Effects of Prior Stress Relaxation on the Prediction of Creep Life Using Time and Strain Based Methods

2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Warwick M. Payten ◽  
Ken U. Snowden ◽  
David W. Dean
Keyword(s):  
Stress Relaxation ◽  
Nuclear Power ◽  
Inelastic Strain ◽  
Life Time ◽  
Fatigue Loading ◽  
Creep Life ◽  
Strain Behavior ◽  
Power Stations ◽  
Creep Fatigue ◽  
Inelastic Design

A critical requirement for both next generation conventional and nuclear plants is the development of simplified inelastic design and fitness for purposes procedures that give a reasonably accurate prediction of the complex multiaxial time dependent stress strain behavior. The accumulation of this inelastic strain in the form of coupled creep-fatigue damage over time is one of the principal damage mechanisms which will eventually lead to crack initiation in critical, high temperature equipment. Two main procedures that address creep-fatigue loading are generally used, either a time fraction or a ductility exhaustion approach. It is generally accepted that these methods enable conservative predictions within a factor of 2 to 3 and hence are reliable methods for code based design and fitness for purpose type assessments. However, for complex cycles, this may not be the case; for example, prior relaxation cycles are found to accelerate the creep rupture of the material with the result that a significant reduction in creep life can be observed. An investigation was undertaken into the influence of prior relaxation on resultant failure using a typical low alloy ferritic power station steel. Both time-based and strain based methods were used to predict the damage caused by the stress relaxation cycles followed by operation at steady state. The predictions found that while ductility exhaustion methodologies based on mean properties were adequate in predicting the failure life, time fraction methods were found to be extremely nonconservative for mean properties and only lower bound solutions provided an estimate of remaining creep life. The ASME time fraction approach, using isochronous curves was found to be extremely conservative for K = 0.67, but was able to predict similar damages to ductility exhaustion when K = 1 was used. The Monkman-Grant approach resulted in predictions that erred on the conservative side. The results have implications for both current and future conventional and nuclear power stations as it may be difficult for time based approaches to account accurately for complex cycling, shakedown conditions or stress relaxation at welds.

Effects of Prior Stress Relaxation on the Prediction of Creep Life Using Time and Strain Based Methods

2010 ◽  
Author(s):  
Warwick M. Payten ◽  
Ken U. Snowden ◽  
David W. Dean ◽  
Samuel R. Humphries ◽  
Lyndon Edwards
Keyword(s):  
Stress Relaxation ◽  
Nuclear Power ◽  
Inelastic Strain ◽  
Life Time ◽  
Fatigue Loading ◽  
Creep Life ◽  
Strain Behavior ◽  
Power Stations ◽  
Creep Fatigue ◽  
Inelastic Design

A critical requirement for both next generation conventional and nuclear plants is the development of simplified inelastic design and fitness for purposes procedures that give a reasonably accurate prediction of the complex multi-axial time dependent stress strain behavior. The accumulation of this inelastic strain in the form of coupled creep-fatigue damage over time is one of the principal damage mechanisms which will eventually lead to crack initiation in critical high temperature equipment. Two main procedures that address creep-fatigue loading are generally used, either a time fraction or a ductility exhaustion approach. It is generally accepted that these methods enable conservative predictions within a factor of 2 to 3 and hence are reliable methods for code based design and fitness for purpose type assessments. However, for complex cycles, this may not be the case, for example prior relaxation cycles are found to accelerate the creep rupture of the material with the result that a significant reduction in creep life can be observed. An investigation was undertaken into the influence of prior relaxation on resultant failure using a typical low alloy ferritic power station steel. Both time based and strain based methods were used to predict the damage caused by the stress relaxation cycles followed by operation at steady state. The predictions found that while ductility exhaustion methodologies based on mean properties where adequate in predicting the failure life, time fraction methods were found to be extremely non-conservative for mean properties and only lower bound solutions provided an estimate of remaining creep life. The Monkman-Grant approach resulted in predictions that erred on the conservative side. The results have implications for both current and future conventional and nuclear power stations as it may be difficult for time based approaches to accurately account for complex cycling, shakedown conditions or stress relaxation at welds.

Sophisticated Creep-Fatigue Life Estimation Scheme for Pressure Vessel Components Based on Stress Redistribution Locus Concept

2004 ◽  
Author(s):  
Takashi Shimakawa ◽  
Kyotada Nakamura ◽  
Ken-Ichi Kobayashi
Keyword(s):  
Thermal Stress ◽  
Pressure Vessel ◽  
Inelastic Strain ◽  
Stress Redistribution ◽  
Inelastic Behavior ◽  
Strain Concentration ◽  
Strain Behavior ◽  
Creep Fatigue ◽  
Current Design ◽  
Estimation Scheme

High temperature components are operated under cyclic thermal transient. Creep-Fatigue is the most dominant failure mode to be considered in Elevated Temperature Design of these components. Design limit for computed thermal stress is allowed to exceed yielding, because thermal stress is generally regarded as a displacement controlled one. Since creep deformation is considered as additional inelastic behavior, methodology to estimate inelastic strain concentration should be prepared in a design standard. Though inelastic FEM analyses can be applied to calculate inelastic strain concentration magnitude, it is well known that prediction is affected by applied constitutive model. Current design codes recommend to apply elastic FEM and to estimate inelastic strain behavior by simplified method. This paper presents sophisticated technique to estimate inelastic strain behavior based on Stress Redistribution Locus (SRL) method. Applicability of SRL concept is discussed with a help of FEM results for representative components of pressure vessel components such as nozzle, skirt and tube sheet.

Accuracy and Margin in Creep-Fatigue Evaluation for Modified 9Cr-1Mo Steel

2007 ◽  
Author(s):  
Yukio Takahashi
Keyword(s):  
Stress Relaxation ◽  
Nuclear Power ◽  
Creep Damage ◽  
Assessment Method ◽  
Fatigue Tests ◽  
Simple Method ◽  
Component Design ◽  
Creep Fatigue ◽  
Grade 91 ◽  
Fatigue Evaluation

Modified 9Cr-1Mo steel (ASME SA-213, Grade 91) is regarded as a promising candidate for structural materials in some of the nuclear power generation plants considered in Generation-IV project. If it is used at high temperature conditions, consideration of creep-fatigue interaction in addition to simple creep rupture is needed in component design. The author has been conducting many creep-fatigue tests for the steel at temperatures between 550°C and 650°C in order to search for a suitable creep-fatigue assessment method. It was found that creep damage at failure estimated by applying the time fraction approach to measured stress relaxation data strongly depended on the test temperature and became quite small at 550°C. However, application of calculated stress relaxation brought about the increase of creep damage over the linear damage summation line. Furthermore, addition of design factors significantly increased the values of creep and fatigue damages, making margin against failure quite large. A new definition of creep damage as a ductility consumer in strain based approach gave a simple method to estimate creep damage more properly and stably with a much smaller sensitivity on the stress relaxation behavior.

scholarly journals Effect of cyclic hardening on stress relaxation in SUS316HTP under creep-fatigue loading at 700ºC: experiments and simulations

2018 ◽  
pp. 497 ◽  
Author(s):  
Nobutada Ohno ◽  
Tatsuya Sasaki ◽  
Takehiro Shimata ◽  
Kenji Tokuda ◽  
Kimiaki Yoshida ◽  
...  
Keyword(s):  
Stress Relaxation ◽  
Cyclic Hardening ◽  
Fatigue Loading ◽  
Creep Fatigue

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

2004 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Steam Generators ◽  
Life Management ◽  
Power Stations ◽  
Reactor Pressure

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.

Thermo-Hydraulic Analysis in a Assembly Representative of a Feedwater Nozzle

2003 ◽  
Author(s):  
Xavier Boissiere ◽  
Christian Laine ◽  
Louis Doubliez
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Loading ◽  
Thermal Stresses ◽  
Experimental Testing ◽  
Flow Behavior ◽  
Life Time ◽  
Technical Problems ◽  
Power Stations ◽  
The Impact

One of the ways to reduce costs in modern nuclear power stations is to increase the life time of the steam generator. Much research is being carried out to gain a better understanding of the mechanical and thermal loads, which are generally overestimated in the interest of safety. One of the main technical problems is the feedwater of the reactor at a relatively cold (40 °C) temperature into the hot steam generator (270 °C). The connection zone, i.e. where the cold pipe is connected to the steam generator, is protected by an annulus inside the feedwater nozzle to reduce the thermal stresses. We have to identify the thermo-hydraulic behavior in that zone in order to accurately assess thermal information. This will give reliable boundary conditions for thermo-mechanical calculations. The configuration of the flows is a superposition of well known elementary fluid mechanic problems which interact strongly. No databases are available for this typical configuration, so we need to qualify the flow before using CFD modelisations. Therefore, a specific experimental testing bench was developed. In this paper, we focus our attention on the understanding of the flow in the one eyed cavity. PIV measurements allows us to identify the flow behavior. Our hypothesizes are demonstrated by a frequencies analysis, and finally confirmed with a numerical model. We also present and discuss the impact on a low thermal loading.

An evaluation of the equipment tagging process in nuclear power stations

1983 ◽  
Author(s):  
Peter Doyle ◽  
Lothar Schroeder ◽  
Stephen Brewer
Keyword(s):  
Nuclear Power ◽  
Power Stations

Environmental Protection Meets Security of Electricity Supply

2021 ◽  
pp. 1-18
Author(s):  
Ilina Cenevska
Keyword(s):  
Environmental Protection ◽  
Nuclear Power ◽  
Habitats Directive ◽  
Electricity Supply ◽  
The European Union ◽  
Power Stations ◽  
Court Of Justice ◽  
National Energy Policy ◽  
The Relationship ◽  
Practical Implications

Abstract This case comment explores the relationship between two intertwined objectives – ensuring security of electricity supply and environmental protection – in the context of the judgment of the Court of Justice of the European Union in Inter-Environnement Wallonie ASBL and Bond Beter Leefmilieu Vlaanderen ASBL v. Conseil des ministres. The analysis focuses on the application of the Environmental Impact Assessment Directive and the Habitats Directive to the facts of the case, which concerns the extension by a ten-year period of the operation of two Belgian nuclear power stations (Doel 1 and Doel 2) as part of a national energy policy strategy to ensure the security of Belgium's electricity supply. The case comment also considers the legal and practical implications that arise as a result of employing the ‘security of electricity supply’ exemption to enable derogation from the requirements of the aforementioned Directives in circumstances where a Member State considers the security of its electricity supply to be under threat.

Evaluation of the residual stress in stainless steel 304L hydraulically expanded joints

2020 ◽  
pp. 095440892096401
Author(s):  
Ying Hong ◽  
Xuesheng Wang ◽  
Yan Wang ◽  
Zhao Zhang ◽  
Yong Han
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Yield Strength ◽  
Residual Stresses ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
The Finite Element Method ◽  
Power Stations ◽  
Stainless Steel 304 ◽  
Initial Clearance

Stainless steel 304 L tubes are commonly used in the fabrication of heat exchangers for nuclear power stations. The stress corrosion cracking (SCC) of 304 L tubes in hydraulically expanded tube-to-tubesheet joints is the main reason for the failure of heat exchangers. In this study, 304 L hydraulically expanded joint specimens were prepared and the residual stresses of a tube were evaluated with both an experimental method and the finite element method (FEM). The residual stresses in the outer and inner surfaces of the tube were measured by strain gauges. The expanding and unloading processes of the tube-to-tubesheet joints were simulated by the FEM. Furthermore, an SCC test was carried out to verify the results of the experimental measurement and the FEM. There was good agreement between the FEM and the experimental results. The distribution of the residual stress of the tube in the expanded joint was revealed by the FEM. The effects of the expansion pressure, initial tube-to-hole clearance, and yield strength of the tube on the residual stress in the transition zone that lay between the expanded and unexpanded region of the tube were investigated. The results showed that the residual stress of the expanded joint reached the maximum value when the initial clearance was eliminated. The residual stress level decreased with the decrease of the initial tube-to-hole clearance and yield strength. Finally, an effective method that would reduce the residual stress without losing tightness was proposed.

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