Nuclear Fatigue Analysis Codified Design Rules: Comparison of Cyclic Plasticity Effects

2015 ◽  
Author(s):  
Claude Faidy ◽  
Andrew Wasylyk
Keyword(s):  
Strain Amplitude ◽  
Task Force ◽  
Cyclic Plasticity ◽  
Elastic Plastic ◽  
Inelastic Analysis ◽  
Code Comparison ◽  
Life Evaluation ◽  
Piping Systems ◽  
Fatigue Life Evaluation ◽  
Nuclear Association

Large uncertainties exist in fatigue life evaluation using existing elastic stress analysis. The usage factor is obtained by comparison of strain amplitude evaluation on different points of the components and the fatigue S-N curve of the material. This paper will review existing codified rules in major nuclear Codes that are proposed as simplified elastic-plastic analysis of strain amplitude. The different formulas proposed by the different Codes are described with their own background document and compared on typical cases. Methods are based on simplified elastic-plastic approach (elastic analysis plus correction factor) with associated material data. The Code comparison includes nuclear Codes, as ASME Boiler and Pressure Vessel Code Section III, French RCC-M and RCC-MRx and German KTA; Russian PNAEG Code and JSME rules are also considered based on specific English version of the Codes for fatigue rules. Two types of components are concerned by the comparison: vessels and piping systems. All these Codes are proposing different Ke and Kv rules based on different assumption. Finally, a first set of recommendation to perform reference inelastic analysis will be presented to improve existing codes on an harmonized way, associated to analytical recommendations, all material properties needed and criteria to apply this modern methods. This review is mainly done in World Nuclear Association (WNA), Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group, Codes & Standards Task Force.

Seismic Qualification of Piping Systems by Detailed Inelastic Response Analysis: Part 1 — A Code Case for Piping Seismic Evaluation Based on Detailed Inelastic Response Analysis

2017 ◽  
Author(s):  
Masaki Morishita ◽  
Akihito Otani ◽  
Tomoyoshi Watakabe ◽  
Izumi Nakamura ◽  
Tadahiro Shibutani ◽  
...  
Keyword(s):  
Strain Amplitude ◽  
Seismic Design ◽  
Equivalent Strain ◽  
Design Rule ◽  
Design Evaluation ◽  
Response Analysis ◽  
Inelastic Response ◽  
Inelastic Analysis ◽  
Piping Systems ◽  
Equivalent Strain Amplitude

A Code Case in the framework of the Nuclear Codes and Standards of Japan Society of Mechanical Engineers (JSME) is currently under development to incorporate seismic design evaluation methodologies for piping systems by detailed inelastic response analysis and strain-based fatigue criteria as an alternative design rule to the current rule, in order to provide a more rational seismic design evaluation by taking directly the response reduction due to plasticity energy absorption into account. The Code Case provides two strain-based criteria; one is a limit to maximum amplitude of equivalent strain amplitude derived from detailed analysis and the other is a limit to the fatigue usage factor also based on the equivalent strain amplitude. The Code Case also provides an evaluation method by simplified inelastic analysis with an additional damping taking the response reduction due to plasticity into account. Some discussions are provided on the adequacy of additional damping in the simplified inelastic analysis and the safety margin and reliability of fatigue evaluation by the detailed inelastic response analysis provided in the Code Case.

Is it Possible to Get-Rid of S-N Curve for Fatigue Evaluation?: A Fully Mechanistic Model of 316SS Reactor Steel for Fatigue Life Evaluation

2017 ◽  
Author(s):  
Bipul Barua ◽  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Fatigue Life ◽  
3D Structure ◽  
Cyclic Plasticity ◽  
Fe Model ◽  
Plasticity Model ◽  
Modeling Framework ◽  
Stress Strain ◽  
Life Evaluation ◽  
Fatigue Life Evaluation ◽  
Cyclic Plasticity Model

The present methods for fatigue life evaluation of nuclear reactor components have large uncertainties due to the overdependence on approaches that involve empirical fatigue life estimation, such as use of test-based curves of stress/strain versus life (S∼N) and Coffin-Manson type empirical relations. To reduce the uncertainty in fatigue life evaluation, we are trying to develop a fully mechanistic modeling approach. The aim is to capture the time/cycle-dependent material ageing behavior such as stress hardening/softening through multi-axial stress-strain evolution of the components based on which the life of the component can be predicted. In this paper, we introduce an implementation of the ANL developed evolutionary cyclic plasticity model for 316 SS reactor steel within the commercial finite element (FE) software ABAQUS. A user subroutine is developed to enable the incorporation of the ANL developed evolutionary cyclic plasticity model [1] into ABAQUS. The FE model, developed in this work, can be used for predicting the time-dependent stress hardening/softening of 3D structure. A strain-controlled constant amplitude fatigue experiment scenario is 3D modeled using the developed ABAQUS based FE modeling framework and is verified through experimental data.

A Study on Fatigue Life Evaluation of Pressure Vessel Made of ASME SA240-316L Material Using Numerical Analysis

2019 ◽  
Vol 893 ◽  
pp. 1-5 ◽  
Author(s):  
Eui Soo Kim
Keyword(s):  
Finite Element ◽  
Fatigue Life ◽  
Pressure Vessel ◽  
Life Prediction ◽  
Pressure Vessels ◽  
Physical Damage ◽  
Element Analysis ◽  
Internal Damage ◽  
Life Evaluation ◽  
Fatigue Life Evaluation

Pressure vessels are subjected to repeated loads during use and charging, which can causefine physical damage even in the elastic region. If the load is repeated under stress conditions belowthe yield strength, internal damage accumulates. Fatigue life evaluation of the structure of thepressure vessel using finite element analysis (FEA) is used to evaluate the life cycle of the structuraldesign based on finite element method (FEM) technology. This technique is more advanced thanfatigue life prediction that uses relational equations. This study describes fatigue analysis to predictthe fatigue life of a pressure vessel using stress data obtained from FEA. The life prediction results areuseful for improving the component design at a very early development stage. The fatigue life of thepressure vessel is calculated for each node on the model, and cumulative damage theory is used tocalculate the fatigue life. Then, the fatigue life is calculated from this information using the FEanalysis software ADINA and the fatigue life calculation program WINLIFE.

Seismic Qualification of Piping Systems by Detailed Inelastic Response Analysis: Part 3 — Variation in Elastic-Plastic Analysis Results on Carbon Steel Pipes From the Benchmark Analyses and the Parametric Analysis

2017 ◽  
Author(s):  
Izumi Nakamura ◽  
Akihito Otani ◽  
Masaki Morishita ◽  
Masaki Shiratori ◽  
Tomoyoshi Watakabe ◽  
...  
Keyword(s):  
Carbon Steel ◽  
Evaluation Procedure ◽  
Response Analysis ◽  
Seismic Load ◽  
Inelastic Response ◽  
Elastic Plastic ◽  
Analytical Results ◽  
Plastic Analysis ◽  
Piping Systems ◽  
Parametric Analyses

It is recognized that piping systems used in nuclear power plants have a significant amount of the safety margin, up to the point of boundary failure, even when the input seismic load exceeds the allowable design level. The reason is attributed to the large strength capacity of the piping systems in the plastic region. In order to establish an evaluation procedure, in which the inelastic behavior of piping systems is considered in a rational way, a task group activity under the Japan Society of Mechanical Engineers (JSME) has been conducted. As a deliverable of this activity, a Code Case in the framework of the JSME Nuclear Codes and Standards is now being developed. The Code Case provides the strain-based criteria, an evaluation procedure using the response-spectrum based inelastic analysis, and detailed inelastic response analysis based on a finite element model. For developing the Code Case, inelastic benchmark and parametric analyses of the tests of a pipe element and piping system made of carbon steel were conducted to investigate the variation of the elastic-plastic analyses results. Based on these analytical results, it is assumed that setting the yield stress has a significant influence on the inelastic analytical results, while the work hardening modulus in the bi-linear approximation of the stress-strain curve has little influence. From the results of the parametric analyses, it is confirmed that the variation in the analytical results among the analysts would be reduced by having a unifying analysis procedure. In this paper, the results of the parametric analyses and the variation in the elastic-plastic analysis are discussed.

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