Study on Simplified Prediction Method of Thermal Ratchet Deformation Based on Parallel Bar Model

2014 ◽  
Author(s):  
Toshiaki Kokufuda ◽  
Naoto Kasahara
Keyword(s):  
Elevated Temperature ◽  
Inelastic Deformation ◽  
Evaluation Method ◽  
Failure Modes ◽  
Prediction Method ◽  
Inelastic Strain ◽  
Primary Stress ◽  
Redistribution Mechanism ◽  
Creep Fatigue ◽  
Calculation Results

For elevated temperature structures such as fast breeder reactor components, inelastic deformation is likely to occur because of reduction of yield stress and occurrence of creep deformation. The typical failure modes for elevated temperature structures are excessive deformation caused by the accumulation of inelastic deformation and creep fatigue caused by inelastic strain concentration at structure discontinuities. In order to prevent such failures, it is necessary to evaluate inelastic deformation adequately. Thermal ratchet deformation, namely the progressive plastic deformation induced by cyclic thermal stress with uniform primary stress, has some possibility resulting in excessive deformation. ASME boiler & pressure vessel code provides elastic evaluation methods for thermal ratchet. However, these methods are so focused on preventing thermal ratchet deformation, that it could be too conservative under some conditions. Therefore, a simplified elastic evaluation method to quantify thermal ratchet deformation is desired. In this paper, the simplified prediction method for thermal ratchet deformation is proposed using parallel bar model, which represents stress redistribution mechanism of cylindrical vessels. The solution of thermal ratchet deformation of parallel bar model was derived and compared with FEM calculation results of cylindrical vessels. This theoretical solution is proposed as a prediction method for thermal ratchet deformation of cylindrical vessels. The applicable area of the proposed prediction method is the cylindrical vessel under linear and parabolic temperature distribution through the wall thickness.

Study on Life Prediction Method for Creep-Fatigue Interaction at Elevated Temperature

2007 ◽  
Vol 353-358 ◽  
pp. 190-194
Author(s):  
Nian Jin Chen ◽  
Zeng Liang Gao ◽  
Wei Zhang ◽  
Yue Bao Le
Keyword(s):  
Elevated Temperature ◽  
Prediction Model ◽  
Life Prediction ◽  
Low Cycle Fatigue ◽  
Hold Time ◽  
Prediction Method ◽  
Fatigue Tests ◽  
Cycle Fatigue ◽  
Creep Fatigue ◽  
316L Austenitic Stainless Steel

The law of low-cycle fatigue with hold time at elevated temperature is investigated in this paper. A new life prediction model for the situation of fatigue and creep interaction is developed, based on the damage due to fatigue and creep. In order to verify the prediction model, strain-controlled low-cycle fatigue tests at temperature 693K, 823K and 873K and fatigue tests with various hold time at temperature 823K and 873K for 316L austenitic stainless steel were carried out. Good agreement is found between the predictions and experimental results.

Methodology for the Determination of a Set of Safety Factors That are Consistent for Design Code and Fitness-for-Service Code: Case Study for Fast Breeder Reactors

2009 ◽  
Author(s):  
Tai Asayama
Keyword(s):  
Failure Modes ◽  
Fatigue Crack Initiation ◽  
Seismic Load ◽  
Safety Factors ◽  
Design Code ◽  
Secondary Stress ◽  
Primary Stress ◽  
Creep Fatigue ◽  
Fast Breeder Reactors ◽  
Breeder Reactors

This paper proposes a method for determining a set of safety factors taking account of multiple failure modes and their interactions. The purpose of the work is to materialize the System Based Code concept for the applications to fast breeder reactors. Current structural codes prevent failure by limiting primary stress, excessive strain, buckling and fatigue damage. However, the relationships between safety factors in these criterion and failure modes are not necessarily clear. For example, safety factors in the limitation of primary stress are considered to cover not only ductile fracture but also fracture due to crack like defects. When the System Based Code concepts, one of the most important of which is designing to target reliability, it is essential to determine safety factors so that they explicitly correspond to particular failure modes. This paper deals with the most important two failure modes to be prevented in fast breeder reactors (FBRs), that is, primary stress due to seismic load and secondary stress due to creep-fatigue. Safety factors that are consistent for design code and fitness-for-service code are derived by the following steps: 1) Formulations of continuous evaluation of reliability are derived for both fracture by primary stress and creep-fatigue crack initiation and propagation due to secondary stress, with their interaction taken into account, 2) Reliability is calculated for various combinations of loading conditions, 3) Safety factors corresponding to various levels of target reliabilities are investigated and compared with currently used ones. The safety factors thus determined not only have firm physical basis but also contribute to enlarge design windows for fast breeder reactor components. Items to be further investigated for the methodology to be implemented in current code are also discussed.

Pressurized Creep-Fatigue Testing of Alloy 617 Using Simplified Model Test Method

2017 ◽  
Author(s):  
Yanli Wang ◽  
Robert I. Jetter ◽  
T.-L. Sham
Keyword(s):  
Elevated Temperature ◽  
Hysteresis Loop ◽  
Fatigue Damage ◽  
Model Test ◽  
Test Method ◽  
Simplified Model ◽  
Damage Evaluation ◽  
Primary Stress ◽  
Creep Fatigue ◽  
Cyclic Damage

The Simplified Model Test (SMT) is an alternative approach to determine cyclic life at elevated temperature and avoids parsing the damage into creep and fatigue components. The Elastic-Perfectly Plastic (EPP) combined integrated creep-fatigue damage evaluation approach incorporates the SMT data based approach for creep-fatigue damage evaluation into the EPP methodology to avoid the separate evaluation of creep and fatigue damage and to eliminate the requirement for stress classification as in current methods; thus greatly simplifying evaluation of elevated temperature cyclic service. The conceptual basis of the SMT approach is that if the effects of plasticity, creep and strain redistribution in the SMT specimen result in a stress-strain hysteresis loop that envelopes the hysteresis loop at the peak strain location in the component, then the SMT results can be used to assess the cyclic damage in the component. The original SMT concept (Jetter, 1998) considered that the effects of sustained primary stress loading could be safely neglected because the allowable local stress and strain levels were much higher than the allowable sustained primary stress levels. This key assumption requires experimental verification. The influence of the internal pressure on SMT creep-fatigue life is demonstrated and the effect of primary load on the SMT design approach is discussed.

Creep-Fatigue Life Evaluation Method for Perforated Plates at Elevated Temperature

2005 ◽  
Vol 128 (1) ◽  
pp. 17-24 ◽  
Author(s):  
Osamu Watanabe ◽  
Takuya Koike
Keyword(s):  
Elevated Temperature ◽  
Fatigue Strength ◽  
Stress Concentration ◽  
Evaluation Method ◽  
Three Dimensional ◽  
Perforated Plate ◽  
Dimensional Structure ◽  
Creep Process ◽  
Plastic Behavior ◽  
Creep Fatigue

The accurate evaluation scheme for creep-fatigue strength is one of the continuing main issues for elevated temperature design; particularly, the three-dimensional structure having stress concentration is becoming more important. The present paper investigates fatigue strength and creep-fatigue strength of perforated plate having stress concentration as an example. The specimens are made of type 304 SUS stainless steel, and the temperature is kept to 550°C. The whole cycles of the experiment record are analyzed, and the characteristics of the structure having stress concentration are discussed. The present paper employs stress redistribution locus (abbreviated as SRL) in evaluation plastic behavior in cyclic fatigue process as well as stress relaxation in creep process, and the feasibility is discussed in conjunction with the comparison to experimental results.

Fatigue and Creep-Fatigue Testing of Bellows at Elevated Temperature

1988 ◽  
Vol 110 (3) ◽  
pp. 301-307 ◽  
Author(s):  
S. Yamamoto ◽  
K. Isobe ◽  
S. Ohte ◽  
N. Tanaka ◽  
S. Ozaki ◽  
...  
Keyword(s):  
Elevated Temperature ◽  
Life Prediction ◽  
Room Temperature ◽  
Fatigue Testing ◽  
Prediction Method ◽  
Fatigue Tests ◽  
Finite Element Analyses ◽  
Strain Behavior ◽  
Creep Fatigue ◽  
Inner Diameter

Fatigue and creep-fatigue tests at elevated temperature were conducted on two different-sized bellows, φ 1100 mm and φ 300 mm in nominal inner diameter, to investigate the fatigue life and the creep-fatigue interaction in a bellows, and also to provide test data for developing a life prediction method and design-by-analysis rules for bellows in elevated temperature service. A series of tests consisted of strain behavior and fatigue tests at room temperature, and fatigue and creep-fatigue tests at elevated temperature. Also, inelastic finite element analyses were performed on a bellows under internal pressure and cyclic axial deflections. Analytical results were compared with the measured data obtained in the room temperature testing to verify the strain prediction method.

Land Subsidence Prediction Method of Power Cables Pipe Jacking Based on the Peck Theory

2013 ◽  
Vol 634-638 ◽  
pp. 3721-3724
Author(s):  
Yuan Liang Zhang ◽  
Yi Hu Zhang
Keyword(s):  
Land Subsidence ◽  
Prediction Method ◽  
Dominant Factor ◽  
Engineering Practice ◽  
Foundation Settlement ◽  
Power Cables ◽  
Overhead Transmission Line ◽  
Allowable Range ◽  
Calculation Results ◽  
Range Calculation

Overhead transmission line and cable are generally used for across or crossing the railway, highways and rivers.For higher deformation requirement of operation of the railway and highway foundation settlement,to ensure that the cable through the process of foundation in regulating the allowable range, calculation and prediction of foundation settlement is specially necessary.Based on the Peck theory, the dominant factor in foundation settlement-strata loss calculation method is introduced supplemented with measured settlement observation records in this paper,which confirmes that the calculation results are replicab in engineering practice.

Development of Ultimate Strength Evaluation Method for Piping Subjected to Seismic Loads

2013 ◽  
Author(s):  
Hideo Machida ◽  
Hiromasa Chitose ◽  
Tatsuhiro Yamazaki
Keyword(s):  
Power Plants ◽  
Evaluation Method ◽  
Failure Modes ◽  
Limit State ◽  
Seismic Loading ◽  
Earthquake Resistance ◽  
State Function ◽  
Limit State Function ◽  
Input Acceleration ◽  
Resistance Tests

This paper reports the results of the study on the failure modes and limit loads of piping in nuclear power plants subjected to cyclic seismic loading. By investigating the past fracture tests and earthquake resistance tests, it became clear that dominant failure mode of piping was fatigue, and the effect of ratchet strain was negligible. Until now, the stress generated with the acceleration of an earthquake was classified into the primary stress. However, the relationship between the input acceleration and the seismic response displacement of the pipe observed from earthquake resistance tests is non-linear, and increasing rate of displacement is lower than that of input acceleration in elastic-plastic stress condition. Therefore, the seismic loading can be treated as displacement controlled loading. To evaluate the reliability-based critical acceleration, a limit state function was defined taking the variations in the fatigue strength or some parameters into consideration. By using the limit state function, the reliability was evaluated for the typical piping of boiling water reactor (BWR) plants subjected to cyclic seismic loading, and a partial safety factors were calculated. Based on these results, a fatigue curve corresponding to the target reliability was proposed.

Structural Integrity Code and Regulatory Issues in the Design of High Temperature Reactors

2008 ◽  
Author(s):  
William J. O’Donnell ◽  
Amy B. Hull ◽  
Shah Malik
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Structural Integrity ◽  
Cyclic Creep ◽  
Creep Rupture ◽  
Creep Fatigue ◽  
Asme Code ◽  
High Temperature Reactors ◽  
Gen Iv ◽  
Very High

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.

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