The Application of Combination of Individual Modal Responses According to the Revised RG1.92

2010 ◽  
Author(s):  
Shenghua Liu ◽  
Zufeng Xia
Keyword(s):  
Nuclear Power ◽  
Parametric Design ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Combination Method ◽  
Response Analysis ◽  
Comparison Result ◽  
Combination Methods ◽  
Regulatory Commission ◽  
Double Sum

The U.S. Nuclear Regulatory Commission (NRC) issued the revised regulatory guide 1.92 as Rev. 2 on July 2006, which is believed to efficiently reduce the unnecessary conservative which is introduced by combination methods specified in RG1.92 Rev.1 and to be acceptable for combining modal responses and spatial components in seismic response analysis of nuclear power plant structures, systems, and components (SSCs). But it is still impossible to use this method to do the combination of spectrum response analysis, because most popular general finite element programs do not develop a reasonable command or macros according to RG1.92 Rev.2. In order to perform the combination of individual modal responses according to this new guide conveniently, this paper developed a macro file used the ANSYS Parametric Design Language (APDL) based on ANSYS program which is widely used in nuclear industry. This paper examined the macro file with an actual support pipe, and compared the results got from the new method with the results acquired from previously accepted method such as double sum method and grouped method. The comparison result shows that the revised combination method actually reduced the unnecessary conservative previously accepted by method in RG1.92, Rev.1, also the results prove that the macro file this paper established is reasonable.

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.

TITRAM Fast Transient Licensing Analysis Methodology for Kuosheng BWR/6

2010 ◽  
Author(s):  
Jan-Ru Tang ◽  
Hon-Chin Jien ◽  
Yang-Kai Chiu ◽  
Cheng-Der Wang ◽  
Julian S. C. Chian
Keyword(s):  
Nuclear Power ◽  
Transient Analysis ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Evaluation Methodology ◽  
Uncertainty Evaluation ◽  
Power Company ◽  
Fast Transients ◽  
Fast Transient ◽  
Regulatory Commission

This paper presents the TITRAM (TPC/INER Transient Analysis Method) methodology for the fast transient analysis of Kuosheng Nuclear Power Station (KSNPS) with two units of General Electric (GE) designed BWR/6 (Boiling Water Reactor). The purpose of this work is to provide a technical basis of Taiwan Power Company (TPC)/Institute of Nuclear Energy Research (INER)’s qualification to perform plant specific licensing safety analyses for the Final Safety Analysis Report (FSAR) basis system fast transients, and related plant operational transient analyses for the Kuosheng plant. The major task of qualifying TITRAM as a licensing method for BWR transient analysis is to adequately quantify its analysis uncertainty. A similar approach as the CSAU (Code Scaling, Applicability, and Uncertainty Evaluation) methodology developed by the USNRC (United States Nuclear Regulatory Commission) was adopted. The CSAU methodology could be characterized as three significant processes, namely code applicability, transient scenario specification and uncertainty evaluation based on Phenomena Identification and Ranking. The applicability of the TITRAM code package primarily using the SIMULATE-3 and RETRAN-3D codes are demonstrated with analyses of integral plant tests such as Peach Bottom Turbine Trip Test and plant startup tests of KSNPS. A Phenomena Identification and Ranking Table (PIRT) with uncertainty values for each identified parameter to cover 95% of possible values are established for the selected KSNPS fast transients. The experience from BWR organizations in the nuclear industry is used as a guide in construction of the PIRT. Sensitivity studies and associated statistical analyses are performed to determine the overall uncertainty of fast transient analysis with TITRAM based on the KSNPS Analysis Nominal Model. Finally, the Licensing Model is established for future licensing applications.

A Comparative Study of Combination Methods Used in Response Spectrum Analysis of Nuclear Piping Systems

1984 ◽  
Vol 106 (1) ◽  
pp. 25-31 ◽  
Author(s):  
S. Gupta ◽  
D. P. Jhaveri ◽  
O. Kustu ◽  
J. A. Blume
Keyword(s):  
Spectrum Analysis ◽  
Response Spectrum ◽  
Nuclear Regulatory Commission ◽  
Dynamic Responses ◽  
Combination Method ◽  
Response Spectrum Analysis ◽  
Combination Methods ◽  
Piping Systems ◽  
Regulatory Commission ◽  
The U.S

The different methods of combining responses for individual modes and directions for response spectrum analysis of nuclear piping systems are evaluated. For the purpose of the study, dynamic responses of 20 typical piping systems using nine different combination methods are systematically compared. The study established the relative conservatism in design achieved by each method and the probability that a particular combination method will produce a more conservative estimate of seismic response than obtained using one of the methods accepted by the U.S. Nuclear Regulatory Commission.

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.

Overview of Current Practice for Analysis and Design of Independent Spent Fuel Storage Installation Pads

2019 ◽  
Author(s):  
Gunup Kwon ◽  
Khaled Ata
Keyword(s):  
Nuclear Power ◽  
Seismic Analysis ◽  
Storage System ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Loading Conditions ◽  
Spent Fuel ◽  
Analysis And Design ◽  
License Application ◽  
Regulatory Commission

Abstract Nuclear power plant spent fuels are initially stored in the spent fuel pool. Then, the water cooled fuels are transferred in a concrete or steel cask and transported outside of the Fuel Handling Building (FHB) or the Reactor Building (RB) for long term on site storage. The spent fuel casks are typically stored on a slab-on-grade pad. The slab-on-grade pad is designed according to the U.S. Nuclear Regulatory Commission NUREG-1536 and NUREG-1567. The two Standard Review Plans provide guidance to the regulators for the review of cask storage system license application. The ISFSI pad analysis and design have to consider various loading conditions, such as earthquake and tornado loadings as well as normal operating loading conditions. Seismic analysis of the ISFSI pad requires considering interaction between the pad and the supporting soil. Various cask loading configurations on the pad also have to be considered. Due to the lack of specific guidelines, many ISFSI pad designs show overly conservative reinforcement. This study provides guidelines and procedure for the design of the ISFSI pad that are typically used in the nuclear industry. It is considered that the guidelines and practices described in this study help design engineers understand general guidance provided in the NRC Standard Review Plans.

Operational Assessment of Simulator Fidelity in the Nuclear Industry

1988 ◽  
Vol 32 (11) ◽  
pp. 705-709
Author(s):  
Christopher Plott ◽  
Jerry Wachtel ◽  
K. Ronald Laughery
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Objective Data ◽  
Control Room ◽  
Paper Briefly ◽  
Power Plant Control ◽  
Regulatory Commission ◽  
Behavioral Perspective ◽  
Plant Control

The Nuclear Regulatory Commission (NRC) has recently developed a procedure for inspecting nuclear power plant control room simulators. These inspections will ensure that the simulator has sufficient fidelity to produce an appropriate medium for the conduct of operator licensing examinations. The difficulty in obtaining objective data for the assessment of fidelity, particularly for transient or accident events, requires that the inspection be performed primarily from a behavioral perspective rather than a strictly engineering perspective. This paper briefly describes the procedure developed.

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.

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.

Proposed License Structure of Small Modular Reactor in China

2018 ◽  
Author(s):  
Zheng Hua ◽  
Wei Shuhong
Keyword(s):  
Nuclear Power ◽  
Modular Design ◽  
Nuclear Regulatory Commission ◽  
Light Water Reactor ◽  
Research Progress ◽  
Structure Characteristics ◽  
Common Structure ◽  
Small Modular Reactor ◽  
Regulatory Commission ◽  
Design Construction

Small Modular Reactor (SMR) is getting more and more attention due to its safety and multi-purpose application. License structure is an important issue for SMR licensing. Modular design, construction and operation, shared or common structure, system and components (SSC) challenge existing large light water reactor license structure. Existing nuclear power plant license structure, characteristics of SMR and its effect on license structure, and research progress of U.S Nuclear Regulatory Commission (NRC) are analyzed, SMR license structure in China are proposed, which can be used as a reference for SMR R&D, design and regulation.

Development of Code Case N-755: Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities

2008 ◽  
Author(s):  
Frank J. Schaaf
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Regulatory Review ◽  
Plastic Pipe ◽  
Service Water ◽  
Division 1 ◽  
Minimum Requirements ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Applications

With the increasing failures of metallic pipe in nuclear Service Water Systems, a new material needed to be found. One option is polyethylene (PE) pipe. PE pipe can be used in non-safety applications at a nuclear plant using the American Society of Mechanical Engineers (ASME) B31, Standards of Pressure Piping with no regulatory review. However, the use of PE material in safety applications, which are regulated by the Nuclear Regulatory Commission (NRC), necessitates a new Standard with special requirements. At the request of the Duke Power Corporation, a new ASME Standard was written by a special Project Team. This standard is found in the form of a Code Case under the control of the ASME Boiler & Pressure Vessel Code (B&PVC). The Code Case utilizes Sections of the B&PVC as its foundation and includes the design, procurement, installation, fusing, examination and testing requirements for the use of PE pipe within safety systems. The first version of the Code Case contained only the minimum requirements needed to support Duke Power Corporation’s first phase of PE piping installation into a safety system within a nuclear power plant. The Code Case developed is titled, N-755, Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities. The first version of this case is limited to buried piping using only the following components; straight PE pipe, PE mitered elbows, and transition flanges. The Code Case will be revised as data for material and components becomes available at the completion of testing.

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