High Temperature Creep-Fatigue Evaluation of a Metallic Component

Abstract High temperature nuclear reactors may be exposed to increased cyclic thermal and mechanical loading in the metallic thermal creep regime at relatively low pressures as compared to conventional base load nuclear power plants. Recent advancements in ASME BPVC Nonmandatory Appendix HBB-Z (Guidance on Constitutive Material Models for Design by Inelastic Analysis) have been proposed or codified in the ASME Boiler and Pressure Vessel Code (BPVC) Section III, Division 5 Subsection HBB for high temperature nuclear reactor components. The purpose of this paper is to quantify and assess conservatisms associated with use of a constitutive material model for inelastic creep-fatigue evaluations which utilizes decoupled time fraction creep damage methods. This paper utilizes existing material models benchmarked to physical failure data combined with a quasi-prototypic component geometry. Creep-fatigue evaluations are performed for design methods developed in ASME Section III Division 5 Subsection HBB Appendix T using inelastic methods and applied to representative high temperature reactor component conditions. The results are intended to be used to assess if modifications to the Code methodologies should be considered for specific conditions where conservatisms appear to become relatively large and to quantify the relative magnitudes of various creep and fatigue damage factor effects applied in the existing ASME Section III Division 5 Subsection HBB Appendix T inelastic creep-fatigue methodology.