• 한국산학기술학회논문지
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• 제18권9호
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• pp.9-17
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• 2017

자동차 및 철도 등 수송 시스템에서 무인화 운전으로의 진전으로 인해 시스템 운영 시 안전성의 확보는 필수불가결한 요소로 간주되어 왔다. 자동차 안전설계를 뒷받침하기 위해 제정된 기능안전 표준인 ISO 26262에서는 위험원 분석 및 평가 그리고 안전 설계를 수행할 때 시스템 설계 정보를 적절하게 반영함으로써 안전성이 확보되는 자동차 시스템을 구현하기 위한 절차가 제시되어 있다. 이에 따라 위험원 분석에 관해 많은 연구가 이루어졌는데, 주로 이미 운영되고 있는 유사 시스템 사례에 의존하여 설계 정보를 활용하였다. 먼저 물리 구성품 수준에서 설계정보를 추출하고, 이로부터 기능 들을 역추적 한 후에 위험원을 식별하는 방법이 연구되었다. 이러한 방법은 빠르고 쉽게 위험원의 식별이 가능하기는 하지만, 설계 요구사항이 변경되거나 새로운 시스템을 설계할 때에는 설계 정보를 제대로 반영할 수 없어 일부 위험원이 누락될 수 있는 가능성이 있다. 이러한 점을 해결하기 위해서 본 논문에서는 기능안전표준에서 제시하는 안전수명주기 모델의 위험원 분석 단계에서 효과적인 방법을 연구하였다. 구체적으로 시스템 개념 설계를 Top-Down 방식으로 수행하면서 확보한 설계 정보를 위험원 분석에 적절하게 활용하는 방법을 제안하였다. 먼저 시스템 개념 설계를 수행하고, 획득된 기능 설계 결과를 분석하였다. 그러고 나서 기능 분석 결과를 활용하는 기능기반 Fault Tree Analysis 방법을 제시하고 위험원 분석을 수행하였다. 또한 자동차 시스템에서의 안전 설계 사례 연구를 통하여 본 논문에서 제시하는 방법이 대상 시스템의 설계 정보가 체계적으로 반영되어 누락 가능성이 줄어든 위험원 분석이 가능함을 보여 주었다.

• Nuclear Engineering and Technology
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• 제28권3호
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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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• 제41권8호
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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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• 제53권2호
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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.

• 한국화재소방학회:학술대회논문집
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• 한국화재소방학회 2003년도 춘계학술논문발표회논문집
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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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• 제55권6호
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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.

• 한국정보통신학회논문지
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• 제15권7호
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• pp.1423-1432
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• 2011

해양에서의 선박의 사고는 교통량의 증가와 더불어 끊임없이 증가하는 추세에 있으며, 해상에서의 안전운항을 확보하기 위하여 사고 발생 후의 대처도 중요하지만 해양사고를 미연에 방지하는 운항 안전을 위한 사전 감시기술이 부각되고 있는 실정이다. 따라서, 본 논문에서는 안전운항을 위한 통합적인 안전관리 플랫폼 기반의 시스템을 제안하였고, 항해 근무상태 감시 및 항해/추진/기관/발전 등 이상상태 및 오동작 상태 감시를 위한 선교운용 단말장치 및 경보시스템을 구현하였으며, 선박 탑재를 위하여 선급 표준에 근거한 성능 평가를 수행하여 실증화를 위한 기반을 갖추었다.

• 한국압력기기공학회 논문집
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• 제15권1호
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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.

• 대한안전경영과학회지
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• 제10권3호
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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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• 제40권4호
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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.