• Nuclear Engineering and Technology
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• v.41 no.5
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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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.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2003.05a
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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)

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.641-642
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• 2005

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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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.

• Journal of the Korean Society of Safety
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• v.16 no.4
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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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• v.50 no.4
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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.

• Journal of KSNVE
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• v.6 no.6
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• pp.827-834
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• 1996

The effect of pipe breaks in the secondary side is investigated as a part of the fuel assembly qualification program. Using the detailed dynamic analysis of a reactor core, peak responses for the motions induced from pipe breaks are obtained for a detailed core model. The secondary side pipe breaks such as main steam line and economizer feedwater line braksare considered because leak-before-break methodology has provided a technical basis for the elimination of double ended guillotine breaks of all high energy piping systems with a diameter of 10 inches or over in the primary side from the design basis. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are carefully investigated. Also, the stress analysis is performed and the effect of the secondary side pipe breaks on the fuel assembly structural integrity under the faulted condition is addressed.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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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.