Structural Evaluation With Elastic and Inelastic Analysis Methods on a High Temperature Storage Tank Subjected to Static and Dynamic Loadings

2020 ◽  
Author(s):  
Wen Wang ◽  
Xiaochun Zhang ◽  
Xiaoyan Wang ◽  
Maoyuan Cai
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Molten Salt ◽  
Fatigue Damage ◽  
Storage Tank ◽  
Power Plants ◽  
Elastic Analysis ◽  
Inelastic Analysis ◽  
Analysis Methods ◽  
Creep Fatigue

Abstract The structural integrity of reactor components is very essential for the reliable operation of all types of power plants, especially for components operating at elevated temperature where creep effects are significant and where components are subjected to high-temperature alteration and seismic transient loading conditions. In this article, a molten salt storage tank in high temperature thorium molten salt reactor (TMSR) is evaluated according to ASME-III-5-HBB high temperature reactor code. The evaluation based on 3D finite element analyses includes the load-controlled stress, the effects of ratcheting, and the interaction of creep and fatigue. The thermal and structural analysis and the application procedures of ASME-HBB rules are described in detail. Some structural modifications have been made on this molten salt storage tank to enhance the strength and reduce thermal stress. The effects of ratcheting and creep-fatigue damage under elevated temperature are investigated using elastic analysis and inelastic analysis methods for a defined representative load cycle. In addition, the strain range and the stress relaxation history calculated by elastic and inelastic methods are compared and discussed. The numerical results indicate that the elastic analysis is conservative for design and a full inelastic analysis method for estimating input for creep-fatigue damage evaluation need to be developed.

Development of Guideline for Prediction of Inelastic Strain for Creep-Fatigue Evaluation

2013 ◽  
Author(s):  
Osamu Watanabe ◽  
Ken-ichi Kobayashi ◽  
Kyotada Nakamura
Keyword(s):  
High Temperature ◽  
Power Plants ◽  
Inelastic Strain ◽  
Local Area ◽  
Stage Of Development ◽  
Inelastic Analysis ◽  
Design Changes ◽  
Creep Fatigue ◽  
High Temperature Components ◽  
Fatigue Evaluation

Cyclic thermal and mechanical loads are frequently applied to power plants during their service lives due to the regular operation of start-up and shutdown. Design or actual lives of these high temperature machines and structures have been mainly dominated by the creep-fatigue failure life. Since most of these failures happen at limited local area, namely, it may happen at the geometrical or material discontinuities in structures or components, the detail inelastic analyses with a conservative margin are required at the design and maintenance. However, much time and colossal effort should be avoided at the stage of development to reduce the total cost of designing because the design changes many times until the final configuration is fixed. Many materials in the high temperature components are subjected to inelastic behaviors; plastic or creep strain always cause in the components. In the computational analyses such as Finite Element Analyses, constitutive equations of both plasticity and creep affect analytical results. Neuber’s rule is employed in the present design code to achieve the simplified design of component but its result sometimes provides more conservative margin. Stress Redistribution Locus (Hereinafter denoted as SRL) method is a simplified inelastic analysis and was developed in Japan. ETD committee in HPI has studied its applicability to basic problems and actual components.

Simplified Methods for Elevated Temperature Structural Design: An Overview of Some Current Activities

2014 ◽  
Author(s):  
Robert I. Jetter ◽  
Yanli Wang ◽  
Peter Carter ◽  
T.-L. (Sam) Sham
Keyword(s):  
Elevated Temperature ◽  
Fatigue Damage ◽  
Temperature Regime ◽  
Structural Integrity ◽  
Stress And Strain ◽  
Damage Evaluation ◽  
Real Component ◽  
Perfectly Plastic ◽  
Inelastic Analysis ◽  
Creep Fatigue

Elevated temperature design criteria for Class 1 nuclear components employ two fundamental approaches for evaluation of structural integrity in the temperature regime where creep effects are significant: full inelastic analysis to predict the actual stress and strain resulting from time dependent loading conditions and simplified methods which bound the actual response with, conceptually, simpler material models and analytical procedures. However, the current simplified methods have been found to be more complex for real component design applications than originally envisioned. There is an added complication that the current simplified methods are considered inappropriate in the very high temperature regime where there is no distinction between plasticity and creep. Recently, some improved, less complex methods have been proposed which would overcome these objections. One set of criteria are based on elastic-perfectly plastic (E-PP) analysis methods. Draft code cases have been prepared which address the use of the E-PP methodology to primary loading, strain limits and creep-fatigue damage evaluation. Another proposed criterion is based on the use of test specimens which include the effects of stress and strain redistribution due to plasticity and creep to develop creep-fatigue damage evaluation design curves. An overview of the key features, associated analytical and experimental verification, status and path forward are presented. Although targeted to nuclear components, these criteria also have potential application to non-nuclear components and operating temperatures below the creep regime. Paper published with permission.

Creep-Fatigue Assessment Methods Using Elastic Analysis Results and Adjustments

1991 ◽  
Vol 113 (1) ◽  
pp. 34-40 ◽  
Author(s):  
L. K. Severud
Keyword(s):  
Elevated Temperature ◽  
Stress Analysis ◽  
Elastic Analysis ◽  
Practical Application ◽  
New Methods ◽  
Inelastic Analysis ◽  
Creep Fatigue ◽  
Adjustment Factors ◽  
Asme Code

The design of mechanical components that operate in elevated temperature environment where creep effects are significant usually requires creep-fatigue assessments. The ASME Code Case N-47 contains rules for these assessments based on both inelastic and elastic stress analysis. Although an inelastic stress analysis generally more accurately predicts effects from creep and plasticity, an elastic analysis is often preferred since it is much simpler and less costly. New creep-fatigue rules for use with elastic analysis results have recently been proposed to enhance the rules’ accuracy and usefulness. This paper describes such new methods and rules for creep-fatigue assessments. Ingredients of the new methods include elastic follow-up, ratcheting, multiaxiality, plasticity, creep, and relaxation considerations and associated adjustment factors. The basis for the adjustments and a comparison of results to those obtained using inelastic analysis are provided. The new methods will provide a wider range of practical application of elastic creep-fatigue rules than permitted by previous code methods in design of components for elevated temperature service.

Pathway to Evaluate Printed Circuit Heat Exchanger Based on Simplified Elastic-Perfectly Plastic Analysis Methodology for High Temperature Nuclear Service

2019 ◽  
Author(s):  
Urmi Devi ◽  
Machel Morrison ◽  
Tasnim Hassan
Keyword(s):  
High Temperature ◽  
Fatigue Damage ◽  
Analysis Method ◽  
Printed Circuit ◽  
Perfectly Plastic ◽  
Analysis Methodology ◽  
Inelastic Analysis ◽  
Creep Fatigue ◽  
Asme Code ◽  
Elastic Perfectly Plastic

Abstract Printed Circuit Heat Exchangers (PCHEs) are well-suited for Very High Temperature Reactors (VHTRs) due to high compactness and efficiency for heat transfer. The design of PCHE must be robust enough to withstand possible failure caused by cyclic loading during high temperature operation. The current rules in ASME Code Section III Division 5 to evaluate strain limits and creep-fatigue damage based on elastic analysis method have been deemed infeasible at temperatures above 650°C. Hence, these rules are inapplicable for temperatures ranging from 760–950°C for VHTRs. A full inelastic analysis method with complex constitutive material description as an alternative, on the other hand, is time consuming; hence impracticable. Therefore, the simplified Elastic-Perfectly Plastic (EPP) analysis methodology is used as a solution in ASME Code Section III Division 5. The current literature, however, lacks any study on the performance evaluation of PCHE through EPP analysis. To address these issues, this study initiates the pathway of EPP evaluation of an actual size PCHE starting with elastic orthotropic analysis in the global scale. Subsequently, preliminary planning for analyzing intermediate and local submodels are provided to determine channel level responses to evaluate PCHE performance against strain limits and creep-fatigue damage using Code Case-N861 and N862 respectively.

scholarly journals Probabilistic Life Models for Steel Structures Subject to Creep- Fatigue Damage

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Fatmagul Ibisoglu ◽  
Mohammad Modarres
Keyword(s):  
High Temperature ◽  
Fatigue Damage ◽  
Power Plants ◽  
Strain Range ◽  
Steel Structures ◽  
Software Tool ◽  
Bayesian Regression ◽  
Life Assessment ◽  
Total Strain Range ◽  
Creep Fatigue

When metal structures are subjected to long-term cyclic loading at high temperature, simultaneous creep and fatigue damage may occur. In this paper probabilistic life models, described by hold times in tension and total strain range at elevated temperature have been derived based on the creeprupture behavior of 316FR austenitic stainless steel, which is one of the candidate structural materials for fast reactors and future Generation IV nuclear power plants operating at high temperatures. The parameters of the proposed creepfatigue model were estimated using a standard Bayesian regression approach. This approach has been performed using the WinBUGS software tool, which is an open source Bayesian analysis software tool that uses the Markov Chain Monte Carlo sampling method. The results have shown a reasonable fit between the experimental data and the proposed probabilistic creep-fatigue life assessment models. The models are useful for predicting expended life of the critical structures in advanced high temperature reactors when performing structural health management.

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.

Application of Damage Mechanics and Polynomial Chaos Expansion for Lifetime Prediction of High-Temperature Components Under Creep-Fatigue Loading

2020 ◽  
Author(s):  
Felix Koelzow ◽  
Muhammad Mohsin Khan ◽  
Christian Kontermann ◽  
Matthias Oechsner
Keyword(s):  
High Temperature ◽  
Power Plants ◽  
Damage Mechanics ◽  
Fatigue Loading ◽  
Probabilistic Methods ◽  
Lifetime Prediction ◽  
Model Parameters ◽  
Creep Fatigue ◽  
High Temperature Components ◽  
Mechanics Concepts

Abstract Several (accumulative) lifetime models were developed to assess the lifetime consumption of high-temperature components of steam and gas turbine power plants during flexible operation modes. These accumulative methods have several drawbacks, e.g. that measured loading profiles cannot be used within accumulative lifetime methods without manual corrections, and cannot be combined directly to sophisticated probabilistic methods. Although these methods are widely accepted and used for years, the accumulative lifetime prediction procedures need improvement regarding the lifetime consumption of thermal power plants during flexible operation modes. Furthermore, previous investigations show that the main influencing factor from the materials perspective, the critical damage threshold, cannot be statistically estimated from typical creep-fatigue experiments due to massive experimental effort and a low amount of available data. This paper seeks to investigate simple damage mechanics concepts applied to high-temperature components under creep-fatigue loading to demonstrate that these methods can overcome some drawbacks and use improvement potentials of traditional accumulative lifetime methods. Furthermore, damage mechanics models do not provide any reliability information, and the assessment of the resultant lifetime prediction is nearly impossible. At this point, probabilistic methods are used to quantify the missing information concerning failure probabilities and sensitivities and thus, the combination of both provides rigorous information for engineering judgment. Nearly 50 low cycle fatigue experiments of a high chromium cast steel, including dwell times and service-type cycles, are used to investigate the model properties of a simple damage evolution equation using the strain equivalence hypothesis. Furthermore, different temperatures from 300 °C to 625 °C and different strain ranges from 0.35% to 2% were applied during the experiments. The determination of the specimen stiffness allows a quantification of the damage evolution during the experiment. The model parameters are determined by Nelder-Mead optimization procedure, and the dependencies of the model parameters concerning to different temperatures and strain ranges are investigated. In this paper, polynomial chaos expansion (PCE) is used for uncertainty propagation of the model uncertainties while using non-intrusive methods (regression techniques). In a further post-processing step, the computed PCE coefficients of the damage variable are used to determine the probability of failure as a function of cycles and evolution of the probability density function (pdf). Except for the selected damage mechanics model which is considered simple, the advantages of using damage mechanics concepts combined with sophisticated probabilistic methods are presented in this paper.

A Viscoplastic Model for Alloy 617 for Use With the ASME Section III, Division 5 Design by Inelastic Analysis Rules

2021 ◽  
Author(s):  
M. C. Messner ◽  
T.-L. Sham
Keyword(s):  
High Temperature ◽  
Alloy 617 ◽  
Element Analysis ◽  
Material Models ◽  
Experimental Database ◽  
Viscoplastic Model ◽  
Analysis And Design ◽  
Inelastic Analysis ◽  
Wide Range ◽  
Creep Fatigue

Abstract The rules for the design of high temperature reactor components in Section III, Division 5, Subsection HB, Subpart B (HBB) of the ASME Boiler and Pressure Vessel Code contain two options for evaluating the deformation-controlled design limits on strain accumulation and creep-fatigue: design by elastic analysis and design by inelastic analysis. Of these options design by inelastic analysis tends to be less overconservative and produce more efficient designs. However, the HBB currently does not provide approved material models for use with the inelastic analysis rules, limiting their widespread use. A nonmandatory appendix has been developed to provide general guidance on appropriate material models and provide reference material models suitable for use with the design by inelastic analysis approach. This paper describes a viscoplastic model for Alloy 617 suitable for use with the HBB rules proposed for incorporation into the new appendix. The model represents the high temperature creep, creep-fatigue, and tensile response of Alloy 617 and accurately accounts for rate sensitivity across a wide range of temperatures. The focus in developing the model was on capturing key features of material deformation required for accurately executing the HBB rules and on developing a relatively simple model form that can be implemented in commercial finite element analysis software. The paper validates the model against an extensive experimental database collected as part of the Alloy 617 Code qualification effort as well as against specialized experimental tests examining the effect of elastic follow up on stress relaxation and creep deformation in the material.

The Mechanical Interaction of Clad and Base Metal for Molten Salt Reactor Structural Components

2018 ◽  
Author(s):  
M. C. Messner ◽  
V.-T. Phan ◽  
R. I. Jetter ◽  
T.-L. Sham
Keyword(s):  
High Temperature ◽  
Molten Salt ◽  
Mechanical Response ◽  
Base System ◽  
Molten Salt Reactor ◽  
Structural Components ◽  
Mechanical Interaction ◽  
Base Materials ◽  
Creep Fatigue ◽  
Axisymmetric Structures

Cladding structural components with a corrosion resistant material may greatly extend the design life of molten salt reactor concepts. A complete design methodology for such cladded, high temperature nuclear components will require addressing many issues: fabrication, corrosion resistance, metallurgical interaction, and the mechanical interaction of the clad and base materials under load. This work focuses on the final issue: the mechanical interaction of the base and clad under creep-fatigue conditions. Depending on the relative mechanical properties of the two materials the clad may substantially influence the long-term cyclic response of the structural system or its effect might be negligible. To quantify the effect of different clad material properties we develop an efficient method for simulating pressurized cladded components in the limiting case where the section of interest is far from structural discontinuities. Using this method we evaluate the mechanics of the clad/base system and identify different regimes of mechanical response. The focus is on situations relevant to high temperature nuclear components: thermal-cyclic Bree-type problems and similar axisymmetric structures. The insights gained from these structural studies will form the basis for developing design rules for high-temperature, nuclear, cladded components.

Assessment of Fatigue Life for High-Temperature Pipeline Welds Using X-Ray Diffraction Technique

2007 ◽  
Vol 353-358 ◽  
pp. 130-133
Author(s):  
Keun Bong Yoo ◽  
Jae Hoon Kim
Keyword(s):  
High Temperature ◽  
Fatigue Life ◽  
Fatigue Damage ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Diffraction Method ◽  
Life Assessment ◽  
Fatigue Properties ◽  
X Ray Diffraction ◽  
X Ray

The objective of this study is to examine the feasibility of the X-ray diffraction method for the fatigue life assessment of high-temperature steel pipes used for main steam pipelines, re-heater pipelines and headers etc. in power plants. In this study, X-ray diffraction tests were performed on the specimens simulated for low cycle fatigue damage, in order to estimate fatigue properties at the various stages of fatigue life. As a result of X-ray diffraction tests, it was confirmed that the full width at the half maximum (FWHM) decreased with an increase in the fatigue life ratio, and that the FWHM and the residual stress due to fatigue damage were algebraically linearly related to the fatigue life ratio. From this relationship, a direct assessment of the remaining fatigue life was feasible.

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