Quantified PIRT for Reactor System Safety Analysis Code Validation Using Characteristic Time Ratios
US Nuclear Regulatory Commission (NRC) has revised the reactor licensing rules in 1988 to allow the use of realistic best-estimate (BE) computer code if its uncertainties are quantified following the rigorous Code Scaling, Application and Uncertainty (CSAU) methodology. To identify major sources of uncertainty in best-estimate code calculations, CSAU methodology introduced a systematic approach called Phenomena Identification and Ranking Table (PIRT), which is based on expert panel subjective conceptual judgments that may be overly conservative. To support and examine the traditional PIRT with objective quantified judgments, this study proposes a quantitative approach, the Quantified PIRT (QPIRT), together with an example application to a 2″ Loss-of-Coolant Accident (LOCA) in APEX test facility at Oregon State University (OSU). Through dimensionless analysis to code models and closure relations, phasic and mixture fluid dimensionless groups (Π groups) are obtained for specific transient. The numerical values of Π groups are calculated using best-estimate nuclear reactor safety codes results. Since Π groups of greater values have higher impact on the transient, the ranking of the numerical Π groups can be employed as the QPIRT for the specific transient. The QPIRT, together with the traditional PIRT, identifies the dominating physical phenomenon during different phases of the transient and can be adopted for further code sensitivity and uncertainty analysis.
PIRT Computer Code Uncertainty Quantification Dimensionless Analysis
Hu Luo Qiao Wu Vincent A. Mousseau
Oregon State University 100 Radiation Center, Corvallis, OR 97331-5902, USA Idaho National Laboratory P.O. Box 1625, MS 3840, Idaho Falls, ID 83415-3840, USA
国际会议
上海
英文
425-437
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）