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.

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.

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.

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.

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.

Fatigue-Strength-Reduction Factors for Welds in Pressure Vessels and Piping

2000 ◽  
Vol 122 (3) ◽  
pp. 297-304 ◽  
Author(s):  
Carl E. Jaske
Keyword(s):  
Fatigue Strength ◽  
Weld Joint ◽  
Pressure Vessels ◽  
Strength Reduction ◽  
Current Status ◽  
Published Data ◽  
Weld Joints ◽  
Class 1 ◽  
Asme Code ◽  
Reduction Factors

Fatigue-strength-reduction factors (FSRFs) are used in the design of pressure vessels and piping subjected to cyclic loading. This paper reviews the background and basis of FSRFs that are used in the ASME Boiler and Pressure Vessel Code, focusing on weld joints in Class 1 nuclear pressure vessels and piping. The ASME Code definition of FSRF is presented. Use of the stress concentration factor (SCF) and stress indices are discussed. The types of welds used in ASME Code construction are reviewed. The effects of joint configuration, welding process, cyclic plasticity, dissimilar metal joints, residual stress, post-weld heat treatment, the nondestructive inspection performed, and metallurgical factors are discussed. The current status of weld FSRFs, including their development and application, are presented. Typical fatigue data for weldments are presented and compared with the ASME Code fatigue curves and used to illustrate the development of FSRF values from experimental information. Finally, a generic procedure for determining FSRFs is proposed and future work is recommended. The five objectives of this study were as follows: 1) to clarify the current procedures for determining values of fatigue-strength-reduction factors (FSRFs); 2) to collect relevant published data on weld-joint FSRFs; 3) to interpret existing data on weld-joint FSRFs; 4) to facilitate the development of a future database of FSRFs for weld joints; and 5) to facilitate the development of a standard procedure for determining the values of FSRFs for weld joints. The main focus is on weld joints in Class 1 nuclear pressure vessels and piping. [S0094-9930(00)02703-7]

scholarly journals Application of the FAVOR-OCI Fracture Mechanics Computer Program to ASME Code Section XI, IWB-3610 Flaw Acceptance Criteria Evaluations

2018 ◽  
Author(s):  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Critical Values ◽  
Initiation Toughness ◽  
Acceptance Criteria ◽  
Oak Ridge ◽  
The Us ◽  
Asme Code

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Applications of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is maintained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all previously-existing probabilistic capabilities of FAVOR. The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws. The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These compute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected cleavage initiation toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Combination Criterion for Multiple Laminar Flaws in Steel Components

2017 ◽  
Vol 741 ◽  
pp. 63-69
Author(s):  
Valéry Lacroix ◽  
Vratislav Mareš ◽  
Bohumír Strnadel ◽  
Kunio Hasegawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Rolling Direction ◽  
Separation Distance ◽  
Pressure Vessels ◽  
New Combination ◽  
Mechanical Interaction ◽  
Single Plane ◽  
Asme Code ◽  
American Society

A laminar flaw is a planar subsurface flaw parallel to the rolling direction of the plate, where the applied stress is typically parallel to the rolling direction. The laminar flaw oriented within 10 degree of a plane parallel to the component surface is defined as a laminar flaw, in accordance with the definition of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Section XI. The ASME Code provides combination criterion for multiple laminar flaws. If there are two or more laminations, these laminations are projected to a single plane and, if the separation distance of the projected laminations is less than or equal to 25.4 mm, the laminations shall be combined into a single large laminar flaw in the assessment. The combination criterion was established on the basis of the non-destructive examination capabilities in the 1970’s. However, this methodology did not consider the offset distance of the laminations nor the mechanical interaction between the flaws. Therefore that combination methodology is not suited in case of a large number of laminar flaws. This may occur e.g. in case of hydrogen flaking in steel forging components. Actually, when multiple discrete laminar flaws are close to each other, interaction between the flaws has to be taken into account and these flaws shall be combined to a single laminar flaw for assessment. Stress intensity factor interactions for inclined laminar flaws were analyzed in the frame of hydrogen flaking issue in reactor pressure vessels of Doel 3 and Tihange 2 Belgian nuclear power plants. Based on the mechanical interaction between flaws, new combination criterion was developed and was presented in this paper.

Comparison of Design Requirements for Nuclear Steam Generators in the ASME and RCC-M Codes

2013 ◽  
Author(s):  
Junjie Cao ◽  
Peter Caple
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Criteria ◽  
Steam Generators ◽  
Class 1 ◽  
Asme Code ◽  
Historic Development ◽  
Design Requirements ◽  
Plant Components

As two major international codes for class 1 nuclear power plant components, both ASME BPVC Section III and RCC-M Code have been applied to the design of steam generators for nuclear power plants. With respect to the design requirements for steam generators, there are many similarities between the ASME and RCC-M rules as a result of the historic development of the RCC-M Code which evolved from a foundation based on the ASME Code. However, differences do exist in the two codes. In this paper, detailed requirements related to steam generator pressure boundary design in ASME Subsection NB are compared with corresponding requirements in RCC-M Subsection B to identify the differences between the two codes. Supplemental requirements to the ASME Code requirements are proposed for steam generator pressure boundary design to meet the design criteria in both codes. This approach may be also applied to other Class 1 components.

Quantification of the Conservatisms in the Flaw Acceptability Assessment of Doel 3 and Tihange 2 RPVs Containing Hydrogen Flakes

2019 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Assessment Methodology ◽  
Operating Modes ◽  
Safety Case ◽  
Asme Code ◽  
Component Manufacturing ◽  
Accident Conditions ◽  
Reactor Pressure

Abstract During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected 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. One of the most challenging parts of this demonstration was the Flaw Acceptability Assessment, aiming at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This analysis was done by using a methodology: innovative, in line with existing ASME Code Section XI requirements, specific, sufficiently wide to be accepted and, first and foremost, conservative. Through a brief reminder of the Flaw Acceptability Assessment methodology, the paper presents the main hypotheses done for the calculation and quantifies the conservatism related to each of them. This quantification clearly highlights the reliability of final result i.e., the demonstration of the Fitness-for-Service for continued operation of both Doel 3 and Tihange 2 RPVs.

Comparison Between Different Calculation Methods for Determining Bolting Up Moments

2016 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
European Standard ◽  
Calculation Algorithm ◽  
Operational Conditions ◽  
Class 1 ◽  
Asme Code ◽  
Floating Type

There are several different standards for flange calculation used in the European and so in the Suisse context. The European Standard EN 1591-1 that is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined were reissued in 2013 and in 2014, respectively. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components in Suisse nuclear power plants it is also used for class 1 flange connections. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint, different m and y values for different kind of gaskets are invented in ASME BPVC Section III. As cited in the Note of table XI-3221.1-1 the values m and y are not mandatory. In Switzerland, mainly the ASME BPVC should be used for the calculation of flange connections. The aim of the ASME Code is more or less not the tightness of the flanges but the integrity. Therefore, stresses are derived for dimensioning the flanges. Following loads are not considered neither for calculation of stresses nor for calculation of tightness. Considering the experience with flanges in general it could be asked, if it is more useful to look at the tightness than at the stresses. The codes KTA 3201.2 and KTA 3211.2 regulate the calculation of flange connections in German nuclear power plants. Stresses in floating type and in metal-to-metal contact type of flange connections and the tightness are calculated for the different load cases. In this paper, the differences in the calculations are shown between KTA 3211.2, ASME BPVC, Section III, Appendix 11, EN 1591-1 and Finite element calculations. In all load cases leakage shouldn’t occur. Therefore, internal pressure and temperature in test and operational conditions after bolting-up are also considered for the stress calculation if it is possible in the calculation algorithm.

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