• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.713-718
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• 1998

Henry-Fauske critical flow model was incorporated into TRAC-PF1 to correct some errors in the original TRAC-PFI critical flow model. Henry-Fouske mode1 was numerically implemented and tested against steady-state steam-water experimental data. The model was incorporated into TRAC-PFI and code assessment against Marviken Critical Flow Tests 15 and 24 was carried out. Calculations using RELAP5/MOD3 were also made for comparison. Ten cases were calculated each test and sensitivity study on nodalization as well as critical flow or model was performed Stand-alone numerical model test and code assessment were done for verification and validation of code modification. Calculation results show that the modified version of TRAC-PF1 has a capability to model critical flow correctly in various conditions.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.332-339
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• 2015

The leak before break (LBB) concept is widely used in designing pipe lines in nuclear power plants. According to the concept, the amount of leaking liquid from a pipe should be more than the minimum detectable leak rate of a leak detection system before catastrophic failure occurs. Therefore, accurate estimation of the leak rate is important to evaluate the validity of the LBB concept in pipe line design. In this paper, a program was developed to estimate the leak rate through circumferential cracks in pipes in nuclear power plants using the Henry-Fauske flow model and modified Henry-Fauske flow model. By using the developed program, the leak rate was calculated for a circumferential crack in a sample pipe, and the effect of the flow model on the leak rate was examined. Treating the crack morphology parameters as random variables, the statistical behavior of the leak rate was also examined. As a result, it was found that the crack morphology parameters have a strong effect on the leak rate and the statistical behavior of the leak rate can be simulated using normally distributed crack morphology parameters.

• 에너지공학
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• 제23권4호
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• pp.73-88
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• 2014

Various studies on the critical flow models for sub-cooled and/or saturated water were reviewed, especially on Fauske, Moody, and Henry for basic theoretical models; Zaloudek for insight into physical phenomena for a critical flow in an orifice type flow path; Sozzi & Sutherland for a critical flow test of saturated and sub-cooled water at high pressure for orifice and nozzles; and a Marviken test on a full-scale critical flow test. In addition, critical flow tests of sub-cooled water for the break simulators in integral effect test (IET) facilities were also investigated, and a hybrid concept using Moody's and Fauske's models was considered by the authors. In the comparison of the models for the selected test data, discussions of the effect of the diameters, predictions of the critical flow models, and design aspects of break simulator for SBLOCA scenarios in the IET facilities were presented. In the effect of diameter on the critical flow rate with respect to all dimensional scales, it was concluded that the effect of diameter was found irrespective of diameter sizes. In addition, the diameter effect on slip ratio affecting the critical flow rate was suggested. From a comparison of the critical flow models and selected test data, the Henry-Fauske model of the MARS-KS code was found to be the best model predicting the critical flow rate for the selected test data under study.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1264-1272
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• 2016

The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB) concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry-Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.

• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.