existing Nuclear Power Plants (NPPs) under extended Station Blackout (SBO) accident conditions. One of the key Severe Accident Management (SAM) strategies developed post Fukushima accident is the In-Vessel Retention (IVR) Strategy which aims to retain the structural integrity of the Reactor Pressure Vessel (RPV). RELAP/SCDAPSIM/MOD3.4 is selected to predict the thermal-hydraulic response of APR1400 undergoing an extended SBO. To assess the effectiveness of the IVR strategy, it is essential to quantify the underlying uncertainties. In this work, both the epistemic and aleatory uncertainties are considered to identify the success window of the IVR strategy. A set of in-vessel relevant phenomena were identified based on Phenomena Identification and Ranking Tables (PIRT) developed for severe accidents and propagated through the thermal-hydraulic model using Wilk's sampling method. For this work, a Systems Engineering (SE) approach is applied to facilitate the development process of assessing the reliability and robustness of the APR1400 IVR strategy. Specifically, the Kossiakoff SE method is used to identify the requirements, functions and physical architecture, and to develop a design verification and validation plan. Using the SE approach provides a systematic tool to successfully achieve the research goal by linking each requirement to a verification or validation test with predefined success criteria at each stage of the model development. The developed model identified the conditions necessary for successful implementation of the IVR strategy which maintains the vessel integrity and prevents a melt-through.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Journal of the Ergonomics Society of Korea
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• v.32 no.1
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• pp.117-124
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• 2013

Objective: This paper presents additional considerations related to organization and safety culture extracted from recent human error incidents in Korea, such as station blackout(i.e., SBO) in Kori#1. Background: Safety culture has been already highlighted as a major cause of human errors after 1986 Chernobyl accident. After Fukushima accident in Japan, the public acceptance for nuclear energy has taken its toll. Organizational characteristics and culture became elucidated as a major contributor again. Therefore many nuclear countries are re-evaluating their safety culture, and discussing any preparedness and its improvement. On top of that, there was an SBO in 2012 in the Kori#1. Korean public feels frustrated due to the similar human errors causing to a catastrophe like Fukushima accident. Method: This paper reassesses Japan's incidents, and revisits Korea's recent incidents. It focuses on the analysis of the hazards rather than the causes of human errors, the derivation of countermeasures, and their implementation. The preceding incidents and conclusions from Japanese experience are also re-analyzed. The Fukushima accident was an SBO due to the natural disaster such as earthquakes and a successive tsunami. Unlike the Fukushima accident, the Kori#1 incident itself was simple and restored without any loss and radioactive release. However, the fact that the incident was deliberately concealed led to massive distrust. Moreover, the continued violation of rules and organized concealment of the accident are serious signs of a new distorted type of human errors, blatantly revealing the cultural and fundamental weakness of the current organization. Result: We should learn from Japanese experiences who had taken pride in its safety technology and fairly high confidence in safety culture. Japan's first criticality accident in JCO facility splashed cold water on that confidence. It has turned out to be a typical case revealing the problems in the organization and safety culture. Since Japan has failed to gain lessons and countermeasure, the issue persists to the Fukushima incident. Conclusion: Safety culture is not a specific independent element, which makes it difficult to either evaluate it properly or establish countermeasures from the lessons. It may continue to expose similar human errors such as concealment of incident and manipulation of bad data. Application: Not only will this work establish the course of research for organization and safety culture, but this work will also contribute to the revitalization of Korea's nuclear industry from the disappointment after the export contract to UAE.

• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.17 no.2
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• pp.245-250
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• 2017

From the instant of installation and operation, power cables start deteriorating. Cable systems can be maintained not only by monitoring the insulation status of the insulation layer and oversheath, but also the insulation status of the terminal and junction in high-voltage power cables. When the cable system (the cable itself and cable junctions combined) deteriorates, fire accidents happen due to dielectric breakdowns. We have invented a device to monitor the deteriorating status of cables, and installed it at Korea Western Power Co. Ltd. located in Taean, Chungcheongnam-do Province. In this paper, we present the results obtained using our device, through analysing and calculating the standard deviation of leakage current from cable insulators attached to the cables. When the standard deviation of analysed leakage current falls below a critical value, a cable system is deemed to be operating safely. But when the standard deviation of analysed leakage current is larger than the critical value, the insulation status of the terminal and junction in the cable system is considered to have seriously deteriorated. The terminal and junction in the relevant system should then be replaced preemptively in order to prevent blackout accidents of cables caused by the suspension of power supply.