Inservice Examination and Testing Issues for Dynamic Restraints (Snubbers) in Nuclear Power Plants

2017 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Regulatory Commission ◽  
The U.S

This paper discusses recent issues related to the inservice examination and testing of dynamic restraints (snubbers) at U.S. nuclear power plants. The U.S. Nuclear Regulatory Commission (NRC) staff identified these issues during its review of examination and testing snubber programs and relief requests, as well as operating experience. This discussion includes information that could apply generically to the implementation of effective snubber programs at U.S. nuclear power plants. Paper published with permission.

scholarly journals Inservice Examination and Testing Issues for Dynamic Restraints (Snubbers) in Nuclear Power Plants

2014 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Regulatory Commission ◽  
The U.S

This paper discusses recent issues related to inservice examination and testing of dynamic restraints (snubbers) at U.S. nuclear power plants. These issues were identified during the U.S. Nuclear Regulatory Commission (NRC) staff review of snubber examination and testing programs, relief requests, and applicable operating experience. This discussion includes information that could have generic applicability in the implementation of effective snubber programs at U.S. nuclear power plants. 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.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

US Nuclear Regulatory Commission Perspective on Buried and Underground Piping Issues at Nuclear Power Plants in the United States

2014 ◽  
Author(s):  
David Alley
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Action Plan ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Buried Pipe ◽  
Development Organizations ◽  
Regulatory Commission ◽  
License Renewal

This paper provides a historical perspective on the need for, and development of, buried and underground piping tanks programs at nuclear power plants. Nuclear power plant license renewal activities, Nuclear Regulatory Commission Buried Piping Action Plan, and the rationale for addressing the issue of buried pipe through an industry initiative as opposed to regulation are discussed. The paper also addresses current NRC activities including the results of Nuclear Regulatory Commission inspections of buried piping programs at nuclear power plants as well as Nuclear Regulatory Commission involvement in industry and standards development organizations. Finally, the paper outlines the Nuclear Regulatory Commission’s future plans concerning the issue of buried piping at US nuclear power plants.

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.

Long Term Operation of Nuclear Power Plants: Status in the U.S.

2013 ◽  
Author(s):  
Garry G. Young
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Safe Operation ◽  
Term Operation ◽  
The Status ◽  
Regulatory Commission ◽  
License Renewal ◽  
The U.S

As of January 2013, the U.S. Nuclear Regulatory Commission (NRC) has renewed the operating licenses of 73 nuclear units out of a total of 104 licensed units, allowing for up to 60 years of safe operation. In addition, the NRC has license renewal applications under review for 15 units and more than 13 additional units have announced plans to submit applications over the next few years [1]. This brings the total of renewed licenses and plans for renewal to over 97% of the 104 operating nuclear units in the U.S. This paper presents the status of the U.S. license renewal process and issues being raised for possible applications for subsequent renewals for up to 80 years of operation. By the end of 2013 there will be 26 nuclear plants in the U.S. (or 25% of the 104 units) that will be eligible to seek a second license renewal and by the end of 2016 this number will increase to about 50% of the 104 licensed units. Although some nuclear plant owners have announced plans to shutdown before reaching 60 years, the majority are keeping the option open for long term operation beyond 60 years. The factors that impact decisions for both the first license renewals and subsequent renewals for 80 years of safe operation are presented and discussed in this paper.

Accident Sequence Precursor Analysis of Daya Bay and Lingao Nuclear Power Plants

2010 ◽  
Author(s):  
Pingping Liu ◽  
Haiying Xi
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Daya Bay ◽  
Probabilistic Safety Assessment ◽  
Regulatory Commission ◽  
Probabilistic Safety

Firstly this article describes how to analyze accident sequence precursor in U.S. nuclear regulatory commission. Following this method, the licensee operational events, internal operational events and inspection reports of Daya Bay and Lingao Nuclear Power Plants were reviewed to identify the precursors with support of probabilistic safety assessment. All the identified precursors were calculated, documented and ranked. Then the trends on precursors can be obtained. Finally this article analyzes the trends available and gains many beneficial insights.

scholarly journals Inservice Testing Program Improvements for New Reactors: Small Modular Reactors

2014 ◽  
Author(s):  
Ronald 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, including more traditional evolutionary designs, passive reactor designs, and small modular reactors (SMRs). ASME (formerly the American Society of Mechanical Engineers) provides specific codes used to perform inspections and testing, both preservice and inservice, 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 certification (DC) and combined license (COL) applications under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” of Title 10, “Energy,” of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code, Operation and Maintenance of Nuclear Power Plants, 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 after January 1, 2000. The ASME New Reactors OM Code (NROMC) Task Group (TG) is assigned the task of ensuring that the preservice testing (PST) and inservice testing (IST) provisions in the ASME OM Code are adequate to provide reasonable assurance that pumps, valves, and dynamic restraints (snubbers) for post-2000 plants will operate when needed. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in nonsafety systems for passive post-2000 plants, including SMRs. (Note: For purposes of this paper, “post-2000 plant” and “new reactor” are used interchangeably throughout.) Paper published with permission.

scholarly journals Combustible Performance of Floor Finishing Materials and Coating Applied to Nuclear Power Plants

2020 ◽  
Vol 20 (4) ◽  
pp. 139-144
Author(s):  
Inkyu Kwon
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Construction Materials ◽  
Nuclear Regulatory Commission ◽  
Plant Industry ◽  
Testing Methods ◽  
Regulatory Commission

Finishing construction materials applied to nuclear power plants and other attached structures are manufactured domestically; however, their fire-related performance has not yet been clarified, and data exist only for common materials with general purposes. Finishing construction materials must meet the requirement of Nuclear Regulatory Commission (NRC), which is regarded as a global standard in the nuclear power plant industry. In this study, to support data when a new guideline for evaluation of fire safety in nuclear power plant and the attached structures thereof are prepared, the finishing materials on the floor and the coating applied onto the floor and other portions were selected and tested using related standards of two nations. The results showed that there were differences in the manner of evaluation and testing. Moreover, certain criteria did not meet Korean standards. Nevertheless, most criteria were satisfied with testing methods suggested by the NRC.

Sign in / Sign up

Export Citation Format

Share Document