• Nuclear Engineering and Technology
• /
• 제26권1호
• /
• pp.126-139
• /
• 1994

본 연구에서는 BETHSY 실험장치에서 수행한 6" 소형 냉각재 상실사고(LOCA) 실험을 최적 열수력 코드인 CATHARE2 V1.2와 RELAP5/MOD3를 이용하여 계산했다. 본 연구의 주 목적은 소형 LOCA시 관심을 가지는 주요 물리현상인 이상 임계유동, 감압과정, 노심수위 감소, loop seal clearing 등에 대한 두 코드의 소형 LOCA 계산모의능력을 평가하는 것이다. 두코드는 이상 유동현상의 전개 경향이나 발생시점을 비교적 잘 예측하는 것으로 나타났고, CATHARE2의 경우가 실험과 더 잘 일치했다. 그렇지만 두 코드는 loop seal clearing 현상, loop seal clearing 발생후의 노심수위, accumulator 유량거동 등의 예측에는 약간의 편차를 보였는데, 편차의 정도는 RELAP5가 CATHARE2보다 더 큰 것으로 나타났다. 두 코드의 편차요인을 보다 상세히 분석하기 위하여 계면 마찰력, mesh크기, 파단노즐 junction에서의 방출계수(Discharge coefficient)등에 대하여 민감도분석을 수행하였다. 그 결과 CATHARE2의 경우는 계면 마찰력을 증가시킴으로써 감압과정시 일차계통의 질량분포, 즉 증기 발생기 입구 공동(SG inlet plenum)에서의 차압과 Cross√er leg의 차압이 개선되었으며, 증기발생기 외측 열전달계수를 증가시킴으로써 중기발생기의 압력변화를 개선할 수 있었다. RELAP5의 경우는 어떤 하나의 입력변수를 변화시켜서 과도기의 결과를 개선할 수 없었으며 다만, 계면 마찰력 모델링에 여전히 많은 불화실성이 내포되어 있음을 확인했다.확인했다.

• Nuclear Engineering and Technology
• /
• 제24권3호
• /
• pp.243-251
• /
• 1992

소형 냉각재 상실사고시 루프밀봉 형성 및 제거에 대하여 LSTF에서 수행된 실험 SB-CL-18의 결과를 RELAP5/MOD2와 /MOD3를 이용하여 예측하였다. 본 연구는 증기발생기 상향 및 하향 유동에서의 비대칭 냉각재수용에 따른 마노메트릭 유동에 의해 노심노출의 조기발생을 야기시키는 열수력학적 현상을 예측하기 위하여 수행되었다. RELAP5/MOD2를 이용한 해석결과는 루프밀봉 형성 및 제거를 포함하여 감압사고시의 주요 현상을 전반적으로 잘 예측하고 있으나 기초 계산외 결과를 볼 때 현상 및 시간적 순서에 관련하여 몇 가지의 차이가 있었다. RELAP5/MOD3는 RELAP5/MOD2보다 전반적인 현상, 특히 증기발생기 액체수용을 보다 잘 예측하고 있으며, 또 한 RELAP5/MOD3를 이용하여 증기발생기 U자관과 펌프 흡입관의 nodalization수를 늘린 경우는 루프 밀봉제거현상과 시간적 순서를 잘 예측할 수 있었다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.745-750
• /
• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

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

• Nuclear Engineering and Technology
• /
• 제44권7호
• /
• pp.719-726
• /
• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• Nuclear Engineering and Technology
• /
• 제31권1호
• /
• pp.17-28
• /
• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• Nuclear Engineering and Technology
• /
• 제50권6호
• /
• pp.829-841
• /
• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
• /
• 제49권5호
• /
• pp.928-940
• /
• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• 에너지공학
• /
• 제28권3호
• /
• pp.26-33
• /
• 2019

ATLAS와 LSTF 장치에서 수행된 저온관(CL) 파단 13% 및 17% IBLOCA 대응실험들을 비교하고 특히, 핵심 관심 인자인 노심 첨두피복온도(PCT)에 대하여 비교 검토하고 아울러 주요 열수력 현상에 대하여 토론하였다. 비교.검토에서 두 건의 CL 파단 IBLOCA 대응실험들은 PCT 거동에 있어서 꽤 큰 차이를 보이고 있는 것을 확인하였는데 이는 두 장치의 척도 차이로 인한 왜곡현상을 벗어나는 경향을 보이고 있다는 점에서 두 장치의 원자로냉각재계통에 대한 자세한 설계 비교를 수행하였다. 이에 두 장치 사이에 핵연료조정판(FAP) 설계에 있어서 차이가 있다는 점을 확인하였다. 이에 따라 IBLOCA 사고시 Reflux 응축수의 노심 유입에 중요한 역할을 하는 CCFL 관련 무차원직경 값에서도 두 장치에서 매우 다른 차이를 보이고 있다는 점에서 CL 파단 IBLOCA 대응실험에서의 PCT 거동의 현격한 차이를 설명할 수 있는 원인일 수 있는 인자라는 것을 발견하였다. 향후 관련 설계 차잇점을 근거로 더 자세한 검토와 분석을 통해 관련 현상을 이해할 수 있을 것으로 판단된다.

• Nuclear Engineering and Technology
• /
• 제41권5호
• /
• pp.655-676
• /
• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.