Development of guideline for structural integrity assessment for fast reactor plants Part I : Evaluation of creep-fatigue damage

2000 ◽  
Vol 2000 (0) ◽  
pp. 607-608
Author(s):  
Yukio TAKAHASHI
Keyword(s):  
Fatigue Damage ◽  
Fast Reactor ◽  
Structural Integrity ◽  
Integrity Assessment ◽  
Creep Fatigue

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.

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.

On a Design Acceptability of 90° Elbow According to Pipe Length at Elevated Temperature Condition

2016 ◽  
Author(s):  
Seong-Yun Jeong ◽  
Min-Gu Won ◽  
Jae-Boong Choi ◽  
Nam-Su Huh ◽  
Young-Jin Oh
Keyword(s):  
Elevated Temperature ◽  
Structural Integrity ◽  
Seismic Damage ◽  
Design Codes ◽  
Finite Element Analyses ◽  
Promising Candidate ◽  
Pipe Length ◽  
Integrity Assessment ◽  
Creep Fatigue ◽  
Elevated Temperature Design

Sodium-cooled Fast Reactor, SFR is promising candidate of Generation-IV reactor. SFR is operated at high temperature and low pressure. For reducing high thermal stress, thin-walled components and structures are employed for SFR. However, thins-walled components are vulnerable to seismic damage[1]. In this paper, the structural integrity assessment are performed to investigate the effect of piping length on creep-fatigue and seismic damage at elevated temperature. L-shaped elbow is considered for piping design and finite element analyses are conducted to calculate creep-fatigue and seismic damage. The evaluation of creep fatigue damage is carried out according to the elevated temperature design codes of ASME B&PV Sec. III Subsec. NH-3200[2]. Seismic damage are evaluated based ASME B&PV Sec. III Subsec. NB-3600[3] and ASME B&PV Sec. III Div.5 HBB-3200[4]. From the results of creep-fatigue and seismic damage, limit length of piping is determined.

Comparison of R5 and ASME NH Creep-Fatigue Damage Assessment Methodologies

2013 ◽  
Author(s):  
Michael Sheridan ◽  
David Knowles ◽  
Oliver Montgomery
Keyword(s):  
Fatigue Damage ◽  
Structural Integrity ◽  
Temperature Response ◽  
Creep Damage ◽  
Basic Principles ◽  
Metallic Structures ◽  
Reference Stress ◽  
Creep Fatigue ◽  
Design Generation

The R5 volume 2/3 procedures [1] were developed by British Energy (now EDF Energy) to assess the high temperature response of uncracked metallic structures under steady state or cyclic loading. They contain the basic principles of: • Application of reference stress methods • Consideration of elastic follow up • A ductility exhaustion approach to calculate creep damage accumulation. These considerations represent a fundamental distinction from ASME BPVC Section III, Subsection NH [2]. This paper draws on literature review and experience to explain the principal differences in the limits of application, cycle construction and damage calculation between these codes/procedures focusing on creep-fatigue damage determination. The implications of the differences between the codes and standards are explored. The output of this work is aimed at two groups of structural integrity engineers; those using these codes and standards to assess existing conventional and nuclear plant, and also those looking to ASME and R5 to design Generation IV PWRs with design temperatures much elevated from those of Generation III and III+. The conclusions from this paper offer some practical guidance to structural integrity engineers which may assist in selecting the more appropriate procedure to assess creep-fatigue damage for a particular situation.

Evaluation method for structural integrity assessment in core disruptive accident of fast reactor

2004 ◽  
Vol 227 (1) ◽  
pp. 97-123 ◽  
Author(s):  
Toshio Nakamura ◽  
Hitoshi Kaguchi ◽  
Iwao Ikarimoto ◽  
Yoshio Kamishima ◽  
Kazuya Koyama ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Evaluation Method ◽  
Structural Integrity ◽  
Integrity Assessment

Structural integrity of fast reactor components

1990 ◽  
Vol 331 (1619) ◽  
pp. 419-434 ◽  
Keyword(s):  
Fast Reactor ◽  
Structural Integrity ◽  
Ambient Pressure ◽  
Failure Process ◽  
Liquid Sodium ◽  
Secondary Circuit ◽  
Deformation And Failure ◽  
Integrity Assessment ◽  
High Boiling Point ◽  
Materials Used

The paper focuses on the generic aspects of the main structural integrity issues in the liquid-sodium-cooled fast reactor. The choice of sodium as a coolant has important consequences for the deformation and failure process in the materials used for the main plant components. For example, its high boiling point means that the prim ary and secondary circuit containment operates at ambient pressure and the system loading is dominated by thermal stress. The resultant low primary stresses make leak-before-break a viable integrity criterion for all sodium boundary components. Sodium coolant operates at comparatively high temperatures and this, together with the good heat-transfer properties, means that thermal fatigue and creep are of concern, particularly in the hotter parts of the plant. A third factor concerns the steam generators, where the integrity of the sodium—water boundary is particularly important. The paper will consider the failure processes that must be addressed in relation to these conditions and the development of the integrity assessment arguments.

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.

Overview of EASICS Weldment Assessment Route Development Through Inelastic Analysis

2020 ◽  
Author(s):  
S. May ◽  
S. Bate ◽  
M. Chevalier ◽  
D. Dean
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Element Analysis ◽  
Validation Data ◽  
Integrity Assessment ◽  
Inelastic Analysis ◽  
Inelastic Material ◽  
Creep Fatigue ◽  
Material Mismatch

Abstract Structural integrity assessment of weldments within metal structures is key to substantiate any nuclear reactor design. The assessment of weldments should consider the localised strain enhancement due to weldment geometry and material mismatch. For high temperature plant designs (operating within the creep regime), R5 Volume 2/3 Appendix A4 provides a procedure for the assessment of creep-fatigue initiation in austenitic and ferritic steel weldments, which accounts for the associated strain enhancement using a Weld Strain Enhancement Factor (WSEF). The current austenitic Type 1 WSEFs in R5 Volume 2/3 have been defined by data attained primarily for plate butt weldments under applied bending loads, and this factor is used for all butt weldments. It has been proposed that the weld strain enhancement may be dependent on loading, geometric and material mismatch conditions, and that adopting a single factor in an assessment may introduce varying levels of conservatism, which are unquantified. This work has included reviewing the current R5 Type 1 WSEF against existing validation data, previous inelastic Finite Element Analysis (FEA) studies and the use of inelastic material models in the FEA of weldments subject to cyclic loading.

Creep-Fatigue Tests and Life Prediction Models for Structural Integrity Assessment of 9Cr-1Mo (Grade 91) Steels

2013 ◽  
Author(s):  
Mauro Filippini ◽  
Riccardo Catelli ◽  
Mihaela E. Cristea
Keyword(s):  
Life Prediction ◽  
Structural Integrity ◽  
Prediction Models ◽  
Martensitic Steel ◽  
Fatigue Tests ◽  
Integrity Assessment ◽  
Creep Fatigue ◽  
Grade 91 ◽  
Temperature Strain ◽  
Calculation Procedures

High temperature strain controlled fatigue and creep-fatigue tests have been carried out on Grade 91 martensitic steel, according to the procedure of the newly developed ASTM E2714 standard. In this paper, the results of these tests are discussed and evaluated according to different damage calculation procedures (e.g. time fraction rule, ductility exhaustion method) for the purpose of validating life prediction models that could be used for the assessment of structural integrity of components in Grade 91 steels operating at high temperatures.

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