Thermal stress intensity factor solutions for reactor pressure vessel nozzles
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To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it isregulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtainedby the internal pressure and thermal load should be obtained through crack analysis of the nozzle cornercrack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G,the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling orheating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomialformat based on the stress linearization analysis results. Inevitably, the SIF can only be obtained throughfinite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal loadunder, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant,three geometric variables were set and 72 cases of RPV models were made, and then the heat transferanalysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions aresuggested
Abstract
To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it isregulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtainedby the internal pressure and thermal load should be obtained through crack analysis of the nozzle cornercrack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G,the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling orheating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomialformat based on the stress linearization analysis results. Inevitably, the SIF can only be obtained throughfinite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal loadunder, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant,three geometric variables were set and 72 cases of RPV models were made, and then the heat transferanalysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions aresuggested
참고문헌 (14)
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