• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.403-410
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• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• 에너지공학
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• 제16권3호
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• pp.103-112
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• 2007

국내 1970년대에 설계 및 건설된 원자력발전소에 대해 침수분석을 수행한 결과 기기냉각수펌프 및 열교환기 건물, 주/보조건물, 중간건물 주증기 헤더 격실, 중간건물 주급수관 지역 및 하부층 등이 침수사고에 매우 취약하며 발전소 안전정지능력을 저해할 정도로 침수 영향이 심각한 것으로 판명되었다. 이들 지역에서의 침수원은 주급수관 파단이다. 현재 원자력발전소 내환경기기검증에서 주급수관 파단 방출량 계산은 수계산(Hand calculation)방법으로 Henry-Fauske 임계유량 모델 사용하고 있다. 이 방법은 배관파단 위치에서의 차압으로 계산되며, 실제 원자력발전소의 각종 제어로직에 의한 격리신호를 반영하지 못하므로 지나치게 보수적으로 파단 방출유량이 계산된다. 이러한 문제점을 개선하기 위해 원자력발전소 열수력계통 해석 전산코드인 RETRAN을 사용하여 원자력발전소 일/이차측 계통과 제어로직을 모사하고, 주급수관 파단 방출량 분석을 위한 입력가정과 해석방법을 개발하였다. 침수위 분석은 웨스팅하우스형 원자력발전소 격납건물 외부 하부격실에 대해 적용하였다. 전산코드 해석에서 각종 제어계통과 로직을 고려하였으며, 가장 제한적 사고조건을 계산하기 위해 노심출력, 파단형태, 면적, 위치 등의 조합으로 구성된 18개 사고 사례를 분석하였다. 그 결과 가장 제한적 사례 분석에서는 기존 수계산 분석에서보다 파단 방출유량이 크게 줄었고, 하부격실의 침수위도 상당히 낮아졌다.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2003년도 춘계 학술발표회 논문집
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• pp.371-376
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• 2003

원자력 발전소에서 격납건물 계통의 건전성 유지는 냉각재상실사고(Loss of Coolant Accident: LOCA) 및 주증기관 파단(Main Steam Line Break : MSLB) 사고와 같은 설계기준사고 시 격납건물의 최대 온도/압력을 평가하는 격납건물 성능 평가는 격납용기 내에 방사능 물질을 효율적으로 가두어 방사능 피해로부터 공공의 안전을 확보할 수 있느냐 하는 관건이다.(중략)

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.641-642
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• 2005

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.1-10
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• 2018

Structures, systems and components of nuclear power plants should be able to maintain safety even in the event of design-basis accidents such as high-energy line breaks. The high-pressure steam jet ejected from the broken pipe may cause damage to the adjacent structures. The ANSI/ANS 58.2 code has been adopted as a technical standard for evaluating the jet impingement load. Recently, the U.S. NRC pointed out the non-conservativeness of the ANSI/ANS 58.2, because it does not take into account the blast wave effect, dynamic behavior of the jet, and oversimplifies the shape and load characteristics of the supersonic steam jet. Therefore, it is necessary to improve the evaluation method for the high-energy line break accident. In order to evaluate the behavior of supersonic steam jet, an appropriate numerical analysis technique considering compressible flow effect is needed. In this study, numerical analysis methodology for evaluating supersonic jet impingement load was developed and verified. In addition, the conservativeness of the ANSI/ANS 58.2 model was investigated using the numerical analysis methodology. It is estimated that the ANSI jet model does not sufficiently reflect the physical behavior of under-expanded supersonic steam jet and evaluates the jet impingement load lower than CFD analysis result at certain positions.

• 한국안전학회지
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• 제16권4호
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• pp.29-38
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• 2001

The purpose of this paper is to evaluate the structural integrity of a reactor pressure vessel subjected to the pressurized thermal shock(PTS) during the transient events, such as main steam line break(MSLB) and small break loss of coolant accident(SBLOCA). For postulated surface or subsurface cracks, variation curves of stress intensity factor are obtained by using the three different methods, including ASME section XI code anlysis, the finite element alternating method and the finite element method. From the stress intensity factor curves, the maximum allowable nil-ductility transition temperatures(RT/NDT/) are determined by the tangent criterion and the maximum criterion for various crack configurations and two initial transient events. As a result of the analysis, it is noted that axial cracks have smaller maximum allowable RT$_{NDT}$ values than same-sized circumferential cracks for both the transient events in the case of the tangent criterion. Axial cracks have smaller RT$_{NDT}$ values than same-sized circumferential cracks for MSLB and circumferential cracks have smaller values than axial cracks for SBLOCA in the case of the maximum criterion.

• Nuclear Engineering and Technology
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• 제50권4호
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• pp.582-588
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• 2018

Accident diagnosis is one of the complex tasks for nuclear power plant (NPP) operators. In abnormal or emergency situations, the diagnostic activity of the NPP states is burdensome though necessary. Numerous computer-based methods and operator support systems have been suggested to address this problem. Among them, the recurrent neural network (RNN) has performed well at analyzing time series data. This study proposes an algorithm for accident diagnosis using long short-term memory (LSTM), which is a kind of RNN, which improves the limitation for time reflection. The algorithm consists of preprocessing, the LSTM network, and postprocessing. In the LSTM-based algorithm, preprocessed input variables are calculated to output the accident diagnosis results. The outputs are also postprocessed using softmax to determine the ranking of accident diagnosis results with probabilities. This algorithm was trained using a compact nuclear simulator for several accidents: a loss of coolant accident, a steam generator tube rupture, and a main steam line break. The trained algorithm was also tested to demonstrate the feasibility of diagnosing NPP accidents.

• 소음진동
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• 제6권6호
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• pp.827-834
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• 1996

본연구에서는 핵연료집합체의 검증계획의 일환으로 2차측 배관파단의 영향을 조사하였다. 원자로노심의 상세모델을 이용한 동적해석으로 배관파단에 의한 응답을 구하였다. 파단적 누설개념의 적용으로 10인치 이상의 고에너지 배관에 대하여 양단 파단이 설계에서 배제됨에 따라 본 연구에서는 주증기관과 급수관의 파단을 가정 하였다. 핵연료 집합체의 전단력, 굽힘모우멘트, 변위 및 지지격자체의 충격하중에 대하여 자세히 고찰하였고 이들 동적해석 결과를 이용하여 핵연료집합체의 구조적 건전성을 평가하였으며 사고조건에서 2차측 배관파단이 핵연료집합체의 구조적 건전성 에 미치는 영향을 검토하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.622-627
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• 1996

Detailed analysis of Feedwater Line Break (FLB) event for the fuel failure point of view are lack because the event was characterized as the increase in reactor coolant system (RCS) pressure. Up to now, the potential of the rapid system heatup case has been emphasized and comprehensively studied. The cooldown effects of FLB event is considered to be bounded by the Steam Line Break (SLB) event since the cooldown effect of SLB event is larger than that of the FLB event. This analysis provides a new possible path which can cause the fuel failure. The new path means that the fuel failure can occur under the heatup scenario because the Pressurizer Safety Valves (PSVs) open before the reactor trips. The 1000 MWe typical C-E plant FLB event assuming Loss of Offsite Power (LOOP) at the turbine trip has been analyzed as an example and the results show less than 1% of the fuel failure. The result is well within the acceptance criteria. In addition to that, a study was accomplished to prevent the fuel failure for the heatup scenario case as an example. It is found that giving the proper pressure gap between High Pressurizer Pressure Trip (HPPT) analysis setpoint and the minimum PSV opening pressure could prevent the fuel failure.