• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3256-3263
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• 2021

The main steam line break accident is an essential initiating event of the pressurized water reactor. In present work, the fuzzy set theory and the signal-based fault detection method has been used to detect the occurrence and diagnosis of the location and break area for the small scale MSLB. The models are validated by the AP1000 accident simulator based on MAAP5. From the test results it can be seen that the proposed approach has a rapid and proper response on accident detection and location diagnosis. The method proposed to evaluate the break area shows good performances for small scale MSLB with the relative deviation within ±3%.

• 한국전산유체공학회지
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• 제20권3호
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• 에너지공학
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• 제12권2호
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• pp.93-100
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• 2003

An estimation methodology of the mass and energy (M/E) release due to the main steam line break (MSLB) has been developed with the RETRAN-3D code. In the case of equipment qualification (EQ), the over-estimated temperature would exceed the design limits of some cables or valves. In order to have a more flexible EQ profiles from the MSLB M/E release, the methodology with the best-estimated code was used. The major conditions affecting the MSLB M/E were found to be the initial SG level, heat transfer between primary and secondary sides, power level, operable protection system, main or auxiliary feedwater availability, and break conditions. The RETRAN-3D models were developed for the Kori unit 1 (KRN-1) which is typical two loop Westinghouse (WH) designed plant. Particularly, a detailed model of the steam generators was developed to estimate a more realistic two-phase heat transfer effect of the steam flow. After the modeling, the methodology has been developed through the sensitivity analyses. The M/E release data generated from the analyses have been used as the input to the inside containment pressure and temperature (P/T) analysis. According to the results at the point of view containment P/T, the Kori unit 1 can have more margin of 5∼15 ㎪ in pressure and 8∼15$^{\circ}C$ in temperature.

• 에너지공학
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• 제23권4호
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• pp.112-122
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• 2014

A main steam line break (MSLB) test at the ATLAS facility was simulated using the best-estimate thermal-hydraulic system code, MARS-KS. This has been performed as an activity at the third domestic standard problem for code benchmark (DSP-03) that has been organized by Korea Atomic Energy Research Institute (KAERI). The results of the MSLB experiment and the MARS input data prepared for the previous DSP-02 using the ATLAS facility were provided to participants. The preliminary MSLB simulation using the base input data, however, showed unphysical results in the primary-to-secondary heat transfer. To resolve the problems, some improvements were implemented in the MARS input modelling. These include the use of fine meshes for the bottom region of the steam generator secondary side and proper thermal-hydraulics calculation options. Other input model improvements in the heat loss and the flow restrictor models were also made and the results were investigated in detail. From the results of simulations, the limitations and further improvement areas of the MARS code were identified.

• 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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• 제47권4호
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• pp.434-442
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• 2015

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

• 에너지공학
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• 제12권2호
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• pp.101-108
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• 2003

The purpose of this study is to develop an entrainment model of the Pressurized Water Reactor (PWR) U-tube Steam Generator (SG) for Main Steam Line Break (MSLB) analyses. Generally, the temperature of the inside containment vessel at MSLB is decreased by introducing the liquid entrainment effect. This effect makes a profit on the aspect of integrity evaluation for Equipment Environmental Qualification (EEQ) in the containment. However, the target plant, Kori unit 1 does not have the entrainment data. Therefore, this study has been performed. RETRAN-3D and LOFTRAN computer programs are used for the model development. There are several parameters that are used for the initial benchmark, such as Combustion Engineerings (CE) experimental data and the RETRAN-3D model which describes the test leg. A sensitivity study is then performed with this model in which the model parameters are varied until the calculated results provide reasonable agreement with the measured results for the entire test set. Finally, a multiplication factor has been obtained from the 95/95 values of the calculated (best-estimate) quality data relative to the measured quality data. With this new methodology, an additional temperature margin of about 40$^{\circ}C$ can be obtained. So, the new methodology is found to have an explicit advantage to EQ analyses.

• 대한기계학회논문집
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• 제7권3호
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• pp.344-352
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• 1983

격납 용기 내에 비응축성 가스(공기)가 존재하는 경우에 증기의 응축 열전달 계수를 평가하는 방 법을 연구하였다. 유일한 대규모 격납 용기 실험인 CVTR자료를 이용하여 응축 열전달 계수를 계산하여, 현재 원자력 발전소의 냉각재 상실 사고(LOCA) 및 주 증기 배관 파열사고(MSLB)시에 격납 용기의 안전 해석에서 공식적으로 사용되고 있는 Tagami와 Uchida열전달 계수 관계식과 비교해 본 결과 좋은 일치를 보여 주었다.

• 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)

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.469-478
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• 2005

The subchannel analysis capability of MARS, a multi-dimensional thermal-hydraulic system code, has been enhanced. In particular, the turbulent mixing and void drift models for the flow-mixing phenomena in rod bundles were improved. Then, the subchannel analysis feature was combined with the existing coupled system thermal-hydraulics (T/H) and 3D reactor kinetics calculation capability of MARS. These features allow for more realistic simulations of both the hot channel behavior and the global system T/H behavior. Using the coupled features of MARS, a coupled analysis of a main steam line break (MSLB) is carried out for demonstration purposes. The results of the calculations are very reasonable and promising.