scholarly journals ASME OM Code Implementation Lessons Learned for the AP1000® Plant

2017 ◽  
Author(s):  
Augustine A. Cardillo ◽  
Natalie M. Rodgers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Federal Regulations ◽  
Plant Design ◽  
Regulatory Requirements ◽  
Implementation Challenges ◽  
Service Testing ◽  
Asme Code ◽  
American Society

The Title 10 of the Code of Federal Regulations (10 CFR) Part 52 process and unique aspects of a passive plant design have presented new challenges for the development and implementation of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements. This paper will discuss lessons learned from the development and implementation of pre-service testing (PST) / in-service testing (IST) program plans for the AP1000® plant, for both international and domestic. Topics to be addressed include the following: • Level of detail in design certification • Treatment of unique passive plant features • Design certification commitments beyond Code requirements • Future regulatory requirements for high-risk non-safety component PST/IST • Implementation challenges for international plants 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.

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.

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.

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.

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.

Lessons Learned in Implementation of Guidelines for Quantifying Support System Initiating Event Frequencies

2008 ◽  
Author(s):  
Rupert A. Weston ◽  
Ashley J. Mossa
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Common Cause Failures ◽  
Areas Of Concern ◽  
Pilot Implementation ◽  
American Society

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

Reconstitution of Reactor Pressure Boundary Components Design Stress Report for License Renewal

2010 ◽  
Author(s):  
Kirby Woods ◽  
Kenneth Thomas
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Cyclic Life ◽  
System Supplier ◽  
Asme Code ◽  
American Society ◽  
Renewal Period ◽  
License Renewal ◽  
Design Stress

The majority of United States Commercial Nuclear Power Plants (CNPP) within the next 10 years will reach the end of their license to operate and can be renewed per the “Atomic Energy Act” of 1954. This act allowed the commission to issue commercial electric power nuclear plants a license to operate (“licensed, but not to exceed 40 years, and maybe renewed upon the expiration of such period, (Chapter 10, Sec. 103(c))).” These CNPP licenses are also governed by the NSSS vendor specification requirements and by the American Society of Mechanical Engineers (ASME) design code standards. This connection is in the form of a stress report that defines the “cyclic life” adequacy criteria for this operational limit of 40 years. The license extension subsequently requires a reconstitution of the initial design stress report input parameters per ASME IWA-4120, 4223 & 4311 (e) for the renewal period extension. This requirement can entail an analysis of the operating conditions and cycles to demonstrate the material elasticity is maintained. The proposed approach for this reconstitution effort was a reanalysis in the form of a study of the Nuclear Steam System Supplier (NSSS) vendors’ original methodology to determine the NSSS vendor specification requirement for ASME code compliance and “cyclic life” adequacy. The information acquired from this evaluation has demonstrated the application to be a complex and simplistic approach. This effort to unravel the composite loading (thermal and pressure transients) condition in relation to specific plant and incident cycles provides both an understanding of component end-of-life limits and supports a comprehensive template for future fatigue life management programs. This paper summarizes this reconstituting effort that utilizes the original vessel stress analysis report to support the license renewal effort, provides a template for future fatigue management programs, demonstrates the conservatism of design, and offers an insight into the design philosophy revealing an elegant process that assures against failures.

Combination Criterion for Multiple Laminar Flaws in Steel Components

2017 ◽  
Vol 741 ◽  
pp. 63-69
Author(s):  
Valéry Lacroix ◽  
Vratislav Mareš ◽  
Bohumír Strnadel ◽  
Kunio Hasegawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Rolling Direction ◽  
Separation Distance ◽  
Pressure Vessels ◽  
New Combination ◽  
Mechanical Interaction ◽  
Single Plane ◽  
Asme Code ◽  
American Society

A laminar flaw is a planar subsurface flaw parallel to the rolling direction of the plate, where the applied stress is typically parallel to the rolling direction. The laminar flaw oriented within 10 degree of a plane parallel to the component surface is defined as a laminar flaw, in accordance with the definition of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Section XI. The ASME Code provides combination criterion for multiple laminar flaws. If there are two or more laminations, these laminations are projected to a single plane and, if the separation distance of the projected laminations is less than or equal to 25.4 mm, the laminations shall be combined into a single large laminar flaw in the assessment. The combination criterion was established on the basis of the non-destructive examination capabilities in the 1970’s. However, this methodology did not consider the offset distance of the laminations nor the mechanical interaction between the flaws. Therefore that combination methodology is not suited in case of a large number of laminar flaws. This may occur e.g. in case of hydrogen flaking in steel forging components. Actually, when multiple discrete laminar flaws are close to each other, interaction between the flaws has to be taken into account and these flaws shall be combined to a single laminar flaw for assessment. Stress intensity factor interactions for inclined laminar flaws were analyzed in the frame of hydrogen flaking issue in reactor pressure vessels of Doel 3 and Tihange 2 Belgian nuclear power plants. Based on the mechanical interaction between flaws, new combination criterion was developed and was presented in this paper.

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.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

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