• Journal of the Ergonomics Society of Korea
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• v.15 no.1
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• pp.91-103
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• 1996

For determining thermal comfort properties of work suit in an atomic power plant, three different coverall ensembles (PVE, PET/Rayon, PP Nonwoven) were selected and the resistance to dry and evaporative heat transfer were measured for each ensemble by using a sweating thermal manikin. Also, PMV (Predicted Mean Vote) and PPD(Predicted Percentage of Dissatisfied) indices were predicted according to ISO 7730. As a result, ideal environmental conditions in an atomic power plant were suggested to make workers feel thermally comfortable. In addition, ideal intrinsic insulation values of coverall ensembles as a work suit under the present environmental conditions in the at6omic power plant were provided. The information given in this paper can be used to control environmental conditions in the atomic power plant thermally comfortable and to select a proper work suit for providing thermal comfort to the workers.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.5
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.265-273
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• 1996

Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.164-164
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• 1999

• Asian Journal of Innovation and Policy
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• v.5 no.1
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• pp.92-115
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• 2016

This paper presents how Korea succeeded in developing an indigenous nuclear power plant model over fifty years. Long-lasting national R&D for technical progress and the Korean government for managerial process were the two pillars in the build-up of indigenous Nuclear Power Plant (NPP) technological capabilities. The concept of technological capabilities is used to examine its evolutionary process with a qualitative and longitudinal approach. The government had a developing country ambition to formulate a strategic plan for technical self-reliance on nuclear power plant while establishing the country’s institutions and organization structure for the plan. Under the government leadership, it was national R&D that led to the resolution of a good number of technological problems, efficiently, by absorbing imported technologies and effectively adapting them to local circumstances.

• Corrosion Science and Technology
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• v.20 no.4
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.273-278
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• 1998

Belief network(or Bayesian network) based on Bayes' rule in probabilistic theory can be applied to the reasoning of diagnostic systems. This paper describes the basic theory of concept and feasibility of using the network for diagnosis of nuclear power plants. An example shows that the probabilities of root causes of a failure are calculated from the measured or believed evidences.

• 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.41 no.6
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• pp.849-858
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• 2009

Risk caused by safety-critical instrumentation and control (I&C) systems considerably affects overall plant risk. As digitalization of safety-critical systems in nuclear power plants progresses, a risk model of a digitalized safety system is required and must be included in a plant safety model in order to assess this risk effect on the plant. Unique features of a digital system cause some challenges in risk modeling. This article aims at providing an overview of the issues related to the development of a static fault-tree-based risk model. We categorize the complicated issues of digital system probabilistic risk assessment (PRA) into four groups based on their characteristics: hardware module issues, software issues, system issues, and safety function issues. Quantification of the effect of these issues dominates the quality of a developed risk model. Recent research activities for addressing various issues, such as the modeling framework of a software-based system, the software failure probability and the fault coverage of a self monitoring mechanism, are discussed. Although these issues are interrelated and affect each other, the categorized and systematic approach suggested here will provide a proper insight for analyzing risk from a digital system.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.544-549
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• 1995

An analysis for Steam Line Break (SLB) events which result in a return-to-power conditions after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip return-to-power SLB is quite different from that of a no return-to-power SLB and is more complicated. Therefore, it is necessary to develop an methodology to analyze the response of the NSSS parameter and the fuel performance for the post-trip return-to-power SLB events. In this analysis, the cases with and without offsite power were simulated by crediting 3-D reactivity feedback effect due to local heatup around stuck CEA and compared with the cases without 3-D reactivity feedback with respect to fuel performance, departure from nucleate boiling ratio (DNBR) and linear heat generation rate (LHGR).