Time Limited Aging Analyses Requirements for License Renewal for U.S. Nuclear Power Plants

2004 ◽  
Author(s):  
Mansoor H. Sanwarwalla
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Management Program ◽  
Metal Fatigue ◽  
Aging Effects ◽  
Electric Equipment ◽  
Design Calculations ◽  
License Renewal

One of the requirements for license renewal for US nuclear plants stated in the United States Nuclear Regulatory Commission (USNRC) regulations in the License Renewal Rule (LRR) 10CFR Part 54 (Ref. 1) is the identification and updating of Time-Limited Aging Analyses (TLAA). During the design phase for a plant, certain assumptions about the length of time the plant would be operated were made and incorporated into design calculations for several of the plant’s systems, structures and components (SSCs). Examples of TLAAs are analyses of metal fatigue, environmental qualification (EQ) of electric equipment, etc. For a renewed license, these calculations have to be reviewed to verify that they remain valid for the period of extended operation. However, the LRR does allow TLAA-associated aging effects to be managed by an aging management program. This paper discusses the USNRC regulatory requirements for TLAAs and the industry’s response for addressing the TLAAs. It also discusses the issues regarding the generic set of TLAAs that have been identified by the NRC in NUREG-1801 (Ref. 2), and how these have been addressed by all the plants that have received their renewed license. The paper also identifies certain plant specific TLAAs.

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.

License Renewal Demonstration Project

2002 ◽  
Author(s):  
Barry J. Elliot ◽  
Jerry Dozier
Keyword(s):  
United States ◽  
State Of The Art ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Management Program ◽  
Lessons Learned ◽  
Demonstration Project ◽  
Aging Effects ◽  
Regulatory Commission ◽  
License Renewal

Generic Aging Lessons Learned (GALL) report, License Renewal Standard Review Plan (SRP-LR), and regulatory guide were issued by the United States Nuclear Regulatory Commission (NRC) in June 2001. The intent of these documents was to provide the technical and process basis that will lead to a more effective, efficient and predictable license renewal process for industry and the NRC. The GALL report provides the aging effects on components and structures, identifies the relevant existing plant programs, and evaluates the program attributes to manage aging effects for License Renewal. The GALL report also identifies when existing plant programs would require further evaluation for License Renewal. The SRP-LR allows the applicant to reference the GALL report to demonstrate that the programs at the applicant’s facility correspond to those reviewed and approved in the GALL report. Programs that correspond to those in the GALL report will not need further detailed review by the staff. Implementation of the aging management program are verified as part of the license renewal inspection program. The GALL report identifies one acceptable way of demonstrating that components and structures have adequate aging management programs. However, applicants may propose alternatives to the programs identified in GALL. During the license renewal review, the NRC primarily focuses on areas where existing programs should be augmented or new programs developed for License Renewal. This paper will provide an overview of these documents and some of the lessons learned during a demonstration project in the application of the new guidance. This topic will be of interest to the U.S. participants considering License Renewal and desiring to know state-of-the-art information about License Renewal in the United States.

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.

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.

Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Stephen A. Hambric ◽  
Samir Ziada ◽  
Richard J. Morante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Cycle Fatigue ◽  
Structural Integrity ◽  
Fatigue Cracking ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Boiling Water ◽  
Cycle Fatigue ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

License Renewal in the U.S. and Plant Life Management (PLIM)

2003 ◽  
Author(s):  
Garry G. Young ◽  
Mark A. Rinckel
Keyword(s):  
United States ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Nuclear Industry ◽  
Life Management ◽  
Technical Requirements ◽  
Plant Life ◽  
License Renewal

License renewal of operating nuclear power plants in the United States has become one of the most successful nuclear industry activities in the past few years. It is anticipated that over 90% of the 103 operating nuclear power plants in the United States will pursue license renewal and seek an additional 20 years of operation. Some plants may pursue operation to 80 years or longer since the license renewal rule does not limit the operating life of a nuclear power plant. The requirements for renewing the operating license of a nuclear reactor in the United States are contained in Nuclear Regulatory Commission (NRC) Regulation 10 CFR Part 54, which addresses general, technical, technical specification, and environmental requirements. The most labor intensive element of the requirements are the technical requirements, which include addressing an integrated plant assessment (IPA) and time limited aging analyses (TLAA). The cost of performing the needed reviews and obtaining a renewed license ranges between $10M to $15M. The license renewal rule focuses on aging of passive long-lived components and aging management programs that manage those structures and components. The aging management programs credited to manage aging include both existing programs (e.g., ASME Section XI) and a few new programs (e.g., Reactor Vessel Internals Aging Management Program). Commitments to aging management programs for license renewal may be implemented and tracked through a comprehensive plant life management (PLIM) program. PLIM is the process to integrated equipment aging management with other plant activities to maximize plant value. PLIM can save the operating plant a significant amount of money by effectively planning and implementing component refurbishment and replacement. The ultimate decision to seek license renewal and continue operation is based on PLIM, which considers aging, safety, and economics.

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.

License Renewal Program for Entergy’s Fleet of Nuclear Power Plants

2004 ◽  
Author(s):  
Garry G. Young
Keyword(s):  
Renewal Process ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Virtual Team ◽  
The United States ◽  
Nuclear Industry ◽  
Dedicated Team ◽  
License Renewal ◽  
Renewal Project

License renewal of operating nuclear power plants in the United States has become one of the most successful nuclear industry activities in the past few years. Entergy, the second largest nuclear plant operator in the U.S., was one of the pioneers in this new process. In 2000, Entergy submitted a license renewal application to the Nuclear Regulatory Commission (NRC) for Arkansas Nuclear One, Unit 1. By June 2001, less than 17 months later, the NRC issued a renewed license. Due in part to the efficiency and success of this first Entergy license renewal project, a dedicated team of Entergy and Areva (formerly known as Framatome-ANP) personnel was established using virtual office concepts to work on license renewal for the remaining nine Entergy nuclear units over the next decade. Since each license renewal project takes 4 to 5 years and costs $10 to $15 million to complete [1], the dedicated team has focused on improving the schedule and economics of the license renewal process. By early 2004, the dedicated team has worked on five license renewal projects and expects to begin work on at least two additional projects by 2005. The virtual team organization has established standardized processes for managing data and for performing aging management reviews, environmental reviews, and time limited aging analyses evaluations. In addition, the team has worked with the Nuclear Energy Institute (NEI) and the NRC to further improve the efficiency of the license renewal process. This paper discusses the standardized processes established, the virtual team techniques used to manage multiple license renewal projects, and the plans for further process improvements. The ultimate goal of Entergy’s license renewal work is to achieve a highly efficient and effective license renewal process that ensures the safe continued operation of its fleet of nuclear power plants for decades to come.

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.

Construction of Westinghouse AP1000™ Nuclear Power Plants in China

2010 ◽  
Author(s):  
Grenville Harrop ◽  
Bill P. Poirier
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Republic Of China ◽  
Environmental Consequences ◽  
Containment Vessel ◽  
Plant Safety ◽  
Major Equipment

In July 2007, China entered a new era of sustainable, safe, and ecologically sound energy development by committing to build four AP1000™ units to be constructed in pairs at the coastal sites of Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Both sites have the planned ability to accommodate at least six AP1000 units. The Westinghouse AP1000 is the only Generation III+ reactor to receive design certification from the U.S. Nuclear Regulatory Commission (NRC). With a design that is based on the proven performance of Westinghouse-designed pressurized water reactors (PWRs), the AP1000 is an advanced 1100 megawatt (MW) plant that uses the forces of nature and simplicity of design to enhance plant safety and operations. Excavation commenced for the first of four China AP1000 units in February 2008, and placement of the basemat concrete for Sanmen Unit 1 was completed on schedule in March 2009. This was soon followed by the placement of the first major structural module; the auxiliary building. As part of localization and the Peoples Republic of China (PRC) desire for self-reliance, a China-based module factory is constructing the major modules and manufacturing the containment vessel plates. The fabrication and welding of the containment vessel bottom head for Sanmen Unit 1 is now complete. The 2010 milestones for Sanmen Unit 1 include the setting of major modules such as the reactor vessel cavity, the steam generator, and refueling canal modules, plus containment vessel rings 1, 2, 3, and 4. All major equipment orders have been placed and the first deliveries are beginning to arrive. The technology transfer is also well underway. The Haiyang Unit 1 basemat was placed on schedule in September 2009 and Sanmen Unit 2 Nuclear Island (NI) concrete basemat placement was completed a month earlier than the milestone date of January 2010. Sanmen Unit 1 will be fully operational in November 2013 followed by Haiyang Unit 1 in May 2014. Operational dates for Sanmen Unit 2 and Haiyang Unit 2 are September 2014 and March 2015, respectively. As one of the world’s largest consumers of energy, China’s path in achieving sustainable energy has profound global economic and environmental consequences. The contract with the Westinghouse and Shaw Consortium for four AP1000 units is the largest of its type between the People’s Republic of China and the United States.

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