LBB Under Beyond Design Basis Seismic Loading

2012 ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Event ◽  
Jet Impingement ◽  
Seismic Loading ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Piping Systems ◽  
Heavy Water Reactor ◽  
Regulatory Commission

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

Leak-Before-Break Under Beyond Design Basis Seismic Loading

2013 ◽  
Vol 135 (5) ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Event ◽  
Jet Impingement ◽  
Seismic Loading ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Piping Systems ◽  
Regulatory Commission ◽  
Leak Before Break

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970's using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The United States Nuclear Regulatory Commission (US NRC) developed leak-before-break (LBB) procedures in standard review plan (SRP) 3.6.3 Revision 1 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987. The US NRC is currently developing a draft Regulatory Guide on what is called the transition break size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work (Wilkowski et al., “Robust LBB Analysis for Atucha II Nuclear Plant,” 2011 ASME PVP Conference, July 17–21, Baltimore, MD) on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1 × 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94% of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33% around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1 × 10−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

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.

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.

Human Factor Data in Nuclear Power Plant Applications

1980 ◽  
Vol 24 (1) ◽  
pp. 123-123
Author(s):  
Linda O. Hecht
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Reliability Modeling ◽  
Plant Safety ◽  
Modeling Techniques ◽  
Nuclear Systems ◽  
Factor Data ◽  
Regulatory Commission ◽  
The Impact

Due to the concern for safety the nuclear power industry in the United States has fostered the use of reliability analysis to assess system performance and the impact of system failure on overall plant safety. The need for system and component failure rate data has been recognized and has spurred such efforts as NPRDS (Nuclear Power Research Data System) and IEEE's Std 500 (The Reliability Data Manual). Reliability modeling techniques have been developed for application to nuclear systems and are presently being considered by the Nuclear Regulatory Commission for licensing purposes.

Relating ASME Flaw Evaluation to ASME Task Group on Carbon Fiber Repairs

2017 ◽  
Author(s):  
Tomas Jimenez ◽  
Eric Houston ◽  
Nico Meyer
Keyword(s):  
Carbon Fiber ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Fiber Reinforced Polymers ◽  
Structural Factors ◽  
Operating Life ◽  
The Us ◽  
Piping Systems ◽  
Wave Passage ◽  
Regulatory Commission

As most nuclear power stations in the US have surpassed their initial 40 years of operability, the industry is now challenged with maintaining safe operations and extending the operating life of structures, systems and components. The US Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and Electric Power Research Institute (EPRI) have identified safety related buried piping systems as particularly susceptible to degradation. These systems are required to maintain the structural factors of the ASME Construction Codes under pressure and piping loads, which includes seismic wave passage. This paper focuses on evaluation approaches for metallic buried piping that can be used to demonstrate that localized thinning meets the requirements of the Construction Code. The paper then addresses a non-metallic repair option using carbon fiber reinforced polymers (CFRP) as the new pressure boundary.

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.

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).

Preparing for the Extended Loss of AC Power (ELAP) Event in the USA

2014 ◽  
Author(s):  
Joseph S. Miller
Keyword(s):  
Nuclear Power ◽  
Task Force ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Spent Fuel ◽  
Ac Power ◽  
The Us ◽  
External Events ◽  
Near Term ◽  
Regulatory Commission

The United States utilities started preparing for external events that could lead to a loss of all ac power in the 1980’s, when the Station Blackout (SBO) rulemaking was first introduced by the United States Nuclear Regulatory Commission (USNRC). Following the events at the Fukushima Dai-ichi nuclear power plant on March 11, 2011, the USNRC established a senior-level agency task force referred to as the Near-Term Task Force (NTTF). The NTTF was tasked with conducting a systematic, methodical review of Nuclear Regulatory Commission (NRC) regulations and processes to determine if the agency should make additional improvements to these programs in light of the events at Fukushima Dai-ichi. As a result of this review, the NTTF developed a comprehensive set of recommendations, documented in SECY-11-0093, “Near-Term Report and Recommendations for Agency Actions Following the Events in Japan,” dated July 12, 2011. Documentation of the staff’s efforts is contained in SECY-11-0124, “Recommended Actions to be Taken without Delay from the Near-Term Task Force Report,” dated September 9, 2011, and SECY-11-0137, “Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated October 3, 2011. To satisfy some of the NRC’s recommendations, the industry described its proposal for a Diverse and Flexible Mitigation Capability (FLEX), as documented in Nuclear Energy Institute’s (NEI) letter, dated December 16, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11353A008). FLEX was proposed as a strategy to fulfill the key safety functions of core cooling, containment integrity, and spent fuel cooling. The events at Fukushima Dai-ichi highlight the possibility that extreme natural phenomena could challenge the prevention, mitigation and emergency preparedness defense-in-depth layers. At Fukushima, limitations in time and unpredictable conditions associated with the accident significantly challenged attempts by the responders to preclude core damage and containment failure. During the events in Fukushima, the challenges faced by the operators were beyond any faced previously at a commercial nuclear reactor. NRC Order 12-049 (Ref. 1) and NRC Interim Staff Guidance JLD-ISG-2012-01 (Ref. 6) provided additional requirements to mitigate beyond-design-basis external events. These additional requirements impose guidance and strategies to be available if the loss of power, motive force and normal access to the ultimate heat sink to prevent fuel damage in the reactor and spent fuel pool affected all units at a site simultaneously. The NEI submitted document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation Guide” in August 2012 (ADAMS Accession No. ML12242A378) to provide specifications for the nuclear power industry in the development, implementation, and maintenance of guidance and strategies in response to NRC Order EA-12-049. The US utilities are currently proposing modifications to their plants that will follow specifications provided in NEI 12-06. This paper presents some of the NEI 12-06 requirements and some of the proposed modifications proposed by the US utilities.

Development of Design Methods for Piping Systems in a Seismically Base Isolated Nuclear Plant

2016 ◽  
Author(s):  
Timothy M. Adams ◽  
Eun Woo Ahn ◽  
Sungjune Kim ◽  
Sookyum Kim ◽  
Deasoo Kim
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Seismic Event ◽  
Base Isolation ◽  
Design Methods ◽  
Large Displacements ◽  
Plant Design ◽  
Isolation System ◽  
Piping Systems ◽  
Base Isolation System

KEPCO Engineering and Construction is developing a Nuclear Power Plant Design that places the Nuclear Island on a base isolation system while the remaining plant structures are on standard structural foundations. When subjected to seismic (earthquake) loading, this will result in large differential displacements between the nuclear island and the adjacent buildings. Critical piping systems, which must remain functional during and after the seismic event, must withstand the large displacements. This paper summarizes the studies conducted to develop design methods for such piping systems. Also presented are the recommended approaches to be used in the design of such piping systems.

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