Structural Integrity Assessment of Weldments at High Temperature: A Proposed New Approach for R5

2008 ◽  
Author(s):  
N. G. Smith ◽  
D. W. Dean ◽  
M. P. O’Donnell
Keyword(s):  
High Temperature ◽  
Fatigue Damage ◽  
Structural Integrity ◽  
Fatigue Crack Initiation ◽  
Current Approach ◽  
Assessment Procedure ◽  
New Approach ◽  
Integrity Assessment ◽  
Creep Fatigue ◽  
Match Effect

The majority of problems associated with the structural integrity of components, particularly those operating at high temperature, are associated with welds. The R5 procedures provide a comprehensive methodology for the assessment of structures operating within the high temperature creep regime. This includes advice on the modifications required to the basic procedure to account for weldments in creep-fatigue crack initiation assessments. The current approach is based on the use of a Fatigue Strength Reduction Factor (FSRF) which has a value according to the particular class of welded joint. The FSRF affects the calculation of creep and fatigue damage. However, the current approach can be excessively conservative for as-welded weldments which are the main type of weldments in plant. This paper outlines the proposed changes to R5, which seek to achieve the following objectives: • to simplify and clarify the current advice for creep-fatigue initiation assessments of weldments, whilst maintaining a conservative assessment procedure; • to have a robust procedure which can be applied to complex components and loading conditions. The new approach separates the FSRF into two components which are as follows: • the geometric strain enhancement due to the weldment geometry (if applicable) and the material mis-match effect between parent material and weld metal, which is called the Weld Strain Enhancement Factor (WSEF), and • the fatigue endurance reduction effect due to the presence of small imperfections (e.g. inclusions, porosity, etc.) in the weldment constituent materials, which is called the Weld Endurance Reduction (WER). The WSEF is used to determine the stress at the start of a dwell or hold period and, because it has a lower value than the FSRF (due to the removal of the WER), results in less conservative calculations of creep damage compared to the current procedure, which uses the full FSRF. For fatigue damage predictions, the modified route is broadly similar to the current route, since the combination of the WER and the WSEF in the modified route corresponds to the FSRF used in the current route. Assessments to demonstrate the improved endurance predictions using the proposed new approach have been performed on several creep-fatigue weldment features tests and examples are provided in this paper.

Creep-Fatigue Crack Initiation Assessments of an Instrument Guide Tube Within a Superheater Header

2014 ◽  
Author(s):  
Jinhua Shi ◽  
Hassam Dodia
Keyword(s):  
Fatigue Crack ◽  
Crack Initiation ◽  
Fatigue Damage ◽  
Material Properties ◽  
Fatigue Crack Initiation ◽  
Guide Tube ◽  
New Approach ◽  
Weld Material ◽  
Creep Fatigue ◽  
Creep Fatigue Crack

In order to extend the boiler lives at Advanced Gas-Cooled Reactor (AGR) nuclear power stations in the UK, new temperature measuring instrumentation to monitor reactor gas temperature has been proposed to install on the bore of an intact boiler tube to provide additional boiler operating data to support the station lifetime extension. This paper details a creep-fatigue crack initiation assessment of the proposed installation of an instrument guide tube within the superheater header using the latest R5 high temperature assessment procedures based on detailed finite element thermal transient stress analysis values for a bounding start-up and shutdown cycle. The fatigue damage at welds has been calculated based on both weld and parent material properties. The new approach for assessing weldments has been used in this paper. This new approach involves splitting the existing Fatigue Strength Reduction Factor (FSRF) into a Weldment Endurance Reduction (WER), which accounts for reduced fatigue endurance due to weld imperfections, and a Weldment Strain Enhacement Factor (WSEF), which accounts for material mismatch and local geometry. The creep assessments of the weld material locations have been carried out on both parent and weld material properties including the welding residual stress. The total creep-fatigue damage is then obtained as the sum of fatigue damage, Df, and creep damage, Dc.

Modelling the Complex Manufacturing History of a Pipework Joint and Assessment of Its Through Life Creep-Fatigue Damage Using Finite Element Based Methods

2010 ◽  
Author(s):  
D. P. Bray ◽  
R. J. Dennis ◽  
R. A. W. Bradford
Keyword(s):  
Finite Element ◽  
Fatigue Damage ◽  
Structural Integrity ◽  
Element Model ◽  
General Purpose ◽  
Element Analysis ◽  
Integrity Assessment ◽  
Creep Fatigue ◽  
History Of ◽  
Manufacturing History

The work reported in this paper investigates the complex manufacture and through-life operation of a pipework joint in a UK AGR boiler. Residual stresses resulting from the fabrication process can be a key driver for creep and creep-fatigue damage. The calculation of creep-fatigue damage for assessment purposes is typically undertaken within the framework of an appropriate assessment code (such as British Energy’s R5). The standard assessment approach usually requires the undertaking of elastic finite element analysis followed by Neuber construction to convert elastic stress ranges into elastic-plastic stress and strain ranges prior to the calculation of creep-fatigue damage. A combination of explicit and implicit finite element methods are employed in order to simulate a range of manufacturing processes which influence the material state for a branched pipework joint. The solution is effectively obtained within one finite element model, with re-meshing performed where necessary. This solution then feeds into a finite element based structural integrity assessment. The methods utilise the principles outlined in the British Energy R5 assessment code but utilise the inelastic strains calculated directly from analysis. The methods are based around the general purpose finite-element code Abaqus enhanced by the use of user-defined subroutines CREEP and UVARM. This paper describes analyses performed to simulate the complex manufacturing history of a branched pipework component, and to estimate its subsequent in service creep-fatigue damage using finite element based methods.

Direct Method on Creep Fatigue Damage Assessment Considering Full Creep-Cyclic Plasticity Interaction

2017 ◽  
Author(s):  
Daniele Barbera ◽  
Haofeng Chen ◽  
Weiling Luan
Keyword(s):  
High Temperature ◽  
Fatigue Damage ◽  
Damage Assessment ◽  
Direct Method ◽  
Creep Damage ◽  
Load Cycle ◽  
Cyclic Plasticity ◽  
Assessment Procedure ◽  
Creep Fatigue ◽  
The Uk

This paper introduces the latest research and development of the Linear Matching Method (LMM) on the creep fatigue damage assessment of components subjected to high temperature and cyclic load conditions. The method varies from existing rule-based approaches in both the ASME Boiler and Pressure Vessel Code (NH) and the UK R5 high temperature assessment procedure, where the creep behavior/creep damage and cyclic plastic response /fatigue damage are analyzed separately. In support to these the extended Direct Steady Cycle Analysis (eDSCA) has been proposed to provide a more accurate description of the potentially dangerous interaction between creep and cyclic plasticity during the load cycle, and hence is able to accurately address creep enhanced plasticity and cyclically enhanced creep. The applications of the LMM eDSCA method for creep fatigue damage assessment to three practical problems are then outlined to demonstrate that the proposed direct method is capable of predicting an accurate component life due to creep fatigue and creep ratcheting damages by modeling cyclic plasticity and creep interaction using this new simplified direct method, providing a degree of accuracy and convenience in creep fatigue assessment hitherto unavailable and without the restrictions inherent in other methodologies.

Creep-Fatigue Interaction Effects On Pressure-Reducing Valve Under Cyclic Thermo-Mechanical Loadings Using Direct Cyclic Method

2021 ◽  
Author(s):  
Nak-Kyun Cho ◽  
Youngjae Choi ◽  
Haofeng Chen
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Fatigue Failure ◽  
Material Properties ◽  
Structural Integrity ◽  
Direct Method ◽  
Element Model ◽  
Critical Loading ◽  
Creep Fatigue ◽  
Pressure Reducing Valve

Abstract Supercritical boiler system has been widely used to increase efficiency of electricity generation in power plant industries. However, the supercritical operating condition can seriously affect structural integrity of power plant components due to high temperature that causes degradation of material properties. Pressure reducing valve is an important component being employed within a main steam line of the supercritical boiler, which occasionally thermal-fatigue failure being reported. This research has investigated creep-cyclic plastic behaviour of the pressure reducing valve under combined thermo-mechanical loading using a numerical direct method known as extended Direct Steady Cyclic Analysis of the Linear Matching Method Framework (LMM eDSCA). Finite element model of the pressure-reducing valve is created based on a practical valve dimension and temperature-dependent material properties are applied for the numerical analysis. The simulation results demonstrate a critical loading component that attributes creep-fatigue failure of the valve. Parametric studies confirm the effects of magnitude of the critical loading component on creep deformation and total deformation per loading cycle. With these comprehensive numerical results, this research provides engineer with an insight into the failure mechanism of the pressure-reducing valve at high temperature.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

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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