• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.257-264
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• 1996

The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.498-508
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• 2021

Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical-CO₂-cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints. As the MMR is in its conceptual design phase, design weaknesses and valuable safety insights could be identified during PSA. Level 1 internal event PSA was carried out involving literature survey, system characterization, identification of initiating events, transient analyses, development of event trees and fault trees, and quantification. The initiating events and scenarios significantly contributing to core damage frequency (CDF) were determined to identify design weaknesses in MMR. The most significant initiating event category contributing to CDF was the transients with the power conversion system initially available category, owing to its relatively high occurrence frequency. Further, an importance analysis revealed that the safety of MMR can be significantly improved by improving the reliability of reactor trip and passive decay heat removal system operation. The findings presented in this paper are expected to contribute toward future applications of PSA for assessing unconventional nuclear reactors in their conceptual design phases.

• Proceedings of the Korea Institute of Fire Science and Engineering Conference
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• 2003.04a
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• pp.259-264
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• 2003

The fire protection system of Nuclear Power Plants(NPPs) is an integrated system that is applied multi-field technology. So, it needs synthetic design and analysis, that is, the plan of fire protection, fire compartment, fire detection, fire suppression, and success of safety shut down, etc. In case of a fire in NPPs, secure the safety of reactor and minimize the radioactivity contamination. For this purpose, perform the fire risk analysis and make up the deducted problem through the improvement of design or the change of operation process.

• 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 Korea Institute of Information and Communication Engineering
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• v.15 no.7
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• pp.1423-1432
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• 2011

Since the marine accidents increase with the increased volume of traffic, preventive surveillance technology for safety navigation of the ships before the accident is being emphasized to secure the safety at sea along with post-accident measures. This paper aims to suggest a design based on an integrated safety management platform systems to support nautical safety, implements of bridge operation terminal equipments and alarm system for bridge watch monitoring and abnormal state of navigation/propulsion/machinery/power, and performs a quality evaluation for the actual boarding on the ship based on the classification standards.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.33-39
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• 2019

An 1,500MW advanced power reactor required the standard design approval by a Korean regulatory body in 2014. The reactor has been designed to have a 4-train independent safety concept and a passive auxiliary feedwater system (PAFS). The full power risk or core damage frequency (CDF) of 1,500MW advanced power reactor has been reduced more than that of APR1400. However, the risk during the low power and shutdown (LPSD) operation should be reduced because CDF of LPSD is about 4.7 times higher than that of internal full power. The purpose of paper is to analysis design alternatives to reduce risk during the LPSD. This paper suggests design alternatives to reduce risk and presents sensitivity analysis results.

• Journal of the Korea Safety Management & Science
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• v.10 no.3
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• pp.1-8
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• 2008

Risk is the probability of an adverse event given exposure to hazard. There are many reason for unsafety situation without safety operation. The reason is no safety evaluation system in small enterprise. And then this study purposes safety management activities that is evaluation system for total safety efficiency's maximization. Therefore, in this study, this model that can evaluate quantitative activities in small enterprise that maximize safety efficiency wishes to do design using balanced scorecard. In other words, this study aims to suggest a performance measurement model reflecting the characteristics of safety evaluation system, especially the model for return manufacturing related to safety, and to develop the S-BSC(Safety-Balanced ScoreCard) measurement model using a weight lifetime value to which a relative weight is applied by using AHP based on the BSC.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.269-276
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• 2008

Operational license of WWER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Prolongation by additional 20 years of the operational lifetime is feasible. Moreover, enhancement of the reactor thermal power by 8% will increase both the net power output and the competitiveness of the plant. Paks NPP is a pioneer considering the power up-rate and preparation of long-term operation of WWER-440/213 design. Systematic preparatory work for long-term operation of Paks NPP has been started in 2000. A regulatory framework and a comprehensive engineering practice have been developed. According to the authors view, creation of a gapless engineering system via consequent application of best practices, and feed-back of experiences together with proper consideration of WWER-440/V213 features are the decisive elements of ensuring the safety of long-term operation. That systematic engineering approach is in the focus of recent paper. Key elements of justification and measures for ensuring the safety of long-term operation of Paks NPP WWER-440/213 units are identified and discussed. These are the assessment of plant condition and review of adequacy of ageing management programmes, also the review, validation and reconstitution of time limited ageing analyses as core tasks of licence renewal.

• Proceedings of KOSOMES biannual meeting
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• 2003.11a
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• pp.93-97
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• 2003

This study conducted a basic study to develop ISM Code operation evaluation module, which is part of development of Safety Management System(ISM Code) Implementation supporting module. In particular, the significant of this study is that it established AHP operation evaluation analysis process, the design of questionnaire, and the hierarchical structure of operation evaluation model standard, focusing on the development of ISM Code operation evaluation module. Also, this study has its value in the point that it attempted to apply AHP scheme, which is part of the main decision-making in management science field, to develop operation evaluation module.