scholarly journals Characteristics of the Vertical and Horizontal Response Spectra of Earthquakes in the Jeju Island Region

2021 ◽  
Vol 11 (22) ◽  
pp. 10690
Author(s):  
Jun-Kyoung Kim ◽  
Soung-Hoon Wee ◽  
Seong-Hwa Yoo ◽  
Kwang-Hee Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Strong Dependence ◽  
Response Spectra ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Jeju Island ◽  
Ground Acceleration ◽  
Resonance Frequencies ◽  
Earthquake Series

In this study, we evaluated the response spectra of 24 earthquake series, which includes 15 from the Kumamoto earthquake series and 9 from the Pohang earthquake series, and explored the effects of earthquake magnitude on the resonance frequencies of structures and buildings. Furthermore, the observations of this study were compared with the design response spectra, Regulatory Guide 1.60 (The United States Nuclear Regulatory Commission, 1973) for Korean nuclear power plants, and with the Korean Building Code (MOLIT, 2016, hereinafter referred to as KBC 2016) for general structures and buildings. The response spectra, after normalization with reference to the peak ground acceleration (PGA), were derived using a total of 423 horizontal and vertical accelerations. It was observed that the shapes of the horizontal and vertical response spectra were strongly dependent on the magnitude of the earthquake and the resonance frequency. Given the strong dependence of the response on the magnitude, it is suggested to consider magnitude > ML ~ 6.0 when establishing design response spectra. Compared to inland areas, a fairly higher amplitude of response at significantly lower frequency ranges could be attributed to the local geological environment of Jeju Island, which was formed by a surface volcano eruption and the distribution of unconsolidated Pleistocene marine sediments in the Jeju area. It is necessary to study the characteristic influence of layers with low shear wave velocity distributed in the Jeju region on seismic responses more rigorously while considering the frequency band and amplitudes at the surface of Jeju. The resonance frequencies of general low-rise and mid-rise buildings by the brief formula and those by design response spectra both suggested by KBC 2016 were overlapped, and these indicated that the seismic hazard could be much higher on general buildings in the Jeju region than in inland areas. Lastly, it is necessary to make the design standard criteria for Reg. Guide 1.60 and KBC 2016 more conservative in the lower frequency range of higher than 0.6 Hz and 2.0–6.0 Hz, respectively, which is significantly lower than those of the inland area, and to establish improved design response spectra with site-specific seismic design standards by referencing large amounts of qualitative data from the region around the Korean Peninsula.

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 Seismic Issues in Design Certification and Combined License Application Reviews

2009 ◽  
Author(s):  
Jim Xu ◽  
Sujit Samaddar
Keyword(s):  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Response Spectra ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Site Specific ◽  
License Application ◽  
Standard Design

The U.S. Nuclear Regulatory Commission (NRC) established a new process for licensing nuclear power plants under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” which provides requirements for early site permits (ESPs), standard design certifications (DCs), and combined license (COL) applications. In this process, an application for a COL may incorporate by reference a DC, an ESP, both, or neither. This approach allows for early resolution of safety and environmental issues. The COL review will not reconsider the safety issues resolved by the DC and ESP processes. However, a COL application that incorporates a DC by reference needs to demonstrate that pertinent site-specific parameters are confined within the safety envelopes established by the DC. This paper provides an overview of site parameters related to seismic designs and associated seismic issues encountered in DC and COL application reviews using the 10 CFR Part 52 process. Since DCs treat the seismic design and analysis of nuclear power plant (NPP) structures, systems, and components (SSC) as bounding to future potential sites, the design ground motions and associated site parameters are often conservatively specified, representing envelopes of site-specific seismic hazards and parameters. For a COL applicant to incorporate a DC by reference, it needs to demonstrate that the site-specific hazard in terms of ground motion response spectra (GMRS) is enveloped by the certified design response spectra of the DC. It also needs to demonstrate that the site-specific seismic parameters, such as foundation-bearing capacities, soil profiles, and the like, are confined within the site parameter envelopes established by the DC. For the noncertified portion of the plant SSCs, the COL applicant should perform the seismic design and analysis with respect to the site-specific GMRS and associated site parameters. This paper discusses the seismic issues encountered in the safety reviews of DC and COL applications. Practical issues dealing with comparing site-specific features to the standard designs and lessons learned are also discussed.

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).

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.

Regulatory and Industry Guidance on Use of Risk Technology in Nuclear Power Plant Applications

2007 ◽  
Author(s):  
Mansoor H. Sanwarwalla
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Regulatory Body ◽  
Current Application ◽  
Technical Specifications ◽  
Regulatory Commission ◽  
The Impact

Since the United States Nuclear Regulatory Commission (USNRC) published its landmark “Reactor Safety Study — An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants” in late 1975, commercial nuclear power industry, encouraged by the USNRC, have since then been applying Probabilistic Risk Assessment (PRAs) in their nuclear power units in areas of in-service testing, in-service inspection, quality assurance, technical specifications, maintenance, etc. To guide and regulate the industry in use of PRAs, Regulatory Guides and Standards have been written and are being revised continuously by the USNRC, American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS). The current use of PRA takes credit for single failure criterion based on applicability of codes and standards. The proposed new USNRC regulation 10 CFR Part 53 applicable for all reactor technologies is silent on the applicability of current standards endorsed by the regulatory body. The impact of the proposed new rule to both new and the current application needs to be studied. This paper will review the application of the various guidance documents for their use in commercial nuclear power plants with emphasis on the new generation nuclear power plants.

Reactor Vessel Closure Head Drop Analysis: Sensitivity Study on the Effects of Representing Nonlinear Behavior in the Closure Head Assembly

2009 ◽  
Author(s):  
William C. Castillo ◽  
Joseph M. Remic ◽  
George J. Demetri ◽  
Frank J. Marx ◽  
David H. Roarty
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Sensitivity Study ◽  
Reactor Vessel ◽  
Air Cooling ◽  
Vessel Closure ◽  
Head Assembly

Nuclear power plants need to safely and efficiently remove their reactor vessel closure head assembly during plant outages. This is accomplished by lifting the closure head assembly out of the reactor vessel cavity and placing it on the closure head stand. In order for nuclear power plants to remove their closure head assembly, the United States Nuclear Regulatory Commission has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis [1]. The goal of head drop analyses is to qualify the maximum drop height in air per plant procedures. A significant percentage (greater than 30%) of the closure head assembly’s mass is comprised of components attached to the top of the head (such as: lifting fixtures, a missile shield, air cooling systems, and control rod drive housings). The analytical consideration of large deflection, plastic deformation, and local failure of these components can potentially change the energy imparted to the vessel during impact due to their energy-absorbing capacities during the drop event. This paper contains a sensitivity study to determine the benefits of modeling closure head assembly components, using nonlinear structural behavior. The guidelines of Nuclear Energy Institute Initiative NEI 08-05 [2] are followed for this study.

scholarly journals Spectral Matching of Three-Component Seismic Ground Accelerations for Nuclear Power Plants

2021 ◽  
Author(s):  
Sindur Mangkoesoebroto ◽  
Ediansjah Zulkifli ◽  
Adi P. Yasa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Response Spectra ◽  
Baseline Correction ◽  
Fourier Amplitude ◽  
Spectral Matching ◽  
Ground Acceleration ◽  
Good Convergence ◽  
Target Spectrum ◽  
Phase Spectra

Abstract The aim of the paper is to introduce a new procedure of three-component spectral matching of seismic ground acceleration records. The procedure is straightforward, yet it is general. In principle, the procedure involves varying of both the Fourier amplitude and the phase spectra so that the modified records’ spectra agree with a target. The matching can be performed against either a target Fourier or response spectra. In the former the solution is exact, while in the latter it becomes approximate. A target spectrum representative of three directions should be provided. In the example several three-component records were matched against two target spectra. Good convergence was achieved in velocity and displacement records so that no baseline correction was necessary. The couplings among the components were preserved.

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