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    Characterization by Notched and Precracked Charpy Tests of the in-Service Degradation of Reactor Pressure Vessel Steel Fracture Toughness

    Published: 01 January 1998

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    The current engineering and regulatory practice to estimate fracture toughness safety margins for nuclear reactor pressure vessels (RPV's) relies heavily on the Charpy V-Notch (CVN) impact test. The uncertainties of this correlation ("indexing") approach are known. Techniques to more directly estimate in-service toughness degradation using a variety of precracked specimens are under active development worldwide, with emphasis on their miniaturization. In the nuclear context, it is essential to address many concerns, such as: representativeness of the surveillance programs with respect to the vessel in terms of materials and environment, transferability of test results to the structure (constraint and size effects), conservatism of lower bound toughness estimation procedure, credibility relative to trends of existing data bases, etc.

    An enhanced RPV surveillance strategy is being developed in Belgium. It combines state-of-the-art micromechanical and damage modeling to the evaluation of CVN load-deflection signals, tensile stress-strain curves and slow-bend tests of reconstituted, precracked Charpy specimens. A probabilistic micromechanical model has been established for static and dynamic transgranular cleavage initiation fracture toughness in the ductile-brittle transition temperature range; this model allows one to project toughness bounds for any steel embrittlement condition from the corresponding CVN and static tensile properties, using a single scaling factor defined by imposing agreement with toughness tests in a single condition. The outstanding finding incorporated by this “toughness transfer model” is that the microcleavage fracture stress is affected by temperature in the ductile-brittle transition and that this influence is strongly correlated to the flow stress: this does phenomenologically explain the shape of the KIc, KId-temperature curves as well as the actual magnitude of the strain rate and irradiation effects. Furthermore, CVN crack arrest loads and fracture appearance are also taken advantage of in order to estimate KIa degradation. Last, but not least, the CVN-tensile load-temperature diagram provides substantial information used for the modeling of in-service steel strengthening, of intergranular fracture susceptibility and other irradiation-induced and ageing effects.

    Systematic application of this enhanced surveillance approach is expected to contribute to the improvement of the engineering and regulatory predictive capability, while making optimal use of the limited inventory of available surveillance material.


    Low alloy steel, reactor pressure vessel, embrittlement, fracture oughness, microstructure, damage modeling, micromechanics, safety

    Author Information:

    Fabry, A
    Head, Fracture Mechanics and Modeling, Mol,

    Committee/Subcommittee: E10.02

    DOI: 10.1520/STP37996S