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At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.

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This paper discusses a material hardening models for welds made from 316 stainless steel (SS) to 316 SS. The model parameters were estimated from the strain-versus-stress curves obtained from tensile and fatigue tests conducted under different conditions (air at room temperature, air at 300 o C, and primary loop water conditions for a pressurized water reactor). These data were used to check the fatigue cycle dependency of the material hardening parameters (yield stress, parameters related to Chaboche-based linear and nonlinear kinematic hardening models, etc.). The details of the experimental results, material hardening models, and associated calculated results are published in an Argonne report (ANL/LWRS-15/2). This paper summarizes the reported material parameters for 316 SS-316 SS welds and their dependency on fatigue cycles and other test conditions. 1 Introduction At present, the fatigue life evaluation of nuclear power plant components has large uncertainties [1]. The relevant design codes [2, 3] allow elastic-analysis-based fatigue analysis of nuclear reactor components. Ideally, if stress and strain stay below the elastic limit, no fatigue would occur in the reactor components. However, safety-critical reactor components often fail due to fatigue damage associated with the reactor loading cycles and environmental conditions. In addition to fatigue damage, ratcheting of reactor components could happen due to the presence of stress concentration and/or plastic zones. The stress concentration and the plastic zone formation in the reactor metal could be due to weld residual stress formation, stress corrosion cracking, etc. Hence, for better accuracy, it is essential to estimate the fatigue and ratcheting damage of reactor components based on the results of elastic-plastic stress analysis rather than pure elastic stress analysis alone. Since ratcheting is a phenomenon closely related to the transient plastic deformation behavior, its nonlinear description requires the calculation of material hardening

Materials considered for application in the high temperature gaseous environments that are present in various coal-gasification processes must be resistant to corrosive and erosive wear. Possible interactions between the materials and the gaseous environment that lead to oxidation, hot corrosion, sulfidation, carburization, and metal dusting are discussed. Existing formalisms for predicting erosive wear loss of materials are presented. The need for additional research on corrosion of materials in complex multicomponent gas environments and on erosion of materials in the relevant range of process conditions to develop suitable models for corrosion-erosion behavior of materials is discussed.