Supplementary Technical Basis for ASME Section XI Code Case N-597-2

2006 ◽  
Author(s):  
Douglas A. Scarth ◽  
Kunio Hasegawa ◽  
Lee F. Goyette ◽  
Phil Rush
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Code Design ◽  
Acceptance Criteria ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Class 1 ◽  
American Society ◽  
Technical Basis

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provides rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Code Case N-597-2 of Section XI, Requirements for Analytical Evaluation of Pipe Wall Thinning, provides evaluation procedures and acceptance criteria to justify continued operation of Class 1, 2 and 3 piping items subjected to wall thinning by a mechanism such as flow-accelerated corrosion. The acceptance criteria ensure that margins equivalent to those of the ASME B&PV Code are maintained. The technical basis for Code Case N-597-2 was previously presented at the 1999 ASME Pressure Vessels and Piping Conference. Since then, the ASME Section XI Working Group on Pipe Flaw Evaluation has identified the need for further explanation of the technical basis for the Code Case, such as the procedures for evaluation of wall thickness less than the Construction Code Design Pressure-based minimum allowable wall thickness, tmin. This paper provides an additional description of the Code Case technical basis and validation against experimental and historic wall thinning events.

Supplementary Technical Basis for ASME Section XI Code Case N-597

2005 ◽  
Author(s):  
Doug Scarth
Keyword(s):  
Wall Thickness ◽  
Structural Integrity ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Original Design ◽  
Acceptance Criteria ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Technical Basis ◽  
Flow Accelerated Corrosion

Efforts to develop clear and conservative methods to measure and evaluate wall thinning in nuclear piping have been underway since the late 1980’s. The Electric Power Research Institute (EPRI) carried out a successful campaign to address programmatic issues, such as locating and predicting flow-accelerated corrosion (FAC) degradation. This included developing a computer code (CHECWORKS), a users group (CHUG), and a comprehensive program guideline document for the effective prediction, identification and trending of flow-accelerated corrosion degradation. U.S. Nuclear Regulatory Commission (NRC) guidelines are provided in the NRC Inspection Manual Inspection Procedure 49001. At the same time, committees under Section XI of the ASME Boiler and Pressure Vessel Code have addressed evaluation of structural integrity of piping subjected to wall thinning. Code Case N-480 of Section XI provided acceptance criteria that focused on primary piping stresses, with evaluation based on a uniform wall thinning assumption for evaluating the minimum wall thickness of the piping. However, when applying this methodology to low pressure piping systems, Code Case N-480 was very conservative. Code Case N-597 was first published in 1998, and supercedes Code Case N-480. The current version is N-597-2. Code Case N-597-2 provides acceptance criteria and evaluation procedures for piping items, including fittings, subjected to a wall thinning mechanism, such as flow-accelerated corrosion. Code Case N-597-2 is a significant improvement over N-480, containing distinct elements to be satisfied in allowing the licensee to operate with piping degraded by wall thinning. The Code Case considers separately wall thickness requirements and piping stresses, and maintains original design intent margins. The Code Case does not provide requirements for locations of inspection, inspection frequency or method of prediction of rate of wall thinning. As described in the original technical basis document published at the 1999 ASME PVP Conference, the piping stress evaluation follows very closely the Construction Codes for piping. Five conditions related to industry use of Code Case N-597-1 have been published by the NRC in Regulatory Guide 1.147, Revision 13. A number of these issues are related to a need for additional explanation of the technical basis for the Code Case, such as the procedures for evaluation of wall thickness less than the ASME Code Design Pressure-based minimum allowable wall thickness. This presentation addresses these NRC conditions by providing additional description of the technical basis for the Code Case.

Technical Basis for Revisions to ASME Section XI Code Case N-597-2 on Requirements for Analytical Evaluation of Pipe Wall Thinning

2015 ◽  
Author(s):  
Douglas A. Scarth ◽  
Michael Davis ◽  
Phil Rush ◽  
Steven X. Xu
Keyword(s):  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Acceptance Criteria ◽  
Accelerated Corrosion ◽  
Evaluation Procedures ◽  
Technical Requirements ◽  
Wall Thinning ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Flow Accelerated Corrosion

Code Case N-597-2 provides procedures and acceptance criteria for the evaluation of piping items subjected to wall thinning mechanisms such as flow-accelerated corrosion (FAC). The acceptance criteria ensure that margins equivalent to those of the ASME B&PV Code are maintained. Subsequent to the publication of Code Case N-597-2, the U.S. Nuclear Regulatory Commission (NRC) found the Code Case conditionally acceptable. A number of task items have been undertaken by the ASME Section XI Working Group on Pipe Flaw Evaluation (WGPFE) to address the NRC conditions. A 2006 ASME Pressure Vessels and Piping (PVP) Division conference paper was published to provide an expanded explanation of the technical basis for Code Case N-597-2. A 2009 PVP paper was published to provide results of validation of evaluation procedures and acceptance criteria in Code Case N-597-2 against experimental and historic wall thinning events. More recently, revisions to Code Case N-597-2 have been made and were proposed as N-597-3. Significant changes have been made in the proposed revised Code Case to clarify the technical requirements and address the NRC concerns over N-597-2. The technical basis for revising Code Case N-597-2 is provided in this paper.

Assessment of Remaining Life for Class 1 Piping Components Experiencing Wall Thinning

2004 ◽  
Author(s):  
Rajnish Kumar
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Extension ◽  
Remaining Life ◽  
Original Design ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Class 1 ◽  
Plant Components

Assessment of remaining life of power plant components is important in light of plant life management and life extension studies. This information helps in planning and minimizing plant outages for repairs and refurbishments. Such studies are specifically important for nuclear power plants. Nuclear Safety Solutions Limited (NSS) is involved in conducting such studies for plant operators and utilities. Thickness measurements of certain piping components carrying fluids at high temperature and high pressure have indicated higher than anticipated wall thinning rates. Flow accelerated corrosion (FAC) has been identified as the primary mechanism for this degradation. The effect of FAC was generally not accounted for in the original design of the plants. Carbon steel piping components such as elbows, tees and reducers are prone to FAC. In such cases, it is important to establish the remaining life of the components and assess their adequacy for continued service. Section XI of the ASME Boiler and Pressure Vessel Code is applicable for evaluation of nuclear power plant components in service. This Section of the Code does not specifically deal with wall thinning of the piping components. Code Case N-597 provides guidelines for evaluation for continued service for Class 2 and Class 3 piping components. For Class 1 piping components, this Code Case suggests that the plant owner should develop the methodology and criteria for evaluation. This paper presents methodology and procedure for establishing the remaining life and assessment of Class 1 piping components experiencing wall thinning effects. In this paper, the rules of NB-3600 and NB-3220 and Code Case N-597 have been utilized for assessment of the components for continued service. Details of various considerations, criteria and methodology for assessment of the remaining life and adequacy for continued service are provided.

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.

Numerical Calculation of Shear Stress Distribution on the Inner Wall Surface of CANDU Reactor Feeder Pipe Conveying Two-Phase Coolant

2007 ◽  
Author(s):  
Jong Chull Jo ◽  
Dong Gu Kang ◽  
Kyung Wan Roh
Keyword(s):  
Shear Stress ◽  
Stress Distribution ◽  
Wall Thickness ◽  
Shear Stress Distribution ◽  
Cfd Analysis ◽  
Two Phase ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Local Areas ◽  
Flow Accelerated Corrosion

Two-phase flow fields inside feeder pipes of a CANDU reactor have been simulated numerically using a CFD (computational fluid dynamics) code to calculate the shear stress distribution which is the most important factor to be considered in predicting the local areas of feeder pipes highly susceptible to FAC (flow-accelerated corrosion)-induced wall thinning. The CFD approach with schemes used in this study to simulate the turbulent flow situations inside the CANDU feeder pipes had been verified by showing a good agreement between the investigation results for the failed feedwater pipe at Surry Unit 2 plant in U.S. and the CFD calculation. Sensitivity studies of the three geometrical parameters such as angle of the 1st and 2nd bends, length of the 1st span between the grayloc hub and the 1st bend, and length of the 2nd span between the 1st and the 2nd bends had already been performed. In this study, the effects of void fraction of the primary coolant coming out from the exit of pressure tubes containing nuclear fuels on the fluid shear stress distribution at the inner surface of feeder pipe wall have been investigated to find out the local areas of feeder pipes conveying two-phase coolant, where are highly susceptible to FAC (flow-accelerated corrosion)-induced wall thinning. As the results of CFD analysis, it is seen that the local regions of feeder pipes of the operating CANDU reactors in Korea, on which the wall thickness measurements have been performed so far, are not coincided with the worst regions predicted by the present CFD analysis where is the connection region of straight & bend pipe near the inlet part of the bend intrados. Finally, based on the results of the present CFD analysis a guide to the selection of the weakest local positions where the measurement of wall thickness should be performed with higher priority has been provided.

Pressure Blow Out Assessment for Class 1 Piping With Local Wall Thinning

2006 ◽  
Author(s):  
Kunio Hasegawa ◽  
Katsuyuki Shibata
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Tensile Load ◽  
Evaluation Methodology ◽  
Evaluation Procedure ◽  
Seismic Load ◽  
Analytical Evaluation ◽  
Wall Thinning ◽  
Bending Load ◽  
Class 1

Wall thinning caused by the flow of water in power piping systems became a major concern to the nuclear power industries. ASME Code Case N-597-3, “Requirements for Analytical Evaluation of Pipe Wall Thinning,” provides procedures and criteria for Code Class 2 and 3 piping for the evaluation of wall thinning. However, analytical evaluation procedure for Class 1 piping is not provideed in the Code Case. Recent full-scale experiments on locally thinned pipes have supported the development of more contemporary failure strength evaluation methodology for Class 1 piping. These evaluation methodologies are applicable for the loading type of bending, tensile or cyclic bending load. Prior to the failure by bending moment, tensile load or cyclic/seismic load, locally wall thinned pipes shall be considered pressure blow out by the internal pressure itself. This paper introduces the failure of a uniformly thinned cylinder by an internal pressure and describes limitation on wall thinning depth to avoid pressure blow out for Class 1 piping.

Evaluation of Flange Calculations Using Strain-Based Acceptance Criteria

2019 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elastic Stress ◽  
Calculation Methods ◽  
Acceptance Criteria ◽  
Class 1 ◽  
Torque Moment ◽  
Calculation Algorithms ◽  
Calculation Code

Abstract The Nuclear Power Plant KKG in Gösgen, Switzerland was designed according to the ASME Boiler and Pressure Vessel Code. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components, it is also used for class 1 flanges. 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 y and m values for different kinds of gasket are invented in ASME BPVC Section III [1]. The KTA 3201.2[2] and KTA 3211.2[3] regulate the calculation of bolted flanged joints in German nuclear power plants. The gasket characteristics required for these calculation methods are based on DIN 28090-1[4], they can be determined experimentally. In Europe, the calculation code EN 1591-1 [5] and the gasket characteristics according to EN 13555[6] are used for flange calculations. Because these calculation algorithms provide not only a stress analysis but also a tightness proof, it would be preferable to use them also in the NPP’s in Switzerland. Additionally, for regulatory approval also the requirements of the ASME BPVC must be fullfilled. For determining the bolting up torque moment of flanges several tables for different nominal diameters of flanges using different gaskets and different combinations of bolt and flange material were established. As leading criteria for an allowable state, the gasket surface pressure, the allowable elastic stress of the bolts and the strain in the flange should be a good and conservative basis for determining allowable torque moments. The herein established tables show only a small part according to a previous paper [7] where different calculation methods for determining bolting up moments were compared to each other. In this paper the bolting-up torque moments determined with the European standard EN 1591-1 for the flange, are assessed on the strain-based acceptance criteria in ASME BPVC, Section III, Appendices EE and FF. The assessment of the torque moment of the bolts remains elastically which should lead to a more conservative insight of the behavior of the flanges.

Effects of Flow Accelerated Corrosion on Piping Steady-State Vibration Evaluations

2003 ◽  
Author(s):  
Brian J. Voll
Keyword(s):  
Steady State ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Vibration Monitoring ◽  
Accelerated Corrosion ◽  
Monitoring Programs ◽  
Wall Thinning ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Piping steady-state vibration monitoring programs were implemented during preoperational testing and initial plant startup at most nuclear power plants. Evaluations of piping steady-state vibrations are also performed as piping and component failures attributable to excessive vibration are detected or other potential vibration problems are detected during plant operation. Additionally, as a result of increased flow rates in some piping systems due to extended power uprate (EPU) programs at several plants, new piping steady-state vibration monitoring programs are in various stages of implementation. As plants have aged, pipe wall thinning resulting from flow accelerated corrosion (FAC) has become a recognized industry problem and programs have been established to detect, evaluate and monitor pipe wall thinning. Typically, the piping vibration monitoring and FAC programs have existed separately without interaction. Thus, the potential impact of wall thinning due to FAC on piping vibration evaluations may not be recognized. The potential effects of wall thinning due to FAC on piping vibration evaluations are reviewed. Piping susceptible to FAC and piping susceptible to significant steady-state vibrations, based on industry experience, are identified and compared. Possible methods for establishing links between the FAC and vibration monitoring programs and for accounting for the effects of FAC on both historical and future piping vibration evaluations are discussed.

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.

Development of Analytical Evaluation Procedures and Acceptance Criteria for Pipe Wall Thinning in ASME Code Section XI

2004 ◽  
Author(s):  
Kunio Hasegawa ◽  
Gery M. Wilkowski ◽  
Lee F. Goyette ◽  
Douglas A. Scarth
Keyword(s):  
Pipe Wall ◽  
Evaluation Methodology ◽  
Current Status ◽  
Evaluation Procedure ◽  
Analytical Evaluation ◽  
Acceptance Criteria ◽  
New Class ◽  
Wall Thinning ◽  
Class 1 ◽  
Asme Code

As the worldwide fleet of nuclear power plants ages, the need to address wall thinning in pressure boundary materials becomes more acute. The 2001 ASME Code Case N-597-1, “Requirements for Analytical Evaluation of Pipe Wall Thinning,” provides procedures and criteria for the evaluation of wall thinning that are based on Construction Code design concepts. These procedures and criteria have proven useful for Code Class 2 and 3 piping; but, they provide relatively little flexibility for Class 1 applications. Recent full-scale experiments conducted in Japan and Korea on thinned piping have supported the development of a more contemporary failure strength evaluation methodology applicable to Class 1 piping. The ASME B&PV Code Section XI Working Group on Pipe Flaw Evaluation has undertaken the codification of new Class 1 evaluation methodology, together with the existing Code Case N-597-1 rules for Class 2 and 3 piping, as a non-mandatory Appendix to Section XI. This paper describes the current status of the development of the proposed new Class 1 piping acceptance criteria, along with a brief review of the current Code Case N-597-1 evaluation procedure in general.

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