• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.657-662
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• 1997

계통분석 코드인 RELAP5/MOD3와 격납용기 분석 코드인 CONTEMPT4/MOD5에 피동형 격납용기 열전달 모델을 추가하여 개선한 CONTEMP4/MOD5/PCCS 코드를 이용하여 피동형원자로의 원자로 계통과 격납용기의 열수력 연계해석을 위한 통합코드를 구성하였다. 두 코드는 process 제어의 개념을 이용하여 각 코드의 특성을 유지시키면서 explicit coupling되게 하였으며 통합코드를 1000MWe급 피동형 원전의 냉각재 상실사고분석에 적용시켜 검증하였다 통합코드는 원자로 계통과 격납용기의 계산을 동시에 수행함으로써 얻을 수 있는 격납용기-계통 간의 열수력 현상을 파악 할 수 있게 하여줌으로써 피동형 원전의 열수력 분석도구로서 사용할 수 있는 것으로 분석되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.513-518
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• 1996

This report presents a demonstration of application of realistic evaluation methodology to a posturated cold leg large break LOCA in a Westinghouse three-loop pressurized water reactor with 17$\times$17 fuel. The new method of this analysis can be divided into three distinct step: 1) Best Estimate Code Validation and Uncertainty Quantification 2) Realistic LOCA Calculation 3) Limiting Value LOCA Calculation and Uncertainty Combination RELAP5/MOD3/K ［1］, which was improved from RELAP5/MOD3.1, and CONTEMPT4/MOD5 code were used as a best estimate thermal-hydraulic model for realistic LOCA calculation. The code uncertainties which will be determined in step 1) were quantified already in previous study ［2］, and thus the step 2) and 3) for plant application were presented in this paper. The application uncertainty parameters are divided into two categories, i.e. plant system parameters and fuel statistical parameters. Single parameter sensitivity calculations were performed to select system parameters which would be set at their limiting value in Limiting Value Approach (LVA) calculation. Single run of LVA calculation generated 27 PCT data according to the various combinations of fuel parameters and these data provided input to response surface generation. The probability distribution function was generated from Monte Carlo sampling of a response surface and the upper 95$^{th}$ percentile PCT was determined. Break spectrum analysis was also made to determine the critical break size. The results show that sufficient LOCA margin can be obtained for the demonstration NPP.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• 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.

• 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.26 no.4
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• pp.562-569
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• 1994

In this study improvement of transient analysis model, KOTRAC, for the passive reactor has been performed. In the KOTRAC, mixture drift flux model is adopted to simulate thermal hydraulic behavior, which can simulate ECCS injection in the passive plant. However, there is a difficulty to handle complete phase separation phenomena due to the near-zero density, which may occur in the pressurizer surge line or horizontal flow paths. In this study, a couple of model changes to over-come Courant limit feilure has been examined. One of key features is to substitute flow distribution parameters with Ishii's correlation. Corrected results are nil compared to those of RELAP/MOD3 analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.441-447
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• 1996

보수적 방법론(Evaluation Model)으로 계산된 냉각재상실사고 해석의 결과는 너무 비현실적이고 보수적이라는 문제점이 제기되어 왔으며, 이를 해결할 수 있는 방안으로 미국 원자력규제위원회(USNRC)에서는 1988년에 최적 방법론(Best Estimate Model)을 적용할 수 있도록 규정을 개정하였다. 이에 따라, 한국원자력연구소에서는 수정된 RELAP5/MOD3를 근간으로 대형냉각재 상실사고 최적 방법론을 개발하였다. 개발된 최적 방법론을 울진 3,4호기에 적용하여 해석을 수행하였으며 그 결과를 보수적 방법론으로 계산된 결과와 비교하여 주요 변수들의 거동을 분석하였다.

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.476-485
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• 1999

In order to evaluate the gravity-injection capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Yong Gwang Units 3&4 were reviewed. The six cases of possible gravity-injection paths from the refueling water tank (RWT) were identified and the thermal-hydraulic analyses were performed using the RELAP5/MOD3.2 code. The core cooling capability was significantly dependent on the gravity-injection path, the RCS opening, and the injection rate. In the cases with the pressurizer manway opening higher than the RWT water level, the coolant was held up in the pressurizer and the system pressure continued increasing after gravity-injection. The gravity injection eventually stopped due to the high system pressure and the core was uncovered. In the cases with the injection path and opening on the same leg side, the core cooling was dependent on whether the water injected from the RWT passed the core region or not. However, in the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. In addition, from the sensitivity study on the gravity-injection flow rate, it was found that about 54 kg/s of injection rate was required to maintain the core cooling and the core cooling could be provided for about 10.6 hours after event with that injection rate from the RWT. Those analysis results would provide useful information to operators coping with the event.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.37-50
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• 1993

To evaluate the sequence of event and the Thermohydraulic behavior during total loss of feedwater accident and recovery procedure, a RELAP5/MOD3 calculation is performed and compared with the LOFT L9-l/L3-3 experiment. Also, the predictability of the code for the major Thermohydraulic phenomena following the accident is assessed. As a result, it is found that a pressure control using the spray until the time the water level reaches the top of the pressurizer, an overpressure protection by pressurizer PORV, a recovery of the secondary heat removal capability by refilling steam generator, and an effective cooldown by the continued natural circulation can be performed without core uncovery. It is also found that the plant-specific evaluation is necessary to confirm the effectiveness of the current symptom-oriented emergency operating procedure, especially in an overpressure protection performance and steam generator recovery performance.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.