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French RPV Assessment — Contribution of Expertises in Mechanical Analyses

[+] Author Affiliations
G. Bezdikian, Y. Rouillon, J. Bourgoin

Electricité De France, Saint-Denis Cedex, France

Paper No. PVP2002-1351, pp. 59-71; 13 pages
doi:10.1115/PVP2002-1351
From:
  • ASME 2002 Pressure Vessels and Piping Conference
  • Fatigue, Fracture and Damage Analysis, Volume 2
  • Vancouver, BC, Canada, August 5–9, 2002
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4654-7
  • Copyright © 2002 by ASME

abstract

The process used by the French utility, concerning the Reactor Pressure Vessel assessment, applied on 54 PWR NPPs (3-loop and 4-loop Reactors), involves the verification of the integrity of the component by mechanical studies, in the most severe conditions of loading in relation with RTndt (Reference Nil Ductility Transition Temperature), and considering major parameters. This approach, is based on mechanical safety studies, to demonstrate the absence of risk of failure by brittle fracture. For these mechanical studies two major input data are necessary: 1 - the fluence distribution and the values during the lifetime in operation for each NPPs, 2 - the thermal-hydraulic evaluation and temperature distribution values in the downcomer. The main results must show significant margins against initiation of the brittle fracture. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • more precise assessment of the fluence and neutronic calculations, • better knowledge of the vessel material properties, including the effect of radiation, • the NDE inspection program based on the inspection of the vessel wall, with a special NDE tool to inspect the area in subcladding zone, • the evaluation of vessel integrity, the mechanical analysis of margins in major loading conditions. The principal actions conducted during recent years are: • the fuel management optimisation (low-leakage core design) and the new development to evaluate the fluence, • the data gathered from radiation specimen capsules, removed from the vessels (3 loop reactor), within the framework of the radiation surveillance program, and • the thermal-hydraulic-mechanical calculations based on finite element thermal-hydraulic computations and three dimensional elastic-plastic mechanical analyses.

Copyright © 2002 by ASME

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