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.

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.

Implementation of Regulatory Rules on Fitness-for-Service for Nuclear Power Plants in Japan

2004 ◽  
Author(s):  
Takehiko Nakamura ◽  
Tetsuya Yamamoto ◽  
Iwao Oshima ◽  
Kiyoshi Takasaka ◽  
Yukio Hirano ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Industrial Safety ◽  
Non Destructive Testing ◽  
Destructive Testing ◽  
Class 1 ◽  
Further Development ◽  
Non Destructive

A set of new regulatory rules on nuclear power plants have been implemented in Japan effective October 1, 2003, in order to conduct the Inservice Inspection (ISI) effectively and to examine fitness-for-service properly. The new regulation utilizes the Japan Society of Mechanical Engineer (JSME) Codes on fitness-for-service for nuclear power plants. The rules are applied for the ISIs and for evaluation of structural integrity of the class 1 components and the core shrouds with flaws. Additional components will be subjected to the structural integrity examination, in accordance with the further development of the codes and with establishment of the performance demonstration system for non-destructive testing. Endorsement of other codes and standards are foreseen after technical review by the Nuclear and Industrial Safety Agency (NISA) of Japan. This new regulatory process, which utilizes the consensus codes and standards established by academic/public societies, would allow the NISA to accommodate recent technical progress in a more timely manner.

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.

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.

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.

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.

Comparison of Nuclear Fitness for Service Codes: ASME Section XI and RSE-M AFCEN Codes

2014 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Pressurized Water ◽  
Term Operation ◽  
Mechanical Components ◽  
Water Reactors ◽  
Test Requirements ◽  
Service Inspection

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally. The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs): - Pre-service and In-service inspection - Surveillance in operation or during shutdown - Flaw evaluation - Repairs-Replacements parts for plant in operation - Pressure tests The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments… The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes… These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.

Sign in / Sign up

Export Citation Format

Share Document