scholarly journals Application of the Advanced Surface Modification Process to the ASME Code Case for Sections III and XI of Nuclear Power Plants

Metals ◽  
2020 ◽  
Vol 10 (2) ◽  
pp. 210
Author(s):  
Sungwoo Cho ◽  
Hyun-Uk Hong ◽  
Nicholas Mohr ◽  
Marc Albert ◽  
John Broussard ◽  
...  
Keyword(s):  
Surface Modification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cost Savings ◽  
Performance Criteria ◽  
Laser Peening ◽  
Modification Process ◽  
Asme Code ◽  
American Society

The advanced surface modification process is known as a promising solution to improve the performance of machine components and systems, especially for vehicles, nuclear power plants, biomedical device, etc. There have been several successful applications of water jet peening and underwater laser peening to nuclear components in Japan since 2001 which resulted in inspection and repair cost savings. The prerequisite condition for the application of the advanced surface modification process to nuclear power plants is the approval of the American Society of Mechanical Engineers (ASME) Code Case, so performance criteria and requirements (PCRs) in the ASME Code Case for repair and maintenance of nuclear power components are explained. A challenging project to apply advanced surface modification processes, such as ultrasonic nanocrystal surface modification and air laser peening to new nuclear power plants and new canisters, was created with the goal to develop a technical basis and the PCRs for ASME Section III (New Manufacturing). The results of this work will be an ASME Section III Code Case which is currently in progress. An initial draft of the new Code Case with the intermediate results of this work is introduced. Four kinds of advanced surface modification processes are explained and compared briefly.

Regulatory Perspectives on ASME OM Code, Mandatory Appendix IV, Revision 0, “Inservice Testing of Air-Operated Valves in Light-Water Reactor Nuclear Power Plants”

2014 ◽  
Author(s):  
Jason Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Diagnostic Testing ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Light Water ◽  
Water Reactor ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses and mandates the use of the ASME OM Code for testing air-operated valves in 10 CFR 50.55a(b)(3)(ii) and 10 CFR 50.55a(f)(4), respectively. ASME has recently approved Mandatory Appendix IV, Revision 0. NRC currently anticipates that Mandatory Appendix IV will first appear in the 2014 Edition of the ASME OM Code. Publication of the 2014 Edition of the ASME OM Code begins the NRC rulemaking process to modify 10 CFR 50.55a to incorporate the 2014 Edition of the ASME OM Code by reference. NRC staff has actively participated in the development of Mandatory Appendix IV, Revision 0, through participation in the ASME OM Code Subgroup on Air-Operated Valves (SG-AOV). The purpose of this paper is to provide NRC staff perspectives on the contents and implementation of Mandatory Appendix IV, Revision 0. This paper specifically discusses Mandatory Appendix IV, Sections IV-3100, “Design Review,” IV-3300, “Preservice Test,” IV-3400, “Inservice Test,” IV-3600, “Grouping of AOVs for Inservice Diagnostic Testing,” and IV-3800, “Risk Informed AOV Inservice Testing.” These topics were selected based on input received during NRC staff participation in the SG-AOV and other industry meetings. The goal of this paper is to provide NRC staff perspectives on the topics of most interest to NRC staff and members of the SG-AOV. Paper published with permission.

Regulatory Requirements and Perspectives on the Use of Inservice Testing Software in the Nuclear Industry

2014 ◽  
Author(s):  
Jason B. Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Industry ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Regulatory Requirements ◽  
Operational Readiness ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing (IST) and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses the use of the ASME OM Code in 10 CFR 50.55a(b)(3) . This paper focuses on applicable regulatory requirements and regulatory perspectives associated with the use of IST software in the nuclear industry. Paper published with permission.

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.

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.

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.

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.

Engineering, Design, and Fabrication of Flanges and Reinforced Elbows for an Essential Service Water (ESW) Polyethylene Pipeline

2009 ◽  
Author(s):  
Harvey Svetlik
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Water System ◽  
Steel Pipe ◽  
Raw Water ◽  
Statistical Probability ◽  
Internal Corrosion ◽  
Service Water ◽  
Asme Code

30 years ago, steel pipe was the standard in the ASME code for raw water supply and raw water handling at Nuclear Power Plants. At some power plants, despite best efforts, that steel pipe has suffered intermittent leakage from external and severe MIC internal corrosion. Additionally, internal tuberculation and mineral build-up has severely constricted flow in other pipelines. Advanced, engineering pipe-grade polyethylene pipe has been extruded and used in some nuclear power plants as the effective method to eliminate corrosion and tuberculation of raw water system pipelines. Implementing the change to earth-quake tolerant polyethylene pipelines has resulted in decreased maintenance, increased system reliability, and improved plant longevity. The expectation is that the advanced polyethylene will provide continuous service up to 100 years from initial installation, with a very low statistical probability of any pressure rupture during its service life. Herein discussed is the engineering grade of polyethylene material, its design basis, the conversion of extruded heavy-wall pipe (Picture 1) into fabricated components, and the final production of fully pressure-rated, fabricated pipe fittings with wall thickness of up to 5-inches. Fabrication pictures 1 to 12 are included.

ASME Operation and Maintenance (OM) Code Overview

2016 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Power ◽  
Regulatory Requirements ◽  
General Overview ◽  
Operation And Maintenance ◽  
American Society ◽  
General Organization

The purpose of this paper is to provide a general overview of the organization and content of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code. This will involve a brief description of the regulatory requirements associated with Inservice Testing (IST) as well as a brief overview of the OM Code scope and requirements. This paper will discuss, in general, the regulations requiring IST as well as a brief discussion on when Preservice Testing (PST) and IST become required. A general organization of the ASME OM Code will be provided as well as general topics associated with how to determine when testing and examination intervals are established; what documentation is required; and general discussion regarding the various subsections of the OM Code and the components associated with the OM Code. Alternatives to the OM Code requirements and how to obtain these alternatives will also be provided as well as how the edition applicability of the ASME OM Code is determined. There is also discussion regarding a few general issues associated with the OM Code regarding existing reactor power plants as well as the “new builds” and advanced reactor plants and designs.

Design Strength of Primary Structural Welds in Freestanding Structures

1991 ◽  
Vol 113 (3) ◽  
pp. 471-475
Author(s):  
K. P. Singh ◽  
A. I. Soler ◽  
S. Bhattacharya
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Structural Components ◽  
Rational Analysis ◽  
Free Standing ◽  
Design Strength ◽  
Analysis Technique ◽  
Asme Code

A rational analysis technique to evaluate structural integrity of primary welds in free-standing structures in accordance with the ASME Code is presented. This paper is intended to fill the void in the ASME Code rules for analyzing welds under “faulted” (level D) conditions in nonlinear free-standing structural components used in safety-related applications in nuclear power plants.

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