• 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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• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• 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.

• Toxicological Research
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• 제35권3호
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• pp.209-214
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• 2019

Cancer is the leading cause of mortality worldwide. In cancer progression, sex hormones and their receptors are thought to be major factors. Many studies have reported the effects of estrogen and estrogen receptors (ERs) in cancer development and progression. Among them, G protein-coupled estrogen receptor (GPER), a G protein-coupled receptor, has been identified as an estrogen membrane receptor unrelated to nuclear ER. The mechanism of GPER, including its biological action, function, and role, has been studied in various cancer types. In this review, we discuss the relation between GPER and estrogen or estrogen agonists/antagonists and cancer progression.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1916-1938
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• 2019

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

• Molecules and Cells
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• 제27권3호
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• pp.359-363
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• 2009

Chfr, a checkpoint with FHA and RING finger domains, plays an important role in cell cycle progression and tumor suppression. Chfr possesses the E3 ubiquitin ligase activity and stimulates the formation of polyubiquitin chains by Ub-conjugating enzymes, and induces the proteasome-dependent degradation of a number of cellular proteins, including Plk1 and Aurora A. While Chfr is a nuclear protein that functions within the cell nucleus, how Chfr is localized in the nucleus has not been clearly demonstrated. Here, we show that nuclear localization of Chfr is mediated by nuclear localization signal (NLS) sequences. To reveal the signal sequences responsible for nuclear localization, a short lysine-rich stretch (KKK) at amino acid residues 257-259 was replaced with alanine, which completely abolished nuclear localization. Moreover, we show that nuclear localization of Chfr is essential for its checkpoint function but not for its stability. Thus, our results suggest that NLS-mediated nuclear localization of Chfr leads to its accumulation within the nucleus, which may be important in the regulation of Chfr activation and Chfr-mediated cellular processes, including cell cycle progression and tumor suppression.

• Nuclear Engineering and Technology
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• 제23권3호
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• pp.285-298
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• 1991

전형적인 정성적 퍼지형태의 입력데이타를 가진, 주어진 사고진행사건수목의 일부분에 대하여 퍼지집합이론(fuzzy set theory)의 응용 예를 먼저 보여주고, 이 예를 통해서 퍼지집합이론을 사고 진행사건수목에 적용하기 위해 적절한 계산알고리즘을 찾아내고 또 예를 들어 설명하였다. 그리고, 간단한 예제에 사용한 계산절차를 많은 현상학적 불확실성 인자를 포함한 아주 복잡한 사고진행사건수목 즉, 최근 Zion 발전소 위험도평가(PRA)에 사용된 전형적인 발전소 손상군의 하나인‘SEC’에 응용해서 적용하였다. 퍼지집합이론으로 평가한 계산값들의 퍼지평균치들은 최근 통계적 PRA 평가 방법론으로 얻는 값들의 평균치와 거의 같은 결과를 보여주고 있다. 본 논문의 주요목적은 부정확하고 또 정성적인 분기점확률이나 또는 많은 현상학적 불확실성 인자들을 가진 사고진행사건수목들에 이 퍼지집합이론을 적용하기 위한 공식적 계산절차를 제공하는데 있다.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.