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논문 기본 정보

자료유형
학술저널
저자정보
황돈관 (포항공과대학교) 강순호 (한국원자력안전기술원) 최낙준 (포항공과대학교) 조항진 (포항공과대학교)
저널정보
한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology Vol.56 No.1
발행연도
2024.1
수록면
19 - 33 (15page)
DOI
10.1016/j.net.2023.08.023

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초록· 키워드

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In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphonbreaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as 〈jg〉/〈j〉 increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

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