Abstract

Abstract The assessment of the structural integrity of nuclear vessels is based on a series of procedures developed in the 1970s and 1980s. On one hand, curves that, according to the American Society of Mechanical Engineers code, describe the relationship between steel toughness and temperature in the ductile-to-brittle transition region, based on the reference temperature concept RTNDT, were adopted in 1972. On the other hand, the material embrittlement derived from the exposure of steel to neutron irradiation is determined through the model included in “Regulatory Guide 1.99 Rev. 2,” published in 1988. Since then, there have been enormous advances in this field. For example, the Master Curve, based on the reference temperature T0, describes the relationship between toughness and temperature in the transition zone more realistically and with much more robust microstructural and mechanical foundations and uses the elastic-plastic fracture toughness KJc. Moreover, improved models have been developed to estimate the embrittlement of steel subjected to neutron irradiation, such as ASTM E900, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials. This study is aimed at comparing the results obtained using traditional procedures to the improved alternatives developed later. For this purpose, the behavior of the steel of a nuclear vessel that is currently under construction has been experimentally characterized through RTNDT and T0 parameters. In addition, the material embrittlement has been quantified using “Regulatory Guide 1.99 Rev. 2” and ASTM E900. These experimental results have been transferred to the assessment of the structural integrity of the vessel to determine the pressure-temperature limit curves and size of the maximum admissible defect as a function of the operation time of the plant. The results have allowed the implicit overconservatism present in the traditional procedures to be quantified.

Highlights

  • Neutron irradiation is the most relevant source of degradation for nuclear reactor pressure vessel (RPV) steels

  • The main goal of these analyses is to determine the amount of those chemical elements that enhances the irradiation embrittlement, such as copper and nickel, in order to obtain the shift in ductileto-brittle transition (DBT) according to the models introduced in the “Correlations to Predict the Material Embrittlement” section

  • According to the current legislation, the initial toughness of the steel in the DBT region is described in a deterministic way by means of the initial reference temperature RTNDT(U) and its shift over time, ΔRTNDT

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Summary

Introduction

Neutron irradiation is the most relevant source of degradation for nuclear reactor pressure vessel (RPV) steels. The fracture toughness requirements for ferritic materials of nuclear power plants (NPPs) must fulfill the acceptance and performance criteria of Appendix G of Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1,2]. According to this procedure, the fracture resistance of the vessel material in the DBT region before irradiation is described by the reference temperature RTNDT(U), which is obtained from Charpy impact and Pellini drop weight tests through empirical and conservative correlations.

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