• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1814-1820
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• 2023

Safety classification of systems, structures, and components (SSC) is an essential activity for nuclear reactor design and operation. The current regulatory trend is to require risk-informed safety classification that considers first, the severity, but also the frequency of SSC failures. While safety classification for nuclear power plants is covered in many regulatory and scientific publications, research reactors received less attention. Research reactors are typically of lower power but, at the same time, are less standardized i.e., have more variability in the design, operational modes, and operating conditions. This makes them more challenging when considering safety classification. This work presents the Integrated Risk-Informed Safety Classification (IRISC) procedure which is a novel extension of the IAEA recommended process with dedicated probabilistic treatment of research reactor designs. The article provides the details of probabilistic analysis performed within safety classification process to a degree that is often missing in most literature on the topic. The article presents insight from the implementation of the procedure in the safety classification for the MARIA Research Reactor operated by the National Center for Nuclear Research in Poland.

• 방사성폐기물학회지
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• 제14권4호
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• pp.423-434
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• 2016

인접 국가인 일본의 후쿠시마 원전에서 극한 자연재해로 인한 중대사고가 발생하면서, 국내에서 중대사고 및 확률론적 안전성 평가 (PSA, Probabilistic Safety Assessment)에 대한 중요성이 재인식되었다. 국내에서는 원전의 소외결말을 평가하는 3단계 PSA에 대한 연구개발이 최근까지 거의 이루어지지 않았다. 본 논문에서는 국외 3단계 PSA 전산코드 중, 미국의 MACCS2 (MELCORE Accident Consequence Code System 2), 유럽의 COSYMA (COde SYstem from Maria) 그리고 일본의 OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents)에 대한 간략한 분석과 미국의 MACCS2에 대한 단점 및 한계점 분석을 수행하였다. 국내 외 전문가들에 의해 공통적으로 지적되어 온 MACCS2의 한계점은 다수호기사고와 사용후핵연료 저장조로부터의 방출 모사의 불가능, 그리고 대기확산모델을 단순 가우시안 플륨모델을 기본으로 한다는 것이며, 이중 일부는 MACCS2업데이트 버전을 통해 개선되어 왔다. Food chain 모델의 모사의 제한, 해양 및 수계 확산모델의 부재, 제한된 범위의 경제영향평가 등 또한 개선되어야 할 사항이다. 기술보고의 결과는 국내 3단계 PSA 관련 기술 개발을 위한 기초자료로 활용될 수 있을 것으로 기대된다.

• 시스템엔지니어링학술지
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• 제16권2호
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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.