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Predictive reactor pressure vessel steel irradiation embrittlement models: Issues and opportunities

  • Materials Issues in Nuclear Reactors / Overview
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Abstract

Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature (ΔT) of reactor pressure vessel steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated ΔT correlation models that provide excellent statistical fits to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low flux-high fluence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of “late blooming phases” in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict ΔT attenuation away from the reactor core, verification of the master curve method to directly measure the fracture toughness with small specimens and predicting ΔT for vessel annealing remediation and re-irradiation cycles.

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Authors and Affiliations

  1. Department of Mechanical Engineering, the University of California, Santa Barbara, CA, 93106, USA

    G. R. Odette

  2. Oak Ridge National Laboratory, Oak Ridge, TN, USA

    R. K. Nanstad

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  1. G. R. Odette
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  2. R. K. Nanstad
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Correspondence to G. R. Odette.

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Odette, G.R., Nanstad, R.K. Predictive reactor pressure vessel steel irradiation embrittlement models: Issues and opportunities. JOM 61, 17–23 (2009). https://doi.org/10.1007/s11837-009-0097-4

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  • DOI: https://doi.org/10.1007/s11837-009-0097-4

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