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.

Apprrvad for public rcltoMg DUtrihution Unlimitod »// tmtmtaumlmiiim ■ >L^w*mpiipipi»i|B^ppppfB»!»Tiwpppppw^w(«iii^i ii wpw» wwpuuww" "i .1 miwrnfmrnim mi" "". > mi^lPBnmnwaiiinjiiiii^iii ■> m^i^mm^mm FOREWORD Recent requirements for increased strength and service life of machines and structures have been m>t by the use of higher strength materials and new fabrication and joining methods. Simultaneously, failures due to fracture have increased raiative to those resulting from excessive deformation. Frequently service conditions are such that low temperature brittle fracture, fatigue fracture, and high temperature creep rupture must be considered in a single system. National concern with increased safety, reliability, and cost has focused attention upon these problems. Methods are now available to predict both fatigue crack initiation life and crack propagation life. Paradoxically the materials properties required for long fatigue crark initiation life are incompatible with the requirements of high fracture toughness. Thus, the conflicting design approaches and requirements placed on the material are confusing and often impossible to satisfy. Numerous publications dealing with a variety of fracture problems have led to many new and useful developments. However, the synthesis of the concepts into methods for design, testing and inspection has lagged. This program of study is intended to contribute to the integration, correlation, and organizaticn of mechanics and materials concepts and research information into a form that will permit enlightened decisions to be made regarding fracture control. Reports are in preparation in three categories: 1. Research reports designed to explore, study and integrate isolated and/or conflicting concepts and methods dealing with life prediction, 2. Reports to introduce and summarize the state-of-the-art concepts and methods in particular areas, and 3. Example problems and solutions intended to illustrate the use of these concepts in decision making.

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

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In USA there are approximately 100 operating light water reactors (LWR) consisting fleet of both pressurized water reactors (PWR) and boiling water reactors (BWR). Most of these reactors were built before 1970 and the design lives of most of these reactors are 40 years. It is expected that by 2030, even those reactors that have received 20 year life extension license from the US nuclear regulatory commission (NRC) will begin to reach the end of their licensed periods of operation. For economical reason it is be beneficial to extend the license beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean and economic electricity without significant green house gas emissions. However, environmental fatigue is one of the major aging related issues for these reactors, and may create hurdles in long term sustainability of these reactors. To address some of the environmental fatigue related issues, Argonne National Laboratory (ANL) with the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program trying to develop mechanistic approach for more accurate life estimation of LWR components. In this context ANL conducted many fatigue experiments under different test and environment conditions on 316 stainless steel (316SS) material that is or similar grade steels are widely used in US reactors. Contrary to the conventional S∼N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to understand material ageing more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help to develop computer based advanced modeling tools to better extrapolate stress-strain evolution of reactor component under multi-axial stress states and hence to help predicting their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In another paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.