A Study of the ASME B&PV Code for Pipe Support and Restraint Design Rules From 1974 to 2013

2015 ◽  
Author(s):  
Phillip E. Wiseman ◽  
Gaurav S. Goel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
First Generation ◽  
Support Design ◽  
Design Rules ◽  
Linear Elastic ◽  
American Society ◽  
Support Engineering ◽  
Allowable Stress Design ◽  
Nuclear Applications

The pipe support and restraint design rules are based on the linear elastic analysis outlined in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Subsection NF. The subarticle NF-3300, primarily introduced by replicating the Allowable Stress Design (ASD) rules from the AISC Steel Construction Manual, differs from most of the ASME B&PV Code. Additionally, the Code must fulfill the Nuclear Regulatory Guide requirements, i.e. Regulatory Guide 1.124, which mandates the criteria for pipe support design in nuclear applications. The Code has been edited on multiple occasions to refine the rules for pipe support engineering since the inaugural publication of Subsection NF in the 1973 Winter Addenda. The changes in the design rules in the Code have been incorporated to meet the nuclear rules and regulations along with the general needs of the industry. A review of some of these modifications in the design rules provides a better insight on the subarticle since the institution of the first publication of Subsection NF. The changes in the design rules are essential to highlight since the first generation of commercial nuclear power plants constructed 40 years ago are approaching the end of their design life.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Integrated Pipe Stress Analysis/Support Pattern Selection/Support Design CAE System

1992 ◽  
Author(s):  
M. Chatterjee ◽  
A. Unemori ◽  
A. Kakaria ◽  
D. Jain
Keyword(s):  
Stress Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Automatic Process ◽  
Pattern Selection ◽  
Flow Process ◽  
Support Design ◽  
Pipe System ◽  
Analysis And Design ◽  
Cae System

Abstract This paper describes the organization and features of the AUTO-PIPE CAE System. AUTO-PIPE is a fully integrated software package which allows the User to perform the entire sequence of piping analysis and design in a streamlined work flow process. Major tasks in this automatic process includes: (1) Pipe Stress Analysis (2) Pipe Support Location Optimization (3) Stress Isometric Drawing Generation (4) Pipe Support Pattern Selection and Member Design (5) 3D Interference Detection for Support At the core of the System is the AUTO-PIPE (Relational) Database which contains all static (project-specific) and dynamic (model-specific) data required for all of the major tasks listed above. The AUTO-PIPE CAE System has been used, and is currently being used, for pipe system design for Nuclear Power Plants in Japan to achieve substantial manpower reduction and cost savings.

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.

Qualification of Motor Actuated Valve Assemblies for Safety Related Nuclear Applications: Lessons Learned

2018 ◽  
Author(s):  
Ronald Farrell ◽  
L. Ike Ezekoye
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Decision Processes ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
Missed Opportunities ◽  
Ieee Standards ◽  
Electrical Components ◽  
Nuclear Applications

Safety related valves in nuclear power plants are required to be qualified in accordance with the ASME QME-1 standard. This standard describes the requirements and the processes for qualifying active mechanical equipment that are used in nuclear power plants. It does not cover the qualification of electrical components that are addressed using IEEE standards; however, QME-1 recognizes that both mechanical and electrical components must be qualified when they are interfaced as an assembly. Qualifying both mechanical and electrical valve assemblies can be challenging. Considerable amount of judgment is used when developing the plan to qualify any valve with an electric motor actuator. If the wrong steps are taken in planning the tests, the results from the tests may not be useful thus triggering the need to perform additional tests to comply with QME-1 requirements. This paper presents lessons learned in the process of qualifying valve assemblies to meet QME-1 requirements. The lessons include the decision processes associated with planning and executing valve testing, analysis of the valve assemblies for natural frequency determination, and missed opportunities to capture relevant test data during the tests. Finally, the paper will discuss challenges associated with justifying the tests and extending the results of the tests to cover untested valve assemblies.

A Study of Axial Compression in the ASME B&PV Code for Pipe Supports and Restraints

2016 ◽  
Author(s):  
Phillip E. Wiseman ◽  
Zara Z. Hoch
Keyword(s):  
Axial Compression ◽  
Pressure Vessel ◽  
Slenderness Ratio ◽  
Elastic Buckling ◽  
Allowable Stress ◽  
Linear Elastic ◽  
Division 1 ◽  
Asme Code ◽  
American Society ◽  
Allowable Stress Design

Axial compression allowable stress for pipe supports and restraints based on linear elastic analysis is detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF. The axial compression design by analysis equations within NF-3300 are replicated from the American Institute of Steel Construction (AISC) using the Allowable Stress Design (ASD) Method which were first published in the ASME Code in 1973. Although the ASME Boiler and Pressure Vessel Code is an international code, these equations are not familiar to many users outside the American Industry. For those unfamiliar with the allowable stress equations, the equations do not simply address the elastic buckling of a support or restraint which may occur when the slenderness ratio of the pipe support or restraint is relatively large, however, the allowable stress equations address each aspect of stability which encompasses the phenomena of elastic buckling and yielding of a pipe support or restraint. As a result, discussion of the axial compression allowable stresses provides much insight of how the equations have evolved over the last forty years and how they could be refined.

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.

Using Limit Analysis for Seismic Evaluation of Component Located in Nuclear Power Plants

2002 ◽  
Author(s):  
Petr Zeman
Keyword(s):  
Limit Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Static Load ◽  
Seismic Load ◽  
Plastic Behavior ◽  
Seismic Evaluation ◽  
Perfectly Plastic ◽  
American Society

Using limit analysis for evaluation of the seismic resistance of the components located in NPPs is compared with the standard evaluation method. This comparison is based on the procedure specified in American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III. Subsection NC, version 1992 standard. The limit analysis uses perfectly plastic behavior of the material. The seismic load is restricted when using limit analysis to the pseudo-static load. The possibility of building of more realistic non-linear model including contacts is another advantage of limit analysis. Using limit analysis is the way to move the evaluation method closer to the real collapse load and to reduce conservatism.

scholarly journals Overview of NRC NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves, and Inservice Examination and Testing of Dynamic Restrains (Snubbers) at Nuclear Power Plants

2014 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Regulatory Guidance ◽  
American Society ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) staff issued Revision 2 to NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plant,” to assist the nuclear power plant licensees in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraints (snubbers) inservice examination and testing programs. Since the Revision 1 issuance of NUREG-1482, certain tests and measurements required by earlier editions and addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) have been clarified, updated, revised or eliminated. The revision to NUREG-1482 incorporates and addresses those changes, and includes the IST programs guidelines related to new reactors. The revised guidance incorporates lessons learned and experience gained since the last issue. This paper provides an overview of the contents of the NUREG-1482 and those changes and discusses how they affect NRC guidance on implementing pump and valve inservice testing (IST) programs. For the first time, this revision added dynamic restraint (snubber) inservice examination and testing program guidelines along with pump and valve IST programs. This paper highlights important changes to NUREG-1482, but is not intended to provide a complete record of all changes to the document. The NRC intends to continue to develop and improve its guidance on IST methods through active participation in the ASME OM Code consensus process, interactions with various technical organizations, user groups, and through periodic updates of NRC-published guidance and issuance of generic communications as the need arises. Revision 2 to NUREG-1482 incorporates regulatory guidance applicable to the 2004 Edition including 2005 and 2006 Addenda to the ASME OM Code. Revision 0 and Revision 1 to NUREG-1482 are still valid and may continue to be used by those licensees who have not been required to update their IST program to the 2004 Edition including the 2005 and 2006 Addenda (or later Edition) of the ASME OM Code. The guidance provided in many sections herein may be used for requesting relief from or alternatives to ASME OM Code requirements. However, licensees may also request relief or authorization of an alternative that is not in conformance with the guidance. In evaluating such requested relief or alternatives, the NRC uses the guidelines/recommendations of the NUREG, where applicable. The guidelines and recommendations provided in this NUREG and its Appendix A do not supersede the regulatory requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.55a, “Codes and standards”. Further, this NUREG does not authorize the use of alternatives to, grant relief from, the ASME OM Code requirements for inservice testing of pumps and valves, or inservice examination and testing of dynamic restraints (snubbers), incorporated by reference in 10 CFR 50.55a. Paper published with permission.

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