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

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Bipul Barua ◽  
Subhasish Mohanty ◽  
Joseph T. Listwan ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Fatigue Loading ◽  
Cyclic Stress ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Evaluation

Although S∼N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, in this work, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

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

2015 ◽  
Author(s):  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Nuclear Reactor ◽  
Axial Stress ◽  
Cyclic Plasticity ◽  
Model Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Test Conditions ◽  
Over Time

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.

Finite Element Cracking Analysis of Nuclear Containment Vessels Under External Impacts

2011 ◽  
Author(s):  
Y. F. Al-Obaid
Keyword(s):  
Finite Element ◽  
Three Dimensional ◽  
Structural Damping ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Pressurized Water ◽  
Nuclear Containment ◽  
Linear Behavior ◽  
Line Elements ◽  
Aircraft Crash

This paper attempts to examine the behavior of nuclear containment vessels under impact loads. Three-dimensional dynamics finite element analysis is proposed. The analysis includes the non-linear behavior of concrete, structural damping and cracking. A combination of solid isoparametric, panel and line elements representing vessel concrete, steel lining, and prestressing tendons or conventional steel, respectively, is suggested. Three-dimensional computer program OBAID is developed which gives time-dependent interactive calculations for stresses, deflections, cracks, reinforcement and suitable wall and dome thicknesses. An existing concrete containment vessels for pressurized water reactor (PWR) is examined under aircraft crash load.

Investigative Study of 2-D vs. 3-D Weld Residual Stress Analyses of the NRC Phase II Mockup

2012 ◽  
Author(s):  
Francis H. Ku ◽  
Shu (Stan) S. Tang
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Phase Ii ◽  
Three Dimensional ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Approximation Techniques ◽  
Weld Residual Stress ◽  
Axisymmetric Modeling ◽  
Stress Analyses

Finite element weld residual stress analyses are performed to investigate the similarities and differences between two-dimensional (2-D) and three-dimensional (3-D) finite element analyses on weld residual stress predictions of the NRC Phase II Mockup. The Mockup resembles a typical pressurized water reactor (PWR) surge nozzle of 14″ in diameter which includes a dissimilar metal weld (DMW) connecting the safe end and a stainless steel weld (SSW) connecting the surge line piping. The 2-D analysis employs axisymmetric modeling approach, while the 3-D analysis utilizes moving heat source approximation techniques. The results demonstrate the variations in residual stresses among the weld bead start and stop locations. Comparing the 2-D and 3-D residual results against experiment measurements also reveal the limitations inherent to the 2-D analysis, while the 3-D analysis can produce results that are of closer match to experimental measurements.

Development of Advanced Fatigue Evaluation Methodology for Monitoring Major Components in Nuclear Power Plant

2010 ◽  
Author(s):  
W. Kim ◽  
Jongjooh Kwon ◽  
Hong Tae Kang ◽  
Gyeong-Hoi Koo ◽  
Tae-Ryong Kim
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Nuclear Power ◽  
Evaluation Methodology ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Monitoring ◽  
Improve Accuracy ◽  
Fatigue Evaluation

In an attempt to develop fatigue monitoring system, two improved fatigue evaluation schemes have been proposed to monitor fatigue degradation in major components and piping of the pressurized water reactor. Proposed methods are both aimed to obtain realistic fatigue usage factors for given plant transients. Developed schemes utilize plant operating signals such as coolant temperature, pressure and flow rate. Finite element method and an improved Green’s function approach were used to calculate stresses and fatigue usage. Case studies were performed to validate effectiveness of each proposed scheme. It has been confirmed that proposed schemes can effectively reduce excessive conservatism in estimating fatigue usage and improve accuracy in stress calculation.

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