Structural Integrity and Life Assessment of Pressure Vessels - Risk Based Approach

2020 ◽  
pp. 274-293
Author(s):  
Aleksandar Sedmak ◽  
Snezana Kirin ◽  
Igor Martic ◽  
Lazar Jeremic ◽  
Ivana Vucetic ◽  
...  
Keyword(s):  
Structural Integrity ◽  
Pressure Vessels ◽  
Life Assessment

Advances in Residual Stress Modeling and Measurement for the Structural Integrity Assessment of Welded Thermal Power Plant

2008 ◽  
Vol 41-42 ◽  
pp. 391-400 ◽  
Author(s):  
Lyndon Edwards ◽  
Mike C. Smith ◽  
Mark Turski ◽  
Michael E. Fitzpatrick ◽  
P. John Bouchard
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Residual Stresses ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Economic Life ◽  
Life Assessment ◽  
Integrity Assessment

The safe operation of both thermal and nuclear power plant is increasingly dependent upon structural integrity assessment of pressure vessels and piping. Furthermore, structural failures most commonly occur at welds so the accurate design and remnant life assessment of welded plant is critical. The residual stress distribution assumed in defect assessments often has a deciding influence on the analysis outcome, and in the absence of accurate and reliable knowledge of the weld residual stresses, the design codes and procedures use assumptions that yield very conservative assessments that can severely limit the economic life of some plant. However, recent advances in both the modeling and measurement of residual stresses in welded structures and components open up the possibility of characterising weld residual stresses in operating plant using state-of–the–art fully validated Finite Element simulations. This paper describes research undertaken to predict residual stresses in stainless steel welds in order to provide validated reliable, accurate Structural Integrity assessment of nuclear power plant components

French RPV PTS Assessment: An Overview of EDF Research and Development in Progress on Thermalhydraulic, Materials and Mechanical Aspects

2003 ◽  
Author(s):  
Dominique Moinereau ◽  
Jean-Michel Frund ◽  
Henriette Churier-Bossennec ◽  
Georges Bezdikian ◽  
Alain Martin
Keyword(s):  
Structural Integrity ◽  
Three Dimensional ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Short Description ◽  
Constraint Effect ◽  
Materials Properties ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

Features Testing of Stud Material Under Low Constraint Ductile Tearing

2017 ◽  
Author(s):  
P. James ◽  
M. Jackson ◽  
P. Birkett ◽  
C. Madew
Keyword(s):  
Structural Integrity ◽  
High Stress ◽  
Crack Tip ◽  
Pressure Vessels ◽  
Materials Testing ◽  
Screw Thread ◽  
Surface Point ◽  
Defect Tolerance ◽  
Material Response ◽  
Crack Tip Constraint

Defect tolerance assessments are carried out to support the demonstration of structural integrity for high integrity components such as nuclear reactor pressure vessels. These assessments often consider surface-breaking defects and assess Stress Intensity Factors (SIFs) at both the surface and deepest points. This can be problematic when there is a high stress at the surface, for example due to the stress concentration at the root of a screw thread. In the past this has led to the development of complex and costly 3D finite element analyses to calculate more accurate SIFs, and still resulting in small apparent limiting defect sizes based on initiation at the surface point. Analysis has been carried out along with supporting materials testing, to demonstrate that the increased SIF at the surface point is offset by a reduction in crack-tip constraint, such that the material exhibits a higher apparent fracture toughness. This enables a more simplistic assessment which reduces the effective SIF at the surface such that only the SIF at the deepest point needs to be considered. This then leads to larger calculated limiting defect sizes. This in turn leads to a more robust demonstration of structural integrity, as the limiting defect sizes are consistent with the capability of non-destructive examination techniques. The high SIF at the surface location, and the concomitant reduction in crack-tip constraint, meant that it was not possible to demonstrate the material response with conventional tests, such as those using shallow-notched bend specimens. Instead it was necessary to develop modified specimens in which semielliptical defects were introduced into a geometry which replicated the notch acuity at the root of a screw thread. These feature tests were used to demonstrate the principle, prior to testing with more conventional specimens to fit more accurately the parameters required to represent the material response in a defect tolerance assessment. Margins in defect tolerance assessments are usually measured against the initiation of tearing, even though the final failure for the material may occur at a higher load following stable crack extension. This work measured and assessed the benefit of reduced crack-tip constraint on both the point of initiation and on the development of the tearing resistance curve. This demonstrated that the effect of constraint was valid with tearing for this material and that there was additional margin available beyond the onset of tearing. The feature test geometry also provided evidence of the tearing behaviour at the surface and deepest points of a surrogate component under representative loading. This paper provides an overview of the range of tests performed and the post-test interpretation performed in order to provide the R6 α and k constraint parameters.

scholarly journals Residual Life Assessment of Electricity Pylons – A Case Study

2009 ◽  
Vol 413-414 ◽  
pp. 219-228 ◽  
Author(s):  
John R. Maguire
Keyword(s):  
Structural Integrity ◽  
Residual Life ◽  
Life Assessment ◽  
Original Design ◽  
Integrity Assessment ◽  
Specific Safety ◽  
Residual Life Assessment ◽  
Independent Assessment ◽  
Overhead Power Line

This case study describes a structural integrity assessment of a 220 kV overhead power line. The line comprises 70 pylons over a distance of approximately 30 km, predominantly in a valley location. The pylons are spaced at intervals of approximately 400 m and each pylon is approximately 32 m in height. The line was originally constructed in the 1950’s, approximately 50 years prior to the requested structural integrity assessment. This paper describes the independent assessment that was carried out. The review established site-specific safety factors at the time of original design and construction; at the time of the review (2007), accounting for the possible presence of the “Thomasstahl” steel; and in the future, at the anticipated end of pylon life (in 2012).

ORNL Evaluation of Safety Cases for Two Belgian Reactor Pressure Vessels Containing Quasi-Laminar Defects

2017 ◽  
Author(s):  
Hilda B. Klasky ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
Sarma B. Gorti ◽  
Randy K. Nanstad ◽  
...  
Keyword(s):  
New York ◽  
Finite Element ◽  
Structural Integrity ◽  
National Laboratory ◽  
Relevant Literature ◽  
Pressure Vessels ◽  
Oak Ridge ◽  
Safety Margins ◽  
Safety Cases ◽  
Reactor Pressure

The Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasi-laminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, “Fracture Toughness Criteria for Protection Against Failure,” New York (both 1992 and 2004 versions). That screening assessment of the EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, “Alternative Characterization Rules for Quasi-Laminar Flaws”. Results and conclusions derived from comparisons of the ORNL flaw acceptance assessments of D3/T2 with those from the 2015 EBL Safety Cases are presented in the paper. The ORNL screening analyses identified fewer flaws than EBL that were not compliant with the ASME Section XI (1992) criterion; the EBL criterion imposed additional conservatisms not included in ASME Section XI. Furthermore, ORNL’s application of the updated ASME Section XI (2004) criterion produced only four non-compliant flaws, all due to design-basis loss-of-coolant loading transients. Among the latter, only one flaw remained non-compliant when analyzed using the warm-prestress (WPS) cleavage fracture model typically applied in USA flaw assessments. ORNL’s independent refined analysis of that flaw (#1660, which was also non-compliant in the EBL screening assessments) rendered it compliant when modeled as a more realistic individual quasi-laminar flaw using a 3-dimensional XFEM (eXtended Finite Element Method) approach available in the ABAQUS© finite element code. Taken as a whole, the ORNL-specific results and conclusions confirmed the structural integrity of Doel 3 and Tihange 2 under all design transients with ample margin in the presence of the 16,196 detected flaws.

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