Fatigue Life Estimation for Internal Threads in Class 1 Components

1998 ◽  
Vol 120 (1) ◽  
pp. 81-85 ◽  
Author(s):  
R. Kumar
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Simplified Approach ◽  
Fatigue Life Estimation ◽  
Class 1 ◽  
Inspection And Maintenance ◽  
Service Conditions ◽  
Service Inspection

Heat exchangers, steam generators, and other pressure vessels in nuclear power plants are equipped with bolted closures for the purpose of in-service inspection and maintenance. The ASME Boiler and Pressure Vessel Code specifies that all Class 1 components meet the fatigue life requirements for level A and B service conditions. In the case of bolted closures, it is often found that the bolt/stud is the most critical part. In many situations, the bolts fail to meet the fatigue requirements for the design life of the equipment. In such cases, the bolts can be replaced after certain duration based upon their fatigue life. However, the mating threads in the flange (which is an integral part of the vessel) are still a concern. While the replacement of the bolts is relatively easy and inexpensive, the corrective action (e.g., replacement or repair) for the flange is usually difficult and expensive, or impossible. Hence, it is important to have a reasonable estimate of the fatigue life of internal threads to alleviate or minimize the concern. In this paper, a simplified approach is presented for this purpose. Considering various bolt sizes, commonly used thread series, and typical Class 1 component materials, it is shown that the fatigue life of the internal threads is about three times the fatigue life of the bolt threads. This conclusion greatly reduces or eliminates the concern for in-service replacement or repair of the components with internal threads.

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.

Current Status of Japanese Code on Fitness-for-Service for Nuclear Power Plants

2004 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Current Status ◽  
Japan Society ◽  
The Third ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection

Following a recognition of the need to establish a FFS (Fitness-for-Service) Code in Japan, JSME (Japan Society of Mechanical Engineers) published its first edition 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, which also referred to the ASME Code Section XI incorporating independent Japanese concepts. In addition, individual inspection rules for specific structures, such as shroud and shroud support for BWR plants, were prescribed in consideration of aging degradation by SCC. Furthermore, the third edition, which includes requirements on repair and replacement methods, will be published in 2004. Along with the efforts of the JSME on the preparation of the FFS Code, the Japanese Regulatory Agency has approved and endorsed this Code as the national rule, which has been in effect since October 2003.

Flaw Evaluation and Allowable Flaw Sizes of Nuclear Piping in JSME Code

2003 ◽  
Author(s):  
Kunio Hasegawa ◽  
Hideo Kobayashi ◽  
Koichi Kashima
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Japan Society ◽  
Acceptance Criteria ◽  
Evaluation Code ◽  
Evaluation Procedures ◽  
Class 1 ◽  
American Society ◽  
Service Inspection

A flaw evaluation code for nuclear power plants has been developed at the Japan Society of Mechanical Engineers (JSME) in 2000 and revised adding inspection rules in 2002. Then the code consists of inspection for nuclear components and evaluation procedures of flaws in Class 1 components detected during in-service inspection. This paper introduces the summary of the JSME Code and describes two kinds of allowable flaw sizes, Acceptance Standards and Acceptance Criteria, for Class 1 pipes in the flaw evaluation procedures. In addition, these allowable flaws are compared with those in the ASME (American Society of Mechanical Engineers) Code Section XI.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs: Fatigue Crack Growth Analysis of Hydrogen Flakes

2017 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Simplified Approach ◽  
Integrity Assessment ◽  
The Core

During the 2012 outages at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the Reactor Pressure Vessels (RPVs). The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated. The Structural Integrity Assessment of the RPVs, through the Flaw Acceptability Assessment, aimed at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This demonstration has been done on the basis of a specific methodology inspired by the ASME B&PV Code Section XI procedure but adapted to the nature and the number of indications found in the Doel 3 and Tihange 2 RPVs. As requested by Article IWB-3610(a) of ASME B&PV Code Section XI, one of the parts that have to be addressed through the Flaw Acceptability Assessment is the Fatigue Crack Growth (FCG) Analysis of the flaws in the core shells until the end-of-service lifetime of the RPVs. Due to the large number of flaws in the core shells, a specific methodology has been developed in order not to perform the FCG Analysis of each flaw separately. The paper describes this simplified approach aiming at distributing the flaws according to their inclination and at defining envelope flaws covering the actual flaws to carry out FCG Analysis. Furthermore, the paper highlights and quantifies the conservatisms of this analysis leading finally to demonstrate that the FCG of hydrogen flakes is not a concern in Doel 3 and Tihange 2 RPVs.

New Needs of Fracture Mechanic Analysis at Design and Operation Level: Status of French Nuclear Mechanical Codes

2018 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Fracture Mechanic ◽  
Pressure Vessels ◽  
Regulatory Requirements ◽  
Mechanic Analysis ◽  
Term Operation ◽  
Fracture Mechanic Analysis ◽  
Class 1 ◽  
Leak Before Break ◽  
Service Inspection

Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture. All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage: - brittle and ductile analysis of hypothetical large flaw - leak before break approach - break exclusion concept - incredibility of failure of high integrity components - end of fabrication acceptable defect - in-service inspection performance - acceptable standards in operation - Long Term Operation (LTO) All these requirements needs a procedure, an analysis method with material properties and criteria. After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.

scholarly journals Importance of Advanced Planning of Manufacturing for Nuclear Industry

2016 ◽  
Vol 7 (2) ◽  
pp. 42-49
Author(s):  
Nick Shykinov ◽  
Robert Rulko ◽  
Dariusz Mroz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Regulatory Compliance ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Project Schedule ◽  
Nuclear Facilities ◽  
Project Completion ◽  
Advanced Planning ◽  
And Performance

Abstract In the context of energy demands by growing economies, climate changes, fossil fuel pricing volatility, and improved safety and performance of nuclear power plants, many countries express interest in expanding or acquiring nuclear power capacity. In the light of the increased interest in expanding nuclear power the supply chain for nuclear power projects has received more attention in recent years. The importance of the advanced planning of procurement and manufacturing of components of nuclear facilities is critical for these projects. Many of these components are often referred to as long-lead items. They may be equipment, products and systems that are identified to have a delivery time long enough to affect directly the overall timing of a project. In order to avoid negatively affecting the project schedule, these items may need to be sourced out or manufactured years before the beginning of the project. For nuclear facilities, long-lead items include physical components such as large pressure vessels, instrumentation and controls. They may also mean programs and management systems important to the safety of the facility. Authorized nuclear operator training, site evaluation programs, and procurement are some of the examples. The nuclear power industry must often meet very demanding construction and commissioning timelines, and proper advanced planning of the long-lead items helps manage risks to project completion time. For nuclear components there are regulatory and licensing considerations that need to be considered. A national nuclear regulator must be involved early to ensure the components will meet the national legal regulatory requirements. This paper will discuss timing considerations to address the regulatory compliance of nuclear long-lead items.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Lifetime Management of EDF PWR Pressure Vessels and Pipings

2004 ◽  
Author(s):  
F. Hedin ◽  
J. C. Legendre
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cost Benefit Analysis ◽  
Cost Benefit ◽  
Pressure Vessels ◽  
Destructive Testing ◽  
Key Points ◽  
Cast Stainless Steel ◽  
Non Destructive

Lifetime management of EDF PWR vessels and pipings are one of the main technical key points of safety and competitivness. This paper describes the EDF global approach in this field, which is applied to the nuclear fleet i.e 58 nuclear power plants, and particularly to the first 34 three loops, as far as lifetime is concerned: • operating procedures and routine maintenance, special maintenance and ten years safety reassessment, • engineering analysis, based on feed back experience, scientific knowledge, degradations mechanisms, causes and consequences management, • operating loadings decrease, • complementary deterministic and cost-benefit analysis, • fit for service justifications, • anticipation strategy to prepare future, based on Non Destructive Testing investigations, ability to repair and/or to replace components, in situ expertises, ... Some examples are given: lifetime management of reactor vessels heads and bottom penetrations of pressure vessels, fit for service of cast stainless steel primary pipings, primary nozzles and auxiliary pipings special maintenance.

Failure Probability Assessment of Piping Systems Affected by SCC and Pipe Wall Thinning for LBB Analysis

2011 ◽  
Author(s):  
Hiromasa Chitose ◽  
Hideo Machida ◽  
Itaru Saito
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Probability Assessment ◽  
Wall Thinning ◽  
Piping Systems ◽  
Leak Before Break ◽  
Thinning Rate ◽  
Service Inspection

This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Sign in / Sign up

Export Citation Format

Share Document