• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.131.1-131
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.447-452
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• 1996

Realistic containment analyses for large LOCA using a merged torsion of RELAP5/CONTEMPT4 are conducted. Analyzed are Generic LOCA with respect to the mass and energy releases from the RCS and containment pressure and temperature behaviors. The break locations considered are the double-ended guillotine breaks at the RCP discharge and hot legs for UCN 3&4 plants. For discharge leg break. the predicted containment pressure and temperature reach a peak during blowdown phase, thereafter the pressure and temperature decrease gradually without the second reflood peak. For the hot leg break it is found that the bypass break flow through the broken steam generator-during post-blowdown is negligibly small so that the containment atmosphere is not pressurized after the end of blowdown.

• Nuclear Engineering and Technology
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• v.12 no.1
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• pp.9-18
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• 1980

Reflood experiments under atmospheric pressure have been conducted with a single heated tube to investigate basically the rewetting phenomena following a LOCA. Experimental conditions are 180cm length of test tube, wall temperature range of 300-80$0^{\circ}C$, coolant flooding rate of 5-30cm/sec. and subcooling of 35-85$^{\circ}C$. Experiments show that the rewetting velocity is dependent on the initial wall temperature of test tube, coolant flow rate and coolant subcooling. It is required to develop the proper method to evaluate the rewetting temperature and the heat transfer coefficient.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.181-193
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• 1998

The COBRA-TF liquid droplet entrainment models have been assessed and improved through various experiments. The COBRA-TF code uses the Wurtz entrainment model in the film mist flow regime and the mechanistic model based on the critical Weber number and critical vapor velocity in the hot wall flow regimes, respectively. The Wurtz model has been replaced with the modified Sugawara model. The assessment against the experiments by Hewitt, Keeys, Yanai, and Whalley showed the modified Sugawara model better predicts the steam-water as well as the air-water experiments for the film mist flow regime. For hot wall flow regime, the COBRA-TF entrainment model was modified using two methods, one with an increased critical Weber number and the other with the Yonomoto's critical vapor velocity model. The modified models were assessed using the FLECHT-SEASET bottom reflood tests. The results showed that the Yonomoto model best predicts the quenching time, whereas the local maximum rod temperature was not affected much.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.707-712
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• 1998

이종 코드에 의한 CE형 발전소의 대형 냉각재 상실 사고 해석이 수행되었다. 이 연구는 상대적으로 최근에 개발된 웨스팅하우스 대형 냉각재 상실 사고 해석 코드를 사용하여 영광 3&4호기의 대형 냉각재 상실 사고를 계산해 봄으로써 CE 대형 냉각재 상실 사고 해석 코드의 개선 방향을 고찰하는 것을 목적으로 하였다. 계산은 가장 제한적인 대형 냉각재 상실 사고의 Blowdown 및 Refill 기간 동안 수행하였다. 이 기간 동안의 RCS내 열수력적 거동 및 연료봉 온도 변화는 CE 대형 냉각재 상실 사고 해석 코드를 사용하여 계산한 경우와 크게 다르지 않음을 확인하였다. 따라서 향후 CE 대형 냉각재 상실 사고 해석 코드의 성능 개설은 Reflood 해석용 코드의 개선 및 개발을 중심으로 이루어져야 한다는 결론을 얻었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.627-632
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• 1997

FLECHT-SEASET 실험을 이용하여 냉각재상실사고시 Reflood에 대한 TRAC-PF1 전산코드의 예측 능력을 평가하였다. FLECHT-SEASET 실험 장치는 3.657m(12 ft) 높이 161개 전열 봉으로 이루어 져 있으며, 다양한 재관수율, 계통압력, 초기 피복재온도, 재관수온도 노심내 반경방향 출력분포 둥의 조건에 따라 수행된 실험이다. TRAC-PF1은 비균질 비평형 이상유동 열수력(Nonhomogeneous Non-equilibrium Two-Fluid Hydrodynamic)모델을 사용하고 원자로 압력용기는 3차원으로 모델할 수 있는 최적전산코드로서, 이 평가 계산에는 HP Version이 사용되었다. 본 연구에서는 재관수율 변화에 따라 달라지는 연료봉 최대 피복재온도와 Quench 시간에 대한 TRAC-PF1 전산코드의 예측 능력을 중점적으로 평가하였다. 계산 결과 TRAC-PF1은 최대 피복재온도는 약 20-100$^{\circ}$K 낮게, Quench 시간은 실험치와 비교하여 약 40-150초 정도 늦게 예측하는 것으로 나타났는데, 재관수율이 낮을수록 최대피복재 온도는 낮게, Quench 시간은 늦게 예측하는 경향을 보이고 있다. 또한 재관수율이 3 in/sec 이상에서 노심 상부가 일찍 Quenching 되는 것으로 계산되는데, 이는 노심상부 열전달 Regime의 부적절한 계산이 원인으로 보인다.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.200-208
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• 1986

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.684-701
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• 2016

The Gaussian process model (GPM) is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU) and Level 1 probabilistic safety assessment (PSA) success criteria definitions while dealing with a large number of uncertainties.

• Journal of Energy Engineering
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• v.9 no.3
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• pp.261-268
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• 2000

A heat transfer correlation to predict the vaporization of a water droplet in highly superheated steam during a loss-of-coolant accident(LOCA) of a nuclear power plant is provided. Vaporization of liquid fuel or water droplets in superheated air or steam and subsequent interface heat transfer between a liquid droplet and superheated gas is typically correlated by way of a Nusselt number as a function of Reynolds number, Prantl number, and in some cases including mass transfer number. Presently available correlations and experimental data of the evaporation of liquid droplets in air or steam are analyzed and a new Nusselt number correlation is proposed taking Schmidt number into consideration in order to account for binary diffusion of the vapor as well, Nu$\_$f/(1+B)$\^$0.7/=2+0.53Sc$\_$f/$\^$-1/5/Re$\_$M/$\^$$\sfrac{1}{2}$/Pr$\_$f/$\^$$\sfrac{1}{3}$/ for which properties are evaluated at film condition except the density of Reynolds number evaluated at ambient condition. Diverse correlations for various combinations of liquid and gas species are put into single equation. The blowing correction factor of (1+B)$\^$0.7/ is confirmed appropriate, and a criterion to distinguish so-called high- and low-temperature condition of ambient gas is set forth.