The Reactor Pressure Vessel (RPV) has long been considered one of the most reliable components in Pressurized Water Reactor (PWR). Nowadays, a general target for the countries that produce nuclear energy is to extend the operation life of existing plants. From this point of view, the RPV is one of the major components that may limit the useful life of the nuclear plant. The risk for the RPV structural integrity is connected to the presence of a flaw of sufficient size, a high level of embrittlement due to radiation damage, and the occurrence of a thermal-hydraulic transient inducing strong stresses in the vessel wall. Severe loading conditions are produced during a Pressurized Thermal Shock (PTS) event, in which an overcooling may induce strong thermal stresses while the internal pressure can be maintained at high level or the system can be re-pressurized during the transient. Such conditions are generated in a Loss Of Coolant Accident (LOCA) transient during the emergency injection. In recent years, important progresses have been made in the development of analysis methods and tools for the best estimation of the thermal and pressure loads on the vessel wall. In this direction, the US-NRC published in 2007 a document aimed at reviewing the rules adopted in PTS analysis, established in the 1980s, containing significant conservatisms, for a Best Estimate (BE) approach combined with uncertainty assessment. In this paper, the methodology for Fracture Mechanics analysis developed at University of Pisa aimed to perform parametric analysis assuming various shapes and locations of the flaw is applied to a Pressurized Heavy Water Reactors (PHWR) during a LOCA scenario. Four steps can be identified starting from the thermal hydraulic analysis of the Nuclear Power Plant (NPP) behaviour with Relap5-3D© in order to calculate the cooling loads of the Emergency Core Coolant Systems (ECCS). The second step is the analysis by mean a CFD code (CFX) of the mixing phenomena occurring in the Down-Comer (DC) and the calculation of the thermal load on the RPV internal surface. The third step is represented by the evaluation of the stresses inside the RPV wall by mean a Finite Element (FE) code (Ansys) under the thermal and pressure loads calculate in the previous steps. The last step is represented by the calculation of the Stress Intensity Factor (SIF) KI by mean the Weight Function method and the comparison with the critical SIF KIc of the material, once the stresses inside the undamaged RPV wall are known. The goal of this work is the evaluation of the safety margin for the operation of the RPV, adopting a BE approach in all the steps of the analysis. This result will be compared with the one obtained with the application of the ASME XI criteria for the KI evaluation with the aim to show that the BE approach leads to a larger safety margin.

Fracture Mechanics Analysis in a Pressurized Heavy Water Reactor Vessel during LOCA Scenario

D'AURIA, FRANCESCO SAVERIO
2010-01-01

Abstract

The Reactor Pressure Vessel (RPV) has long been considered one of the most reliable components in Pressurized Water Reactor (PWR). Nowadays, a general target for the countries that produce nuclear energy is to extend the operation life of existing plants. From this point of view, the RPV is one of the major components that may limit the useful life of the nuclear plant. The risk for the RPV structural integrity is connected to the presence of a flaw of sufficient size, a high level of embrittlement due to radiation damage, and the occurrence of a thermal-hydraulic transient inducing strong stresses in the vessel wall. Severe loading conditions are produced during a Pressurized Thermal Shock (PTS) event, in which an overcooling may induce strong thermal stresses while the internal pressure can be maintained at high level or the system can be re-pressurized during the transient. Such conditions are generated in a Loss Of Coolant Accident (LOCA) transient during the emergency injection. In recent years, important progresses have been made in the development of analysis methods and tools for the best estimation of the thermal and pressure loads on the vessel wall. In this direction, the US-NRC published in 2007 a document aimed at reviewing the rules adopted in PTS analysis, established in the 1980s, containing significant conservatisms, for a Best Estimate (BE) approach combined with uncertainty assessment. In this paper, the methodology for Fracture Mechanics analysis developed at University of Pisa aimed to perform parametric analysis assuming various shapes and locations of the flaw is applied to a Pressurized Heavy Water Reactors (PHWR) during a LOCA scenario. Four steps can be identified starting from the thermal hydraulic analysis of the Nuclear Power Plant (NPP) behaviour with Relap5-3D© in order to calculate the cooling loads of the Emergency Core Coolant Systems (ECCS). The second step is the analysis by mean a CFD code (CFX) of the mixing phenomena occurring in the Down-Comer (DC) and the calculation of the thermal load on the RPV internal surface. The third step is represented by the evaluation of the stresses inside the RPV wall by mean a Finite Element (FE) code (Ansys) under the thermal and pressure loads calculate in the previous steps. The last step is represented by the calculation of the Stress Intensity Factor (SIF) KI by mean the Weight Function method and the comparison with the critical SIF KIc of the material, once the stresses inside the undamaged RPV wall are known. The goal of this work is the evaluation of the safety margin for the operation of the RPV, adopting a BE approach in all the steps of the analysis. This result will be compared with the one obtained with the application of the ASME XI criteria for the KI evaluation with the aim to show that the BE approach leads to a larger safety margin.
2010
9789616207317
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/11568/139442
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