Continuum Damage Mechanics Modeling of High-Temperature Flaw Propagation: Application to Creep Crack Growth in 316H Standardized Specimens and Nuclear Reactor Components

Abstract

Predicting creep crack growth (CCG) of flaws found during operation in high-temperature alloy components is essential for assessing the remaining lifetime of those components. While defect assessment procedures are available for this purpose in design codes, these are limited in their range of applicability. This study assesses the application of a local damage-based finite-element methodology as a more general technique for the prediction of CCG at high temperatures on a variety of structural configurations. Numerical results for stainless steel 316H, which are validated against experimental data, show the promise of this approach. This integration of continuum damage mechanics (CDM) based methodologies, together with adequate inelastic models, into assessment procedures can therefore inform the characterization of CCG under complex operating conditions, while avoiding excessive conservatism. This article shows that such modeling frameworks can be calibrated to experimental data and used to demonstrate that the degree of tri-axiality ahead of a growing creep crack affects its rate of growth. The framework is also successfully employed in characterizing CCG in a realistic reactor pressure vessel geometry under an arbitrary loading condition. These results are particularly relevant to the nuclear power industry for defect assessment and inspections as part of codified practices of structural components with flaws in high-temperature reactors.

Coupling Thermal-Hydraulics with Reactor Pressure Vessel Fracture Models

“…For instance, the effects of thermal striping could be important in high-temperature reactors cooled by molten salt. Models have been developed for crack growth in alloys subjected to these high temperatures [ 20 ]. With further development, this type of solution transfer method could be used to evaluate the potential for fatigue crack growth under high-temperature conditions in advanced reactors.…”