• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.544-549
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• 1995

An analysis for Steam Line Break (SLB) events which result in a return-to-power conditions after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip return-to-power SLB is quite different from that of a no return-to-power SLB and is more complicated. Therefore, it is necessary to develop an methodology to analyze the response of the NSSS parameter and the fuel performance for the post-trip return-to-power SLB events. In this analysis, the cases with and without offsite power were simulated by crediting 3-D reactivity feedback effect due to local heatup around stuck CEA and compared with the cases without 3-D reactivity feedback with respect to fuel performance, departure from nucleate boiling ratio (DNBR) and linear heat generation rate (LHGR).

• Nuclear Engineering and Technology
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• 제28권6호
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• pp.563-575
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• 1996

A total of 77 cases was identified to have human errors from a total of 255 trips occurred from 1978 to 1992 in Korean NPPs. The cases were analyzed to investigate how many human errors occurred on which work conditions to find out the areas of high priority for human error reduction. For the analysis of the 77 trip cases due to human errors, classifications were made for the following four categories ; plant systems, work situation, job types, and error types. Erroneous tasks critically influencing the plant trips were carefully identified and analyzed according to the classifications. Based on the results for the individual cases, the cases were counted for the classification items in each of the four categories, then also for the group of categories to investigate the relationships among the categories in aspects of human error occurrences. As results, the plant systems, work situations, and job types, and error types that are dominant in human errors related to the trips ore identified.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1587-1596
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• 2020

The location selection for nuclear power plants (NPP) is a strategic decision, which has significant impact operation of the plant and sustainable development of the region. Further, the ranking of the alternative locations and selection of the most suitable and efficient locations for NPPs is an important multi-criteria decision-making problem. In this paper, the non-sequential Monte Carlo probabilistic method and the Latin hypercube sampling probabilistic method are used to evaluate and select the optimal locations for NPP. These locations are identified by the power plant's onsite loads and the average of the lowest number of relay protection after the NPP's trip, based on electricity considerations. The results obtained from the proposed method indicate that in selecting the optimal location for an NPP after a power plant trip with the purpose of internal onsite loads of the power plant and the average of the lowest number of relay protection power system, on the IEEE RTS 24-bus system network given. This paper provides an effective and systematic study of the decision-making process for evaluating and selecting optimal locations for an NPP.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.483-484
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.111.2-111
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• 2001

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전력기술부문
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• pp.110-112
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• 2005

Recently, the reactor trip in nuclear power plants has mostly occurred by the failure of secondary electrical facilities. Some kinds of technical standards or acceptable criteria are required to regulate those facilities. Therefore, in this paper the failures of secondary electrical facilities which cause reactor trip are analyzed, and the inspection technologies including guidelines for secondary electrical facilities are developed.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1184-1191
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• 2016

This paper introduces the development of a transient monitoring system to detect the early stage of a transient, to identify the type of the transient scenario, and to inform an operator with the remaining time to turbine trip when there is no operator's relevant control. This study focused on the transients originating from a secondary system in nuclear power plants (NPPs), because the secondary system was recognized to be a more dominant factor to make unplanned turbine-generator trips which can ultimately result in reactor trips. In order to make the proposed methodology practical forward, all the transient scenarios registered in a simulator of a 1,000 MWe pressurized water reactor were archived in the transient pattern database. The transient patterns show plant behavior until turbine-generator trip when there is no operator's intervention. Meanwhile, the operating data periodically captured from a plant computer is compared with an individual transient pattern in the database and a highly matched section among the transient patterns enables isolation of the type of transient and prediction of the expected remaining time to trip. The transient pattern database consists of hundreds of variables, so it is difficult to speedily compare patterns and to draw a conclusion in a timely manner. The transient pattern database and the operating data are, therefore, converted into a smaller dimension using the principal component analysis (PCA). This paper describes the process of constructing the transient pattern database, dealing with principal components, and optimizing similarity measures.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.430-436
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• 1993

원자력 발전소의 제어계측 장비의 노후화 때문에, 원자력 발전소에서는 제어계측 장비의 고장으로 인한 발전소 불시정지가 가끔 일어난다. 원자력 발전소의 불시정지는 커다란 경제적 손실과 발전소의 신뢰도를 떨어뜨리고 대중적인 인식을 좋지 않게 한다. 그러므로 불필요한 불시정지를 방지하기 위해서 제어계측 부품의 노후화를 고려한 부품교체를 해주는 것이 필요하다. 이 논문은 월성 원자력 발전소의 제어 컴퓨터 부품의 최적교체주기에 대해 연구했다. 우리는 먼저 월성 디지탈 제어 컴퓨터 부품의 최적교체주기에 대한 수학적 모델을 유도했고, 해석적으로 최적교체주기를 계산했다. 우리는 그결과를 현재 월성 원자력 발전소에서 쓰이는 방법과 비교했다. 월성에서 사용되는 방법은 수학적 분석 방법이 아닌 경험적 지식을 토대로 한것이다. 결과적으로 해석적으로 얻어진 최적 교체주기와 월성에서 사용되는 교체주기와는 약간의 차이를 보이고 있다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2009년도 제40회 하계학술대회
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• pp.72_73
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• 2009

The industrial power system has the radial plant distribution system and domestic generators to supply the essential loads. When the system is isolated from the utility tie line, the system frequency drops resulting in the trip of generators. The load shedding scheme shall be properly designed to secure the essential load. In this paper two kinds of load shedding schemes, those are, the Fast Act Load Shedding(FALS) and Under Frequency Load Shedding(UFLS) are simulated and compared taking an example of petrochemical gas separation plant.