• Nuclear Engineering and Technology
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• 제56권11호
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• pp.4880-4886
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• 2024

The approach for determining the break size of recirculation loops in a multiple-loop power plant in the event of a loss of coolant accident (LOCA) is presented in this study. In this study, the MAAP5 simulation program was used. An approach to measuring the size of a crack or break in the cooling system is the temperature difference between the recirculation loops. This method does not require any additional facilities; it compares the temperatures of the cooling loops to determine which one has a rupture. The best data source was the loop monitoring system, which sends temperature data for analysis to the main control room. A real operating power reactor training simulator and the FSAR are applied to evaluate MAAP5, the methodology's engine. The results of the MAAP5 simulation code were consistent with those of the power plant simulator. Therefore, MAAP5 could produce enough analytical data to create the relationship diagram between temperature difference and break size. The study hypothesized that there exists a maximum value of temperature difference corresponding to each break size and suggested that applying the absolute maximum temperature difference can aid in identifying the break size. This approach proposes an assistive method for determining the size of a fracture or break in the recirculation system by leveraging the temperature difference between each loop. This approach eliminates the need for additional facilities, as temperature data from the recirculation loops can be transmitted to the main control room. After the reactor scram, operators can monitor the maximum temperature differences at the inlet to estimate the break size. Although the fitting curve used to preliminary estimate the Large Break Loss of Coolant Accident break size may overestimate the break size, it still provides valuable insights. This novel tool offers a rapid and comprehensive method for detecting LOCA events in the recirculation loops.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.193-206
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• 2022

Several algorithms for nuclear power plant (NPP) break event detection, isolation, localization, and size estimation are proposed. A break event can be promptly detected and isolated after its occurrence by simultaneously monitoring changes in the sensing readings and by employing an interquartile range-based isolation scheme. By considering the multi-sensor data block of a break to be rank-one, it can be located as the position whose lead field vector is most orthogonal to the noise subspace of that data block using the Multiple Signal Classification (MUSIC) algorithm. Owing to the flexibility of deep neural networks in selecting the best regression model for the available data, we can estimate the break size using multiple-sensor recordings of the break regardless of the sensor types. The efficacy of the proposed algorithms was evaluated using the data generated by Maanshan NPP simulator. The experimental results demonstrated that the MUSIC method could distinguish two near breaks. However, if the two breaks were close and of small sizes, the MUSIC method might wrongly locate them. The break sizes estimated by the proposed deep learning model were close to their actual values, but relative errors of more than 8% were seen while estimating small breaks' sizes.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.702-708
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• 2019

This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms. NARX neural network is able to directly deal with time-dependent signals for dynamic estimation of break sizes in real-time. The case studied is a LOCA in the primary system of Bushehr nuclear power plant (NPP). In this study, number of hidden layers, neurons, feedbacks, inputs, and training duration of transients are selected by performing parametric studies to determine the network architecture with minimum error. The developed NARX neural network is trained by error back propagation algorithm with different break sizes, covering 5% -100% of main coolant pipeline area. This database of LOCA scenarios is developed using RELAP5 thermal-hydraulic code. The results are satisfactory and indicate feasibility of implementing NARX neural network for break size estimation in NPPs. It is able to find a general solution for break size estimation problem in real-time, using a limited number of training data sets. This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr NPP.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.495-503
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• 2017

Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN) model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

• Communications for Statistical Applications and Methods
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• 제19권1호
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• pp.177-182
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• 2012

A structural break in the level as well as in the innovation variance has often been exhibited in economic time series. In this paper we propose robust unit root tests based on a sign-type test statistic when a time series has a shift in its level and the corresponding volatility. The proposed tests are robust to a wide class of partially stationary processes with heavy-tailed errors, and have an exact binomial null distribution. Our tests are not affected by the size or location of the break. We set the structural break under the null and the alternative hypotheses to relieve a possible vagueness in interpreting test results in empirical work. The null hypothesis implies a unit root process with level shifts and the alternative connotes a stationary process with level shifts. The Monte Carlo simulation shows that our tests have stable size than the OLSE based tests.

• Journal of Biosystems Engineering
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• 제6권2호
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• pp.65-76
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• 1982

현재(現在) 수입공급(輸入供給)되고있는 농용(農用)트랙터의 고장수리실태(故障修理實態)를 조사분석(調査分析)하여 트랙터의 국산화(國産化) 및 이용합리화(利用合理化)를 위한 기초자료(基礎資料)로 활용(活用)을 목적(目的)으로 1980년(年) 1년간(年間) 8개도(個道) 32개군(個那)에서 기장조사(記帳調査)한 결과(結果)는 다음과 같다. 가. 대형(大型)트랙터의 연평균고장빈도(年平均故障頻度)는 5.0회(回)/대(臺)이고, 소형(小型)은 2.3회(回)/대(臺)로 대형(大型)이 소형(小型)보다 2배이상(倍以上) 고장(故障)이 많았으며, 이용시간당빈도(利用時間當頻度)는 대형(大型)트랙터가 1. 11회(回)/100hr, 소형(小型)은 0.65회(回)/100hr이었다. 나. 전체고장중(全體故障中), 75.6%는 본체(本體)에서 발생(發生)되었으며 24.4%는 작업기(作業機)에서 발생(發生)되었고, 특(特)히 소형(小型)트랙터 작업기고장중(作業機故障中) 쟁기와 로타베이터 고장(故障)이 80%이상(以上)을 차지하였다. 다. 대형(大型)트랙터의 75%, 소형(小型)의 62%가 연(年) 1회이상(回以上)의 고장(故障)이 있었으며, 10회이상(回以上) 발생(發生)된 경우(境遇)도 9%나 되었다. 라. 월별고장발생분포(月別故障發生分布)는 트랙터이용시간(利用時間)과 비례(比例)하는 경향(傾向)을 나타내어 5월(月)에 가장 고장(故障)이 많았다. 마. 운전자(運轉者)의 연령(年齡)이 많거나 학력수준(學歷水準)이 높을 수록 고장빈도(故障頻度)도 높아서 50대(代) 연령(年齡)에 대졸자(大卒者)가 고장(故障)이 가장 많았다. 바. 고장원인(故障原因)은 운전자과실(運轉者過失)이 29.8%, 부품노후(部品老朽)가 30.2%, 품질불량(品質不量)이 20.6%로 나타났다. 사. 연평균수리회수(年平均修理回數)는 5.5회(回)/대(臺)였고 이중 공장수리(工場修理)가 55.6%였으며 자가수리(自家修理)는 44.4%였다. 아. 대당연간수리시간(憂當年間修理時間)은 대형(大型)트랙터가 30.6시간(時間)이었고, 소형(小型)이 19.9시간(時間)이었으며 수리(修理) 1회당소요시간(回當所要時間)은 평균(平均) 3.62시간(時間)으로 나다났다. 자. 대당연간수리비용(臺當年間修理費用)은 대형(大型)이 278천(千)원, 소형(小型)이 70천(千)원 이었고, 이용시간당수리비용(利用時間當修理費用)는 대형(大型)이 621원, 소형(小型)이 198원이었다. 차. 트랙터 본체(本體)의 수리비계수(修理費係數)(Y)는 사용연수(使用年數)(x)'에 대하여 Y=0.752x의 상관(相關)을 가졌으며 총수리비계수(總修理費係數)는 내구년수(耐久年數)가 10년(年)일때 64%가 되었다.

• 한국조경학회지
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• 제24권3호
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• pp.109-114
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• 1996

Generally, the size of our recreation sites is selected through use demand at the peak day. At same time, scale economic and diseconomic are applied to a recreation site. If you are a rational decision-maker, you would like to select the size of your park at profit-maximization point. This study is the first try for modelling a Gain-Loss by the size options of a theme park. For testing the validity of a financial model to select the size of theme parks. Ever-Land, Seoul-Land, Dream-Land's operating size was analyzed. By the size options, the revenue of each park was compared with cost. The profit-maximization point and break-even point of each park were found. Ever-Land and Dream-Land's size was selected between the profit-maximization point and the break-even point. In contrast with Ever-Land and Dream-Land's, Seoul-Land's was selected upper the break-even point. To increase the utility of this model in selecting the size of a theme park, a decision-maker must keep in mind a few limits of this study. That is, 1) this model can not be applied at public parks. 2) Sometimes the others can be more important than financial revenue and cost. Finally, there is the validity of Gain-Loss Model in estimating only the financial revenues and costs through the size options.

• Nuclear Engineering and Technology
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• 제16권1호
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• pp.11-17
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• 1984

매우 작은 규모의 냉각재 상실 사고후($\leq$0.05ft$^2$) 어떤 일이 일어나는 가를 더 잘 이해하기 위해 고리 1호기에 대한 샘플 계산을 수행하였다 깨진 크기가 0.006 ft$^2$ 보다 큰 사고에 대해서는 냉각재 상실이 보충되는 양을 초과한다. 0.008 ft$^2$ 보다 큰 깨진 크기에 대해서는 잔열은 깨진 곳을 통해 완전히 제거된다. 이와 같은 결과에 비추어 고리 1호기는 매우 작은 규모의 냉각재 상실 사고의 전 영역에 걸쳐 비교적 안전하다고 결론지었다. 하지만, 900MWe 나 1200MWe 를 가진 원자로에 있어서, 어떤 깨진 크기에 대해서는 이 사고가 주의깊게 고려되어야 한다. 자연 순환에서 pool boiling 으로 또는 pool boiling에서 자연 순환으로 천이할때, 특별히 운전자와 안전 분석에 문제점을 남긴다. Primary pump shutoff, HPI pump shutoff, break isolation, opening relief valve의 운전자 간섭에 대해서도 논의 되었다. Shutoff 후 HPI pump의 연속적인 운전은 primary system의 건전성을 위협하지 않는다는 것이 증명되었다.

• Journal of Mechanical Science and Technology
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• 제15권1호
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• pp.10-20
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• 2001

The leak before break(LBB) concept is difficult to apply to a structure with a thin tube that is immersed in a water environment. A heat exchanger in a nuclear power plant is such a structure. The present paper addresses an application of the LBB concept to a heat exchanger in a nuclear power plant. The minimum leaked coolant amount(approximately 37.9 liters) containing the radioactive material which can activate the radiation detector device installed in near the heat exchanger is assumed. A postulated initial flaw size that can not grow to a critical flaw size within the time period to activate the radiation detector is justified. In this case, the radiation detector can activate the warning signal caused by coolant leakage from initially postulated flaws of the heat exchanger. The nuclear plant can safely shutdown when this occurs. Since the postulated initial flaw size can not grow to the critical flaw size, the structural integrity of the heat exchanger is not impeded. Particularly the informational scenario presented in this paper discusses an actual nuclear plant.

• 한국환경과학회지
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• 제4권5호
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• pp.481-488
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• 1995

This study was carried out to evaluate the variations of headloss rate and of specific deposit to depths with effective size of media and configuration of filter layer during algae blooming period. 0.51mm size media was disqualified because most of headloss occurred rapidly below 5cm from surface layer however 0.91mm size media acted deep filtration more than 20cm from top, as result 0.91mm sixte media filter had 2~3 times longer filtration time than 0.51mm sixte media filter, but 0.91mm size media have break-through potentiality. multi-layer filter with 1.02mm anthracite and 0.51mm sand had large deposit volume in upper layer that could longer filtration time, moreover smaller media in lower layer that could protect break-through.