• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• 제46권2호
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• pp.207-216
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• 2014

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• Nuclear Engineering and Technology
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• 제38권6호
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• pp.535-550
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• 2006

The objectives of this paper are twofold to summarize the new findings and confirmed results from the Phebus FPT-1 experimental data and to report useful information to MELCOR users regarding the better use of MELCOR. For the core damage behavior, the early stage of a melt progression was predicted well; however, the late phase models, concerned with fuel dissolution, oxide cladding failure, fuel slumping, rubble debris heat up, effects of burn-up fuel, and so on, still showed limitations in MELCOR. For the fission product behavior, the comparison showed unexpected phenomena, various limitations, unresolved issues, and even absence of models. The issues summarized in this study have revealed the main areas where our endeavors need to be intensified in order to improve our understanding of severe accident phenomena. From the analysis of the Phebus FPT-1 test results, not only new core damage features, such as foaming or core expansion, but also possible new fission product release patterns due to effects from a high burn-up fuel have raised alternative challenging phenomena that should be solved in the next severe accident research phase.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.772-777
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• 1997

Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and in the hot leg and SG during the TMLB' scenrio. The objective of this study is to develop a natural circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.688-706
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• 2024

Austenitic stainless steels (ASS) are extensively employed in various sectors such as nuclear, power, petrochemical, oil and gas because of their excellent structural strength and resistance to corrosion. SS304 and SS316 are the predominant choices for piping, pressure vessels, heat exchangers, nuclear reactor core components and support structures, but they are susceptible to stress corrosion cracking (SCC) in chloride-rich environments. Over the course of several decades, extensive research efforts have been directed towards evaluating SCC using diverse methodologies and models, albeit some uncertainties persist regarding the precise progression of cracks. This review paper focuses on the application of Acoustic Emission Technique (AET) for assessing SCC damage mechanism by monitoring the dynamic acoustic emissions or inelastic stress waves generated during the initiation and propagation of cracks. AET serves as a valuable non-destructive technique (NDT) for in-service evaluation of the structural integrity within operational conditions and early detection of critical flaws. By leveraging the time domain and time-frequency domain techniques, various Acoustic Emission (AE) parameters can be characterized and correlated with the multi-stage crack damage phenomena. Further theories of the SCC mechanisms are elucidated, with a focus on both the dissolution-based and cleavage-based damage models. Through the comprehensive insights provided here, this review stands to contribute to an enhanced understanding of SCC damage in stainless steels and the potential AET application in nuclear industry.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.54-64
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• 1988

FLHT-2 실험 결과를 사용하여 원자력발전소의 중대사고발생시 노심의 거동을 해석하기 위한 전산코드인 SCDAP코드를 평가하였다. 계산결과에 의하면 코드는 실험시 측정된 노심의 온도경향, 총수소발생량 및 순간최대수소발생율, 그리고 연료봉내압과 피복재파열시간을 잘 예측하는 것으로 평가되었다. 그러나 이상유체높이와 복사열전달 및 zircaloy의 급격한 산화 시작 온도에 대한 모델은 수정되어야 할 것으로 평가되었다. 또한 핵 연료봉에서의 gap을 고려하여주는 것은 노심손상현상의 정확한 예측에 커다란 도움을 줄 수 있다는 것이 밝혀졌다.

• Molecules and Cells
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• 제46권4호
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• pp.231-244
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• 2023

Leucine-rich repeat containing 15 (LRRC15) has been identified as a contributing factor for cartilage damage in osteoarthritis; however, its involvement in rheumatoid arthritis (RA) and the underlying mechanisms have not been well characterized. The purpose of this study was to explore the function of LRRC15 in RA-associated fibroblast-like synoviocytes (RA-FLS) and in mice with collagen-induced arthritis (CIA) and to dissect the epigenetic mechanisms involved. LRRC15 was overexpressed in the synovial tissues of patients with RA, and LRRC15 overexpression was associated with increased proliferative, migratory, invasive, and angiogenic capacities of RA-FLS and accelerated release of pro-inflammatory cytokines. LRRC15 knockdown significantly inhibited synovial proliferation and reduced bone invasion and destruction in CIA mice. Runt-related transcription factor 1 (RUNX1) transcriptionally represses LRRC15 by binding to core-binding factor subunit beta (CBF-β). Overexpression of RUNX1 significantly inhibited the invasive phenotype of RA-FLS and suppressed the expression of proinflammatory cytokines. Conversely, the effects of RUNX1 were significantly reversed after overexpression of LRRC15 or inhibition of RUNX1-CBF-β interactions. Therefore, we demonstrated that RUNX1-mediated transcriptional repression of LRRC15 inhibited the development of RA, which may have therapeutic effects for RA patients.

• 대한임상검사과학회지
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• 제54권4호
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• pp.249-255
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• 2022

알츠하이머병은 주요한 공중보건 문제로 나타나며 연구분야에서도 최우선적인 과제이다. 알츠하이머병(AD)에서 뇌척수액(CSF)을 활용한 바이오마커인 아밀로이드-β(Aβ42), 총 타우(T-tau) 및 인산화 타우(P-tau)가 알츠하이머병 병태생리학의 핵심 요소를 반영한다. 임상 연구 및 새로운 측정법을 통한 임상적으로 활용되는 진단은 전임상 알츠하이머병에 대해 민감적이고 특이적이며 신뢰할 수 있는 바이오마커의 발굴, 뿐만 아니라 치매의 조기 발견 및 감별 진단과 질병 진행 모니터링에 도움이 되는 검사법의 개발에도 중요할 것이다. 증상 전 단계에서 AD의 조기 발견은 시냅스 손상 및 신경 손실이 확장되기 전에 개입이 수행되기 때문에 치료 개입을 조기에 가능하게 하고 치료 성공을 위한 가능성이 더 큰 좋은 기회로 이어진다. 따라서 새롭고 접근하기 쉽고 비용이 적게 드는 바이오마커를 임상 진단에 활용하는 것이 매우 유익할 것이다. 치매의 초기단계에 일어나는 병리학적 변화나, 질병의 진행정도를 추적할 수 있는 다양한 바이오마커들의 진단방법을 찾는 일은 치료제 개발처럼 중요한 연구 분야이다. 조기진단을 위해 임상증상을 대변하거나(surrogate marker), 증상이 나타나기 이전 상태를 측정할 수 있는 새로운 진단마커가 필요한 상황이다. 이러한 이유로 인지기능 저하정도를 측정하여 정상, 경도인지장애(mild cognition impairment, MCI) 및 전임상(preclinical) 상태의 사람을 판별할 수 있는 바이오마커(biomarker)를 활용한 조기진단법 개발의 중요성이 강조되고 있다.