Type CF8M cast duplex stainless steels used in the primary loop elbows of PWRs can be affected by thermal aging embrittlement at service temperature, which is around 300°C. This mechanism, resulting from the microstructural evolution of the ferrite, can reduce their fracture toughness properties. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects, such as shrinkage cavities. In a context of life extension, it is important to assess the margins for crack initiation and crack propagation instability. This paper reports the present integrity and life expectancy assessment methodologies as carried out by EDF. The French approach is based on engineering code RSE-M guidelines and French regulation requirements on NPPs in operation. This work is supported by an extensive R&D program on one hand and on field experience analysis on the other hand. It is shown how R&D and engineering tools complement each other. This paper details the three main topics of the life assessment methodology: - Fracture toughness estimates are made with predictive formulae based on chemical composition and aging conditions. These formulae are supported by a data base which is regularly up-dated with new measurements. - An inventory of manufacturing quality is drawn up and the most severe defects are characterized regarding mechanical concerns. - Fracture mechanics analyses are performed using both engineering simplified methods and classical finite element analysis. The calculated J parameter is then compared with the estimated fracture toughness of the material. Margin coefficients are included in the calculation process as required by the French regulation and code. Finally, this evaluation enables the utility to: - Identify the sensitive parts regarding the aging of cast components, - Perform a monitoring and a maintenance program as necessary.
The internal core baffle structure of a PWR consists in baffles and formers attached to the barrel. Each baffle being independent, the connection between the core baffle sheets, the formers and the core barrel is done thanks to a large number of bolts (about 1500). After inspection, some baffle bolts have been found cracked. This behaviour is attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). In order to compute accurately the temperature distribution affecting these bolts, EDF has set up a research program. Due to symmetry reasons, only a 45° sector has been accounted for. The three-dimensionnal neutron flux and the gamma induced internal heating are calculated with a Monte-Carlo particle transport code named Tripoli-4. The by-pass flow inside the cavities is computed with the CFD code Code_Saturne with a finite volume technique. Finally, the temperature distribution inside the structure (including all bolts which leads to a considerable solid mesh size — about 236 millions tetraedra) is computed by the thermal code Syrthes using a finite element approach, taking into account both the heating due to the gamma heating deposit and the cooling by the by-pass flow. Calculations show that the solid thermal field obtained exhibit strong temperature gradients and high temperature levels but in very limited zones located inside the material. As expected mainly very limited regions located inside the material and near the corner close to the reactor center are exposed to high temperature levels. On the other hand, calculations clearly confirm that external bolts thightening the core barrel and the formers see temperature much lower than those thightening the baffles.
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