Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

Abstract

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.

Simulation of low cycle fatigue stress‐strain response in 316LN stainless steel using non‐linear isotropic kinematic hardening model—A comparison of different approaches

“…Such an attempt has been made in the study by Avanzini where in the modification of NLIKH model has been proposed and implemented for ultra‐high molecular weight polyethylene by considering the changes in yield stress, Young's modulus, and plastic hardening modulus (C) as a function of cycle number. The cyclic evolution of Chaboche kinematic hardening parameters has been estimated for 316 SS‐316SS welds and its importance emphasized for elastic‐plastic analysis of reactor components for fatigue evaluation and ratcheting . However, the same has not been attempted in the present study, as it is being planned to investigate the variation in these parameters with temperature and cyclic strain amplitude in addition to the cycle number.…”

“…The cyclic evolution of Chaboche kinematic hardening parameters has been estimated for 316 SS-316SS welds and its importance emphasized for elasticplastic analysis of reactor components for fatigue evaluation and ratcheting. 25,26 However, the same has not been attempted in the present study, as it is being planned to investigate the variation in these parameters with temperature and cyclic strain amplitude in addition to the cycle number.…”

Chaboche-based cyclic material hardening models for 316 SS–316 SS weld under in-air and pressurized water reactor water conditions

“…Based on Von-Mises stress criteria and original Chaboche material hardening model, the present work proposes time-dependent linear and nonlinear material hardening models. The details of the material models and parameter estimation technique can be found elsewhere [15,16]. However, the important expressions related to the material models are presented below.…”

“…For that reason, the effect of fatigue cycles on underlying elastic-plastic material parameters and their sensitivity to different test conditions and environments was determined. An earlier paper presented a cyclic plasticity material model for 316 SS pure base metal [15] under different test and environmental conditions. This paper presents tensile and fatigue test results and associated material model parameters under different test and environmental conditions for 316 SS-316 SS welds.…”

Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

“…Thus, the development of advanced models that can address the aforementioned phenomena and the successful incorporation of those models into a generalized finite element code are necessary to ensure more accurate evaluation of the mechanistic-based structural integrity of reactor and other safety-critical components. In previous work [15][16][17][18][19], an evolutionary cyclic plasticity model was presented for key reactor materials, such as 316 SS base, 508 low alloy steel base, and 316 SS-316 SS weld. It is assumed that the material yield surface and the corresponding hardening and softening behavior evolve over time.…”