Development of an Automatic 3D Finite Element Crack Propagation System

2014 ◽  
Author(s):  
Hiroaki Doi ◽  
Hitoshi Nakamura ◽  
Wenwei Gu ◽  
Hiroshi Okada
Keyword(s):  
Finite Element ◽  
Crack Propagation ◽  
Integral Method ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Element Analysis ◽  
3D Finite Element ◽  
Asme Code ◽  
Virtual Crack Extension Method

When cracks are detected in piping in nuclear power plants during in-service inspections, the crack propagation is usually calculated using approximation formulas of stress intensity factor (SIF) provided in the ASME Code, the JSME Rules or the literature. However, when the crack is detected in complicated-shaped locations in components, finite element analysis (FEA) needs to be used to calculate the SIFs. Accordingly, a method of automatically conducting FEA for crack propagations in nuclear power plants is needed. Therefore, we, the Nuclear Regulation Authority (NRA, Japan) have developed an automatic 3D finite element crack propagation system (CRACK-FEM) for nuclear components. The developed CRACK-FEM uses three methods of SIF calculation: the Virtual Crack Extension Method (VCEM), the Virtual Crack Closure-Integral Method (VCCM) and the Domain Integral Method (DIM). Each method uses different meshes, so users can select a method which uses a suitable mesh for the problem. The software includes a geometry generator to create complicated weld models, and a mesh generator which can deal with interior boundaries formed between different materials. The functions and accuracy of the new software are demonstrated by solving several sample problems involving crack propagation. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Finite Element Reliability Analysis of Steel Containment Vessels with Corrosion Damage

2015 ◽  
Vol 10 (3) ◽  
pp. 527-534 ◽  
Author(s):  
Xiaolei Wang ◽  
◽  
Dagang Lu ◽  
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Reliability Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Damage ◽  
Element Analysis ◽  
Containment Vessel ◽  
Reliability Method

Containment vessels, which contain any radioactive materials that would be released from the primary system in an accident, are the last barrier between the environment and the nuclear steam supply system in nuclear power plants. Assessing the probability of failure for the containment building is essential to level 2 PSA studies of nuclear power plants. Degradation of containment vessels of some nuclear power plants has been observed in many countries, so it is important to study how the corrosion has adverse effects on the capacity of containment vessels. Conventionally, the reliability analysis of containment vessels can be conducted by using Monte Carlo Simulation (MCS) or Latin Hypercube Sampling (LHS) with the deterministic finite element analysis. In this paper, a 3D finite element model of an AP1000 steel containment vessel is constructed using the general-purpose nonlinear finite element analysis program ABAQUS. Then the finite element reliability method (FERM) based on the first order reliability method (FORM) is applied to analyze the reliability of the steel containment vessel, which is implemented by combining ABAQUS and MATLAB software platforms. The reliability and sensitivity indices of steel containment vessels under internal pressure with and without corrosion damage are obtained and compared. It is found that the FERM-based procedure is very efficient to analyze reliability and sensitivity of nuclear power plant structures.

Coalescence Criterion of Part-Through Wall Cracks in Steam Generator Tubes of Nuclear Power Plants

2004 ◽  
Author(s):  
Jeries Abou-Hanna ◽  
Timothy McGreevy ◽  
Saurin Majumdar ◽  
Amit J. Trivedi ◽  
Ashraf Al-Hayek
Keyword(s):  
Finite Element ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Local Instability ◽  
Element Analysis ◽  
Failure Pressure ◽  
Ligament Failure ◽  
Steam Generator Tubes

In scheduling inspection and repair of nuclear power plants, it is important to predict failure pressure of cracked steam generator tubes. Nondestructive evaluation (NDE) of cracks often reveals two neighboring cracks. If two neighboring part-through cracks interact, the tube pressure, under which the ligament between the two cracks fails, could be much different than the critical burst pressure of an individual equivalent part-through crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The coalescence criterion, established earlier for 100% through cracks using nonlinear finite element analyses [1–3], was extended to two part-through-wall axial collinear and offset cracks cases. The ligament failure is caused by local instability of the radial and axial ligaments. As a result of this local instability, the thickness of both radial and axial ligaments decreases abruptly at a certain tube pressure. Good correlation of finite element analysis with experiments (at Argonne National Laboratory’s Energy Technology Division) was obtained. Correlation revealed that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for part-through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments. The study revealed that rupture of the radial ligament occurs at a pressure equal to the coalescence pressure in the case of axial ligament with collinear cracks. However, rupture pressure of the radial ligament is different from coalescence pressure in the case of circumferential ligament, and it depends on the length of the ligament relative to crack dimension.

Verification Analyses for Newly Developed Automatic 3D Finite Element Crack Propagation System

2014 ◽  
Author(s):  
Hiroaki Doi ◽  
Hitoshi Nakamura ◽  
Wenwei Gu ◽  
Do-Jun Shim ◽  
Gery Wilkowski
Keyword(s):  
Finite Element ◽  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
3D Finite Element ◽  
Primary Water ◽  
Axial Crack ◽  
Analysis System

In order to calculate the crack propagation in complicated-shaped locations in components such as weld in penetration structures of reactor pressure vessel of nuclear power plants, an automatic 3D finite element crack propagation system (CRACK-FEM) has been developed by the Nuclear Regulation Authority (NRA, Japan). To confirm the accuracy and effectiveness of this analysis system, a verification analysis was performed. The program used for comparison is PipeFracCAE developed by Engineering Mechanics Corporation of Columbus, which has been used for many crack propagation analyses in various applications. In this paper, the axial crack propagation analysis for primary water stress corrosion cracking (PWSCC) in a steam generator inlet nozzle of a pressurized water reactor (PWR) plant is presented. The results demonstrate that the two codes are in good agreement. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

Application and Experience Feedback of ASME Code Case N-155-2 for RTR Piping in Nuclear Plants

2013 ◽  
Author(s):  
Nicolas d’Udekem ◽  
Philippe Art ◽  
Jacques Grisel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sea Water ◽  
Water System ◽  
Moderate Pressure ◽  
Raw Water ◽  
Cooling Systems ◽  
Piping Systems ◽  
Asme Code

Nowadays, the usefulness of RTR (Reinforced Thermosetting Resin) for pressure retaining equipment does not need further proof: they are lightweight, strong, with low thermal elongation and highly corrosion resistant. The use of RTR piping makes all sense for piping systems circulating raw water such as sea water at moderate pressure and temperature for plants cooling. However, this material is rarely used for safety related cooling systems in nuclear power plants. In Belgium, Electrabel and Tractebel have chosen to replace the existing carbon steel pipes of the raw water system by GRE (Glassfiber Reinforced Epoxy) pipes, in accordance with the Authorized Inspection Agency, applying the ASME Code Case (CC) N-155-2 defining the specifications and requirements for the use of RTR pipes, fittings and flanges. After a challenging qualification process, Class 3 GRE pipes are now installed and operating for raw water cooling systems in two Belgian nuclear units and will soon be installed in a third one. The paper will address the followed qualification processes and the implementation steps applied by Electrabel/Tractebel and relate the overcome obstacles encountered during manufacturing, erection and commissioning of Class 3 GRE piping in order to ensure quality, reliability and traceability required for safety equipment in nuclear power plants.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

2006 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Industrial Safety ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection ◽  
Core Shroud

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

scholarly journals CONCEPTION OF REFERENCE BOOKS OF DEFECTS IS IN EQUIPMENT AND PIPELINES OF POWER AND TRANSPORT SYSTEMS

2020 ◽  
Vol 1 (46) ◽  
pp. 387-404
Author(s):  
Kharytonova L ◽  
◽  
Kutsenko O ◽  
Kadenko I ◽  
◽  
...  
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Transport Systems ◽  
Object Of Study ◽  
Service Inspection ◽  
Element Method

The paper focuses on the one of the persperctive approaches to the increasing of thje safety of Nuclear Power Plants - Flaw Handbook Concept. Object of study - equipment and piping of Nuclear Power Plants. Purpose of study - the description of the Flaw Handbook Concept and the application of the concept for the specific example. Method of the study - numerical procedures of the finite-element method and fracture mechanics. In the modern economics the optimization of the performance and operation of industry and power systems is of the main importance. The Flaw Handbook Concept is considered in the paper. This concept is applied on the nuclear power plants in the leading states with the aim of the optimization of the procedures of in-service inspection and repair. The main steps of the concept are considered and applied for the specific example. An example of Flaw Handbook using is analysed. The results of the paper can be incorporated into the procedures of in-service inspection for the safety-significant equipment and piping. KEYWORDS: FLAW HANDBOOK, BRITTLE FRACTURE, FATIGUE, FINITE-ELEMENT METHOD, SURGE PIPE.

Regulatory Perspectives on ASME OM Code, Mandatory Appendix IV, Revision 0, “Inservice Testing of Air-Operated Valves in Light-Water Reactor Nuclear Power Plants”

2014 ◽  
Author(s):  
Jason Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Diagnostic Testing ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Light Water ◽  
Water Reactor ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses and mandates the use of the ASME OM Code for testing air-operated valves in 10 CFR 50.55a(b)(3)(ii) and 10 CFR 50.55a(f)(4), respectively. ASME has recently approved Mandatory Appendix IV, Revision 0. NRC currently anticipates that Mandatory Appendix IV will first appear in the 2014 Edition of the ASME OM Code. Publication of the 2014 Edition of the ASME OM Code begins the NRC rulemaking process to modify 10 CFR 50.55a to incorporate the 2014 Edition of the ASME OM Code by reference. NRC staff has actively participated in the development of Mandatory Appendix IV, Revision 0, through participation in the ASME OM Code Subgroup on Air-Operated Valves (SG-AOV). The purpose of this paper is to provide NRC staff perspectives on the contents and implementation of Mandatory Appendix IV, Revision 0. This paper specifically discusses Mandatory Appendix IV, Sections IV-3100, “Design Review,” IV-3300, “Preservice Test,” IV-3400, “Inservice Test,” IV-3600, “Grouping of AOVs for Inservice Diagnostic Testing,” and IV-3800, “Risk Informed AOV Inservice Testing.” These topics were selected based on input received during NRC staff participation in the SG-AOV and other industry meetings. The goal of this paper is to provide NRC staff perspectives on the topics of most interest to NRC staff and members of the SG-AOV. Paper published with permission.

Effects of Local Wall Thinning With Crack on Stress Intensity Factor for Pipes Subject to Combined Pressure and Bending

2019 ◽  
Author(s):  
Joy (Xiaoya) Tao ◽  
Lei Zhu
Keyword(s):  
Stress Intensity Factor ◽  
Finite Element ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Limit Load ◽  
Minimum Thickness ◽  
Element Analysis ◽  
Accelerated Corrosion

Abstract At ageing power plants, local thinning of pipework or vessel is unavoidable due to erosion/corrosion or other reasons such as flow accelerated corrosion (FAC) — one of the common degradation mechanisms in pipework of nuclear power plant. Local thinning reduces the structure strength, resulting in crack initiation from the corrosion pit or welding defect when subject to cyclic loading. General practice is to use the minimum thickness of the thinned area to calculate both limit load and stress intensity factor (SIF) in performing Engineering Critical Assessment (ECA) using Failure Assessment Diagram (FAD). Using the minimum thickness is normally overly conservative as it assumes that thinning occurs grossly instead of locally, leading to unnecessary early repair/replacement and cost. Performing cracked body finite element analysis (FEA) can provide accurate values of limit load and SIF, but it is time consuming and impractical for daily maintenance and emergent support. To minimise the conservatisms and provide a guidance for the assessment of locally thinned pipework or vessel using existing handbook solutions, a study was carried out by the authors on the effect of local thinning on limit loads. The study demonstrates that local thinning has significant effect on limit load if the thinning ratio of thinning depth to original thickness is larger than 25%. It concluded that the limit load solutions given in handbooks (such as R6 or the net section method) are overly conservative if using the minimum local thickness and non-conservative if using the nominal thickness. This paper discusses the effect of local thinning on SIFs of internal/external defects using cracked body finite element method (FEM). The results are compared with R6 weight function SIF solutions for a cylinder. A modified R6 SIF solution is proposed to count for the effect of local thinning profile. Along with the previous published paper on limit load it provides comprehensive understanding and guidance for fracture assessment of the local thinned pipework and vessel.

Eddy Current Monitoring Aiming at Evaluation of Crack Propagation in Structural Components of Nuclear Power Plants

2004 ◽  
Vol 2004.3 (0) ◽  
pp. 187-188
Author(s):  
Takashi KASUYA ◽  
Takeshi OKUYAMA ◽  
Hisashi ENDO ◽  
Tetsuya UCHIMOTO ◽  
Toshiyuki TAKAGI ◽  
...  
Keyword(s):  
Crack Propagation ◽  
Eddy Current ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Components ◽  
Current Monitoring
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