• 제목/요약/키워드: RETRAN

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Improvement in the DNBR Modeling of RETRAN for Safety Analyses of Westinghouse Nuclear Power Plants

  • Cheong, Ae-Ju;Kim, Yo-Han
    • Nuclear Engineering and Technology
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    • 제34권6호
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    • pp.596-609
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    • 2002

  • Korea Electric Power Research Institute has developed the in-house safety analysis methodologies for non-LOCA(Loss Of Coolant Accident) events based on codes and methodologies of vendors and Electric Power Research Institute . According to the new methodologies, analyses of system responses and calculation of DNBR(Departure from Nucleate Boiling Ratio) during the transient have been carried out with RETRAN code and a sub-channel analysis code, respectively. However, it takes too much time to calculate DNBR for each case using the two codes to search for the limiting case from sensitivity study. To simplify the search for the limiting case, accordingly, RETRAN code has been modified to roughly calculate DNBR using hot channel modeling. The W-3 correlation is already included in RETRAN as one of the auxiliary DNBR models. However, WRB-1 and WRB-2 correlations required to analyze some Westinghouse type fuels are not considered in RETRAN DNBR models. In this paper, the RETRAN DNBR models using the correlations have been developed and the partial and complete loss of forced reactor coolant flow events have been analyzed for Yonggwang units 1 and 2 with the new methodologies to validate the models. The results of the analyses have been compared with those mentioned in the chapter 15 of the Final Safety Analysis Report.

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최적 계통분석 코드를 이용한 웨스팅하우스형 원자력발전소 시뮬레이터용 핵 증기 공급 계통 열수력 프로그램 개발 (Development of An Nuclear Steam Supply System Thermal-Hydraulic Program for the Westinghouse Type Nuclear Power Plant Simulator Using A Best-Estimate Code)

  • 서재승;전규동;이명수;이용관
    • 한국시뮬레이션학회:학술대회논문집
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    • 한국시뮬레이션학회 2004년도 춘계학술대회 논문집
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    • pp.94-100
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    • 2004

  • KEPRI has developed an Nuclear Steam Supply System(NSSS) thermal-hydraulic simulation program (called ARTS-KORIl) based on the best-estimate system code, RETRAN, as a part of the development project for the KORI unit 1 nuclear power plant simulator. To develop the RETRAN code as an NSSS T/H engine for the simulator, a number of code modifications, such as simplifications and removing of discontinuities of the physical correlations, were made to satisfy the simulator requirements of robustness and real time calculation capability Some simplified models and a backup system were also developed to simulate some transients that cannot be efficiently calculated by the RETRAN part of ARTS-KORIl.

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