• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.126-139
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• 1994

Best-estimate thermal-hydraulic codes, CATHARE2 V1.2 and RELAP5/MOD3, hate been assessed against the BETHSY 6.2 tc six-inch cold leg break loss-of-coolant accident (LOCA) test. Main objective is to analyze the overall capabilities of the two codes on physical phenomena of concern during the small break LOCA i.e. two-phase critical flow, depressurization, core water level de-pression, loop seal clearing, liquid holdup, etc. The calculation results show that the too codes predict well both in the occurrences and trends of major two-phase flow phenomena observed. Especially, the CATHARE2 calculations show better agreements with the experimental data. However, the two codes, in common, show some deviations in the predictions of loop seal clearing, collapsed core water level after the loop seal clearing, and accumulator injection behaviors. The discrepancies found from the comprision with the experimental data are larger in the RELAP5 results than in the CATHARE2. To analyze the deviations of the two code predictions in detail, several sensitivity calculations have been performed. In addition to the change of two-phase discharge coefficients for the break junction, fine nodalization and some corrections of the interphase drag term are made. For CATHARE2, the change of interphase drag force improves the mass distribution in the primary side. And the prediction of SG pressure is improved by the modification of boundary conditions. For RELAP5, any single input change doesn't improve the whole result and it is found that the interphase drag model has still large uncertainties.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.745-750
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• 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
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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.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.719-726
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• 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
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• v.31 no.1
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• pp.17-28
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• 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
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• v.50 no.6
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• pp.829-841
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• 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
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• v.49 no.5
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• pp.928-940
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• 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.

• Journal of Energy Engineering
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• v.28 no.3
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• pp.26-33
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• 2019

A comparison of CL 13% and 17% IBLOCA counterpart tests(CPTs) between the ATLAS and LSTF facilities was carried out and the behavior of peak cladding temperatures(PCTs) and related thermal hydraulic phenomena were investigated and discussed. There appeared quite a big difference in PCT behavior between the two CPTs and a further comparison of reactor coolant system design between the two facilities was performed. As a result, there was a difference in fuel alignment plate (FAP) design, e.g., one FAP in ATLAS, a combination of upper core plate and upper end box in LSTF, respectively. The FAP design mainly affects the reflux condensate behavior in IBLOCA tests and any difference in FAP design can be a possible reason for different PCT behavior between the two facilities. It should be a further study to find the reason of different PCT behvior between the two facilites.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.655-676
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• 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.