Comparison of German KTA and ASME Nuclear Design Codes for Class 1, 2, 3 Components and Piping

2011 ◽  
Author(s):  
Daniel Hofer ◽  
Henry Schau ◽  
Hu¨seyin Ertugrul Karabaki ◽  
Ralph Hill
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Codes ◽  
Nuclear Facility ◽  
Design Standards ◽  
Design Rules ◽  
Design Specifications ◽  
Division 1 ◽  
Class 1 ◽  
Standard Organization

This paper compares the design rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Rules for Construction of Nuclear Facility Components, with German nuclear design standards for Class 1, 2, 3 components and piping. The paper is focused on a comparison of the equations for Design by Analysis and on Piping equations. The ASME Section III Code has been used in combination with design specifications for design of German nuclear power plants. Together with manufacturers, inspectors and power plant owners, the German regulatory authority decided to develop their own nuclear design standards. The current versions being used are from 1992 and 1996. New versions of KTA design standards for pressure retaining components (KTA 3201.2 and KTA 3211.2) are currently under development. This comparison will cover the major differences between the design rules for ASME Section III, Div. 1 and KTA standards 3201.2 and 3211.2 as well as code or standard organization by sections, paragraphs, articles and code development.

A 12 MM/S RMS Screening Vibration Velocity for Pipes Using ANSI-OM3 Standard and Regulatory Design Rules

2005 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Yannick Pons ◽  
Pierre Moussou ◽  
Michae¨l Gaudin
Keyword(s):  
Correction Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Vibration Monitoring ◽  
Vibration Velocity ◽  
Concentrated Mass ◽  
Design Rules ◽  
Velocity Relationship ◽  
Regulatory Design

ASME ANSI-OM3 standard is dedicated to the assessment of piping vibrations for nuclear power plants. It provides an allowable zero-to-peak velocity, which is derived from a stress/velocity relationship, where corrections factors (C1, C2K2, C3, C4 and C5) and an allowable stress σal are introduced. In the ANSI-OM3 standard, the C4 correction factor depends on the pipe layout and on its boundary conditions, and is calculated for a few cases. In a former work, it was proposed to extend this factor to a larger number of pipe setups. Besides, the correction factor C1, which stands for the effect of concentrated mass, is established on a given set-up: a clamped-clamped straight pipe span on its first vibrating mode. C1 is then supposed to be conservative on any piping layout. Finally, allowable velocities derived from the ANSI-OM3 stress/velocity relationship may be very conservative. One way to reduce this conservatism is to introduce regulatory design rules. For a larger set of pipe geometries, a new set of C1 and C4 correction factors are computed using weight and pressure designs. Using these numerical results, allowable velocities can be calculated. Then, we propose here to check if a screening vibration velocity of 12 mm/s rms is fulfilled. For the 181 geometries on 3708, which do not meet the criterion, a seismic design checking is applied. Finally, by this way, 99.7% of the tested geometries, which are supposed to be acceptable with respect to static and seismic designs, display allowable velocities above 12 mm/s rms and the minimum allowable vibration velocity is 11.2 mm/s. This screening vibration velocity of 12 mm/s commonly used for vibration monitoring of piping systems in EDF nuclear power plants is then supported.

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.

Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants

2004 ◽  
Author(s):  
Dilip Bhavnani ◽  
James Annett
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Supply System ◽  
Reactor Vessel ◽  
Original Design ◽  
Pressurized Water ◽  
Class 1 ◽  
Current Design ◽  
Successful Design

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.

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.

Seismic Analysis for Axial Fan Used in Nuclear Power Plant

2014 ◽  
Vol 670-671 ◽  
pp. 1134-1139
Author(s):  
Wen Bing Xu ◽  
Wen Yuan Xiang ◽  
Yong Hong Lv ◽  
Guang Yao Lu
Keyword(s):  
Seismic Performance ◽  
Nuclear Power ◽  
Power Plants ◽  
Natural Frequencies ◽  
Seismic Analysis ◽  
Combination Method ◽  
Vibration Modes ◽  
Design Codes ◽  
Axial Fan ◽  
Quadratic Mode

The seismic performance of the axial fan used in nuclear power plants (NPP) is investigated by using the commercial FEM software ANSYS. In the modal analysis, the first five orders of the natural frequencies of the axial fan is obtained, as well as the corresponding vibration modes. And in the spectrum analysis, the complete quadratic mode combination method is applied to determine the dynamic response of the axial fan under a random seismic excitation. The simulation results show that the seismic performance of the axial fan satisfies the requirement of the RCC-M 2007 design codes.

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.

Development of RI-ISI Program for Krško NPP

2013 ◽  
Author(s):  
Scott Kulat ◽  
Robin Graybeal ◽  
Benjamin Montgomery ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Selection Process ◽  
Degradation Mechanism ◽  
Electrical Power ◽  
Implementation Program ◽  
Class 1 ◽  
Process Risk ◽  
Almost All ◽  
The U.S

Risk-informed methodologies for inservice inspections of safety related piping in nuclear power plants were formally established in mid-1990s in the U.S. Since then, they have been adopted and applied by almost all of the U.S. plants. Nowadays, risk-informed inservice inspection (RI-ISI) is considered to be a standard for the operating plants in the U.S. It was not long before the RI-ISI practice started to be “exported” from the U.S. to other countries. By now, RI-ISI had found its way to a number of European and other countries. Among the recent examples is the Krško Nuclear Power Plant (Krško NPP), a two-loop Westinghouse-designed PWR located in Slovenia and owned by Slovenian and Croatian utilities. Krško NPP finished its third inservice inspection (ISI) interval in July 2012 and initiated implementation of the RI-ISI program at the start of the fourth interval. The process used to develop the RI-ISI program conformed to the methodology described in Electrical Power Research Institute (EPRI) Topical Report TR-112657 and included a degradation mechanism evaluation, consequence analysis and risk characterization for ASME Class 1 and Class 2 piping, as well as an element/examination selection process, risk impact assessment and inspection implementation program development. This paper describes the development of the Krško NPP RI-ISI program and the results of its RI-ISI application. A discussion is, also, provided on some aspects relevant for application of RI-ISI approaches developed in the U.S to plants outside of the U.S.

A Stochastic Model for Piping Failure Frequency Analysis Using OPDE Data

2009 ◽  
Vol 131 (5) ◽  
Author(s):  
X.-X. Yuan ◽  
M. D. Pandey ◽  
J. Riznic
Keyword(s):  
Nuclear Power ◽  
Process Model ◽  
Power Plants ◽  
Probabilistic Models ◽  
Accurate Estimation ◽  
Model Parameters ◽  
Pipe Failure ◽  
Failure Frequency ◽  
Failure Data ◽  
Class 1

The accurate estimation of piping failure frequency is an important task to support the probabilistic risk assessment and risk-informed in-service inspection of nuclear power plants. Although probabilistic models have been reported in the literature to analyze the piping failure frequency, this paper proposes a stochastic point process model that incorporates both a time dependent trend and plant-specific (or cohort) effects on the failure rate. A likelihood based statistical method is proposed for estimating the model parameters. A case study is presented to analyze the Class 1 pipe failure data given in the OPDE Database.

A Suggestion of Improvement Plan for ASME Sec. III Requirements Through the Construction Experience of Korea Standard Nuclear Power Plants: Hydrostatic Test Pressure and PWHT Temperature of P-No.4 Materials

2007 ◽  
Author(s):  
Sanghoon Lee ◽  
Youngho Son ◽  
Jaehoon Lee ◽  
Woosung Kim ◽  
Seogchan Yoon
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reduction Rate ◽  
Design Strength ◽  
Tempering Temperature ◽  
Division 1 ◽  
Test Pressure ◽  
Construction Experience ◽  
Static Head

The minimum hydrostatic test pressure for class 2 and 3 components has been reduced from 1.5 to 1.25 times the Design Pressure in ASME B&PV Sec. III, Division 1 since 1999 addenda. If these requirements are applied to the system hydrostatic test as they are, the minimum hydrostatic test pressure of components and system becomes identical. Therefore it may happen that the test pressure imposed on components installed at low locations in the system exceeds the maximum permissible pressure due to the static head during the system hydrostatic test. PWHT temperature requirements for P-No.4 materials in various Construction Codes, such as ASME B31.1, Sec. I and Sec. VIII, except Sec. III have been unified and the minimum PWHT temperature became 649°C (120°F) since 2004 edition. When considering the mechanical properties of the weld, the minimum PWHT temperature of 593°C or that of P No. 4 materials in Sec. III is too low to reduce hardness and to increase toughness. When PWHT is performed on dissimilar material joints (e.g., between P-No.1 or P-No. 3 and P-No. 4) at 649°C (1200°F) in accordance with Sec. VIII etc., it is possible that the strength of the lower P-No. materials is decreased below the design strength because the PWHT temperature will exceed the tempering temperature of the lower P-No. materials. In this study, the cases of system hydrostatic test in UC-3, 4 units and Steam Generator Nozzle to feedwater pipe joint in Korea Standard Nuclear Power Plants (e.g., UC-3, 4 units, YK-3, 4 units and YK-5, 6 units) were reviewed and analyzed. And then problems of two cases were presented. It is suggested that the minimum system hydrostatic test pressure in Sec. III NC, ND should be decreased by the reduction rate of test pressure for components and the minimum PWHT temperature for P No. 4 materials in Sec. III should be 630°C (1166°F).

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