scholarly journals Consistent Creep and Rupture Properties for Creep-Fatigue Evaluation

1979 ◽  
Vol 101 (4) ◽  
pp. 276-285 ◽  
Author(s):  
C. C. Schultz
Keyword(s):  
Fatigue Life ◽  
Factor Of Safety ◽  
304 Stainless Steel ◽  
Code Design ◽  
Class 1 ◽  
Creep Fatigue ◽  
Asme Code ◽  
Accepted Practice ◽  
Fatigue Evaluation ◽  
Rupture Properties

The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a minimum strength heat is also shown to provide reasonable predictions. The viability of using consistent properties (either actual or those of a minimum strength heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations.

Environmental Fatigue Evaluation of a Korean Nuclear Power Plant

2008 ◽  
Author(s):  
Chang-Kyun Oh ◽  
Hyun-Su Kim ◽  
Hag-Ki Youm ◽  
Tae-Eun Jin ◽  
Young-Jin Kim
Keyword(s):  
Fatigue Life ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Moment Stress ◽  
Wide Range ◽  
Class 1 ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Reactor Pressure

In accordance with the recommendation of USNRC and the U.S. license renewal experiences, the effects of reactor coolant environment on the fatigue life have to be considered for the continued operation of operating nuclear power plants as well as for the design of new plants. Although various evaluation methodologies have been suggested to date, a wide range of comparison of the existing methodologies has not been performed. The purpose of this paper is to evaluate the environmental effects on the fatigue life of a reactor pressure vessel and ASME Class 1 piping by the various methodologies and to investigate the effects of the pressure and moment stress histories on the environmental fatigue evaluation. The evaluation results show that the environmental fatigue evaluation results based on the design cumulative usage factors for the reactor pressure vessel and ASME Class 1 piping satisfy the requirement of the ASME code except for charging nozzle. However, when using operating cumulative usage factor, the environmental fatigue evaluation result for the charging nozzle satisfies the ASME code allowable. And the effects of the pressure and moment stress histories on the environmental fatigue evaluation are considered to be small when using the modified rate approach.

Application of Draft Regulatory Guide DG-1144 Guidelines for Environmental Fatigue Evaluation to a BWR Feedwater Piping System

2007 ◽  
Author(s):  
Hardayal S. Mehta ◽  
Henry H. Hwang
Keyword(s):  
Fatigue Life ◽  
Thermal Cycle ◽  
Light Water Reactor ◽  
Piping System ◽  
Low Alloy Steel ◽  
Light Water ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Cycle Diagram ◽  
Analytical Expressions

Recently published Draft Regulatory Guide DG-1144 by the NRC provides guidance for use in determining the acceptable fatigue life of ASME pressure boundary components, with consideration of the light water reactor (LWR) environment. The analytical expressions and further details are provided in NUREG/CR-6909. In this paper, the environmental fatigue rules are applied to a BWR feedwater line. The piping material is carbon steel (SA333, Gr. 6) and the feedwater nozzle material is low alloy steel (SA508 Class 2). The transients used in the evaluation are based on the thermal cycle diagram of the piping. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. In both cases the cumulative fatigue usage factors are shown to be less than 1.0.

Creep-Fatigue Life Assessment Under Multiaxial Strain State

2006 ◽  
Author(s):  
Shengde Zhang ◽  
Masao Sakane ◽  
Takamoto Itoh
Keyword(s):  
Fatigue Life ◽  
Biaxial Tension ◽  
Creep Damage ◽  
Strain Ratio ◽  
Life Assessment ◽  
Principal Strain ◽  
304 Stainless Steel ◽  
Strain Wave ◽  
Element Analysis ◽  
Creep Fatigue

This paper studies the multiaxial creep-fatigue life for type 304 stainless steel at elevated temperature. Strain controlled biaxial tension-compression creep-fatigue tests were carried out using cruciform specimens under four strain waves at three principal strain ratios. The strain wave and the principal strain ratio had a significant effect on creep-fatigue life of the cruciform specimen. The creep-fatigue life ratio decreased as the principal strain ratio increased which indicates that larger creep damage occurred at larger principal strain ratio. The effects of the strain wave and principal strain ratio were discussed in relation to the observations of surface crack and void area density in the gage part of the specimen. Creep-fatigue lives were discussed in relation to the principal stress amplitude calculated by finite element analysis and creep-fatigue damage was evaluated by linear damage rule.

Study of the Effects of Environment in the Fatigue Analysis on Existing LWR as Proposed in USNRC RG 1.207

2009 ◽  
Author(s):  
Eugene Tom ◽  
Milton Dong ◽  
Hong Ming Lee
Keyword(s):  
Fatigue Life ◽  
Correction Factor ◽  
Environmental Effects ◽  
Environmental Parameters ◽  
Computer Code ◽  
Cyclic Behavior ◽  
Simplified Approach ◽  
Asme Code ◽  
Detailed Computer ◽  
Fatigue Evaluation

US NRC Regulatory Guide 1.207 Rev. 0 provides guidance for use in determining the acceptable fatigue life of ASME pressure boundary components, with consideration of the light-water reactor (LWR) environment. Because of significant conservatism in quantifying other plant-related variables (such as cyclic behavior, including stress and loading rates) involved in cumulative fatigue life calculations, the design of the current fleet of reactors is satisfactory. For new plants under design and current operating plants considering applying for License Renewal, the environment effects may need to be considered in the design. RG 1.207 proposes using an environmental correction factor (Fen) to account for LWR environments by correcting the fatigue usage calculated with the ASME “air” curves. The Fen method is presented in NUREG/CR-6909, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials”. By definition, Fen is the ratio of fatigue life of the component material at room temperature air environments to its fatigue life in LWR coolant at operating temperature. To incorporate environmental effects into the fatigue evaluation, the fatigue usage is calculated using provisions set forth in Section III of the ASME Code, and is adjusted by multiplying a correction factor. The calculated Fen values are then used to incorporate environmental effects into ASME fatigue usage factor evaluation. Once the environmental correction factors have been determined, the previously calculated allowable number of cycles for each load set pair based on the current Code fatigue design curve can be adjusted to determine the new fatigue usage factors for environmental effects. This paper presents a study of the effect of the Regulatory Guide if it is to be implemented on the current fleet of LWR. A quick assessment of the sensitivity of the various environmental parameters is also included in this paper. The comparison of environmental effects between the simplified approach in this paper and the results with detailed computer analyses, such as Unisont’s propriety computer code UPIPENB (Ref. 4), will be our next research project to be presented in the future conference.

Low Cycle Fatigue of Nuclear Pipe Components

1974 ◽  
Vol 96 (3) ◽  
pp. 171-176 ◽  
Author(s):  
J. D. Heald ◽  
E. Kiss
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Room Temperature ◽  
Fatigue Testing ◽  
Low Cycle Fatigue ◽  
304 Stainless Steel ◽  
Code Design ◽  
Hydraulic Pressure ◽  
Cycle Fatigue ◽  
Cyclic Bending

This paper presents the results of low-cycle fatigue testing and analysis of 26 piping components and butt-welded sections. The test specimens were fabricated from Type-304 stainless steel and carbon steel, materials which are typically used in the primary piping of light water nuclear reactors. Components included 6-in. elbows, tees, and girth butt-welded straight sections. Fatigue testing consisted of subjecting the specimens to deflection-controlled cyclic bending with the objective of simulating system thermal expansion type loading. Tests were conducted at room temperature and 550 deg F, with specimens at room temperature subjected to 1050 psi constant internal hydraulic pressure in addition to cyclic bending. In two tests at room temperature, however, stainless steel elbows were subjected to combined simultaneous cyclic internal pressure and cyclic bending. Predictions of the fatigue life of each of the specimens tested have been made according to the procedures specified in NB-3650 of Section III[1] in order to assess the code design margin. For the purpose of the assessment, predicted fatigue life is compared to actual fatigue life which is defined as the number of fatigue cycles producing complete through-wall crack growth (leakage). Results of this assessment show that the present code fatigue rules are adequately conservative.

Environmental Fatigue Evaluation for Primary Components of APR1400 Reactor Coolant System

2010 ◽  
Author(s):  
J. M. Kim ◽  
K. W. Kim ◽  
K. S. Yoon ◽  
S. H. Park ◽  
I. Y. Kim ◽  
...  
Keyword(s):  
Environmental Effects ◽  
Integral Method ◽  
Power Reactor ◽  
Light Water Reactor ◽  
Outlet Nozzle ◽  
Design Life ◽  
Class 1 ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Coolant System

USNRC Regulatory Guide (RG) 1.207 provides a guideline for evaluating fatigue analyses due to the environmental effects on the new light water reactor (LWR). The environmental correction factor (Fen) is used to incorporate the LWR environmental effect into fatigue analyses of ASME Class 1 components. In this paper, the environmental fatigue evaluation is applied to some primary components with 60 year design life of Advanced Power Reactor (APR1400). The materials sampled from Class 1 components are the low alloy steel for the reactor vessel (RV) outlet nozzle and the carbon steel for the hot leg which are attached to the outlet nozzle. The simplified method, time-based integral method and strain-based integral method are used to compute the Fen values. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. As the calculated cumulative fatigue usage factor considering environmental effects (CUFen) is below 1.0, there is no concern for the RV outlet nozzle to implement design for environmental fatigue effects.

Proposed Code Case of Creep Fatigue Evaluation of 9Cr-1Mo-V Steels for High Pressure Vessels in ASME Section VIII Division 2

2015 ◽  
Author(s):  
Susumu Terada ◽  
Masato Yamada ◽  
Tomoaki Nakanishi
Keyword(s):  
High Pressure ◽  
High Temperature ◽  
Fatigue Curve ◽  
Pressure Vessels ◽  
Inelastic Analysis ◽  
Creep Fatigue ◽  
Asme Code ◽  
Section Viii ◽  
Fatigue Evaluation ◽  
Division 2

9Cr-1Mo-V steels (Gr. 91), which has an excellent performance at high temperature in mechanical properties and hydrogen resistance, has been used for tubing and piping materials in power industries and it can be a candidate material for high pressure vessels for high temperature processes in refining industries. The current Section VIII Division 2 of ASME code does not permit method A of paragraph 5.5.2.3 to be used for the exemption from fatigue analysis for Gr. 91 steels due to limitation of specified minimum tensile strength (585 MPa > 552 MPa). Method B of paragraph 5.5.2.4 also can’t be used because it requires the use of the fatigue curve which is limited to 371 °C lower than the needed temperature. Therefore new rules for fatigue evaluation of Gr. 91 steels at temperatures greater than 371 °C and less than 500 °C similar to CC 2605 for 2.25Cr-1Mo-0.25V(Gr. 22V) steels are necessary. This paper provides fatigue test results at 500 °C for Gr. 91 steels, the modification of CC 2605, sample inelastic analysis results for nozzles. Then, the new Code Case for Gr. 91 steels is proposed from these results.

scholarly journals Evaluation of Primary-Load Effects on Creep-Fatigue Life of Alloy 617 Using Simplified Model Test Method

2021 ◽  
Author(s):  
Yanli Wang ◽  
Peijun Hou ◽  
Robert I. Jetter ◽  
T.-L. Sham
Keyword(s):  
Fatigue Life ◽  
Internal Pressure ◽  
Model Test ◽  
Test Method ◽  
Simplified Model ◽  
Evaluation Procedure ◽  
Alloy 617 ◽  
Creep Fatigue ◽  
Fatigue Evaluation ◽  
Primary Load

Abstract Current creep-fatigue evaluation approaches based on the creep-fatigue Damage-diagram are complex and very conservative. Simplified Model Test (SMT) method is an alternative approach to determine cyclic life at elevated temperatures. The SMT-based creep-fatigue evaluation methodology avoids parsing the damage into creep and fatigue components and greatly simplifies the evaluation procedure for elevated-temperature cyclic service. In this study, the effects of sustained primary-stress loading are evaluated in support of the development of SMT-based creep-fatigue design curves for Alloy 617. Experiments were designed and performed using internal pressurized tubular specimens at 950 °C on Alloy 617. The sustained primary-load was introduced by the internal pressure. A newly developed SMT technique, single-bar SMT, was extended to these tests and SMT creep-fatigue test data were generated with various elastic follow-ups, internal pressures and strain ranges. The test results from this study along with the original SMT data on Alloy 617 demonstrate that, although internal pressure is within the allowable stress limit per ASME Section III Division 5 Code Case N-898, the SMT creep-fatigue cycles to failure were reduced for the cases tested with primary-pressure load. The reduction of SMT creep-fatigue life due to primary-load was found to be dependent on strain ranges and elastic follow up. Approaches to handle the primary-load effect on SMT design curves are discussed.

Development of Creep–Fatigue Evaluation Method for 316FR Stainless Steel

2015 ◽  
Vol 137 (4) ◽  
Author(s):  
Yuji Nagae ◽  
Shigeru Takaya ◽  
Tai Asayama
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Evaluation Method ◽  
Elevated Temperatures ◽  
Strain Curve ◽  
Creep Damage ◽  
Fatigue Curve ◽  
Time To Failure ◽  
Creep Fatigue ◽  
Fatigue Evaluation

In the design of fast reactor plants, the most important failure mode to be prevented is creep–fatigue damage at elevated temperatures. 316FR stainless steel is a candidate material for the reactor vessel and internal structures of such plants. The development of a procedure for evaluating creep–fatigue life is essential. The method for evaluating creep–fatigue life implemented in the Japan Society of Mechanical Engineers code is based on the time fraction rule for evaluating creep damage. Equations such as the fatigue curve, dynamic stress–strain curve, creep rupture curve, and creep strain curve are necessary for calculating creep–fatigue life. These equations are provided in this paper and the predicted creep–fatigue life for 316FR stainless steel is compared with experimental data. For the evaluation of creep–fatigue life, the longest time to failure is about 100,000 h. The creep–fatigue life is predicted to an accuracy that is within a factor of 2 even in the case with the longest time to failure. Furthermore, the proposed method is compared with the ductility exhaustion method to investigate whether the proposed method gives conservative predictions. Finally, a procedure based on the time fraction rule for the evaluation of creep–fatigue life is proposed for 316FR stainless steel.

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