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          NRC Library  Document Collections  NUREG-Series Publications  Publications Resulting from International AgreementsAssessment of RELAP5/MOD3.2.2 Gamma With the LOFT L9-3 Experiment Simulating an Anticipated Transient Without Scram (NUREG/IA-0192)

On this page: Publication Information AbstractDownload complete document NUREG/IA-0192 (PDF 3.65 MB)Publication Information **Date Published:** January 2001 **Prepared by:** J.K. Suh, Y.S. Bang, H.J.Kim Korea Institute of Nuclear Safety 19 Kusung-Dong, Yusong-Gu Taejeon, KOREA, 305-338 **Prepared as part of:** The Agreement on Research Participation and Technical Exchange under the International Code Application and Maintenance Program (CAMP) Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Availability Notice Abstract The present work is to assess the capability of RELAP5/MOD3.2.2gamma to predict the system response following an Anticipated Transient Without Scram (ATWS) event. The experiment L9-3 which is a unique nuclear experiment simulating an ATWS event induced by loss of feedwater accident in Loss-of-Fluid-Test (LOFT) is calculated. The experimental condition and sequence are reviewed and a calculation modeling is developed with the important test specific features. The result of RELAP5 calculation is compared with the experimental data, and the predictability of the system response of the primary coolant system (PCS), the reactor power, and the steam generator (SG) secondary system is discussed. The base case showed a good agreement for the RCS pressure, temperature and reactor power with the experimental data. Therefore, it is shown that the RCS thermal-hydraulic response, the reactor power response, and the secondary system response following the LOFT L9-3 experiment can be reasonably predicted by the RELAP5 code under the current modeling scheme, and thus, that the code can be reasonably applied to the analysis of the system thermal-hydraulic response during the ATWS event in real plant. In addition, four parameters such as subcooled discharge coefficient of PORV, loss coefficient of spray valve, steam generator nodalization and moderator density coefficient (MDC) were selected and the effect of those parameters on the total discharged energy through the pressurizer safety relief valves is evaluated. 
      
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