Load Combination Reduction Methodology for the US EPR™ Standard Nuclear Power Plant

2012 ◽  
Author(s):  
Se-Kwon Jung ◽  
Joseph Harrold ◽  
Nawar Alchaar
Keyword(s):  
Reinforced Concrete ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Reduction Process ◽  
Shear Walls ◽  
Related Structure ◽  
Design Component ◽  
Load Combination ◽  
The U.S

The Safety-Related structures of the U.S. EPR™ Standard Nuclear Power Plant (NPP) predominantly consist of reinforced concrete shear walls and slabs; thus they are typically modeled using shell finite elements and analyzed and designed for a large number of applicable load combinations. This paper presents a load combination reduction methodology that has been specifically developed for and applied to these types of structural elements in order to methodically reduce the full set of applicable load combinations to a manageable sub-set of load combinations, termed “controlling load combinations” for structural design purposes. Load combination reduction criteria involve code-specified section capacities (i.e., allowables), structural demands (i.e., forces and moments), and demand-to-capacity ratios (DCR) as complemented by reinforcing ratios. For a particular Safety-Related structure or portions thereof, the controlling load combination produces the most demanding forces and moments relative to design allowables in accordance with applicable codes and standards for reinforced concrete design, resulting in the highest DCR among all applicable load combinations. To facilitate the load combination reduction process, portions or segments of a particular Safety-Related structure that are in close proximity and thereby most likely to be designed for a common reinforcement pattern are identified and grouped as a single design component and termed an “evaluation level component.” It is demonstrated that the load combination reduction methodology developed herein is instrumental in narrowing down numerous applicable load combinations to a sub-set of controlling load combinations for the U.S. EPR™ Nuclear Island Safety-Related structures.

Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant

2010 ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joe Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.

Validation and Benchmarking of Comprehensive Vibration Assessment Program for Prototype Reactor Internals

2012 ◽  
Author(s):  
Jianfeng Yang ◽  
Lixin Yu ◽  
Byounghoan Choi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Boiling Water ◽  
Pressurized Water ◽  
Water Reactor ◽  
Assessment Program ◽  
Water Reactors ◽  
The U.S ◽  
Reactor Internals

Reactor internals important to nuclear power plant safety shall be designed to accommodate steady-state and transient vibratory loads throughout the service life of the reactor. Operating experience has revealed failures of reactor internals in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) due to flow-induced vibrations (FIVs). U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) that the NRC staff considers acceptable for use in verifying the structural integrity of reactor internals for FIV prior to commercial operation. A CVAP supports the NRC reviews of applications for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50, as well as design certifications and combined licenses that do not reference a standard design under 10 CFR Part 52. The overall CVAP should be implemented in conjunction with preoperational and initial startup testing. For prototype reactor internals, the comprehensive program should consist of a vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. Validation and benchmarking processes should be integrated into the CVAP throughout each individual program. Based on the authors’ experiences in Advanced Boiling Water Reactor and AP1000® CVAPs and based on detailed reviews of the U.S. Evolutionary Power Reactor and the U.S. Advanced Pressurized Water Reactor CVAPs, this article summarizes the essential CVAP validation and benchmarking processes with proper consideration of bias errors and random uncertainties. This article provides guidance to a successful CVAP that satisfies the NRC requirements and ensures the reliability of the evaluation of potential adverse flow effects on nuclear power plant components.

Assessment of Aging of Nuclear Power Plant Civil Structures

2002 ◽  
Author(s):  
D. J. Naus ◽  
B. R. Ellingwood ◽  
H. L. Graves
Keyword(s):  
Reinforced Concrete ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Reliability ◽  
Operating Experience ◽  
Assessment Tools ◽  
Background Information ◽  
Civil Structures ◽  
Age Related

Research is being conducted by ORNL for the USNRC to address aging of civil structures in light-water reactor plants. The importance and operating experience of nuclear power plant (NPP) civil structures is reviewed. Factors that can lead to age-related degradation of reinforced concrete structures and containment metallic pressure boundaries (i.e., steel containments and liners of reinforced concrete containments) are identified and their manifestations described. Background information and data for improving and developing methods to assess the effects of age-related degradation on structural performance are provided. Techniques for detection of degradation are reviewed and research related to development of methods for inspection of inaccessible regions of the containment pressure boundary presented. Application of structural reliability analysis methods to develop condition assessment tools and guidelines is described.

Application of a Large-Scale Database System for the Analysis and Design of the US EPR™ Standard Nuclear Power Plant

2012 ◽  
Author(s):  
Se-Kwon Jung ◽  
Joseph Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Database System ◽  
Database Management System ◽  
Plane Shear ◽  
Analysis And Design ◽  
Shear Design ◽  
The U.S

Due to the increased size and complexity of large-scale commercial and industrial structures, it is increasingly challenging to manage key engineering data including analysis and design results of these structures. This paper presents a novel approach of using large-scale database systems as a means to gather, organize and manage key analysis and design results of a large-scale structure. Specifically, this paper describes in detail the development process of the backend database management system (DBMS) for the U.S. EPR™ Standard Nuclear Power Plant (NPP) Nuclear Island (NI) structures. The database system consists of three parent database tables to represent three representative groups of load combinations applicable to the U.S. EPR™ Standard NPP. Inheriting all characteristics of an applicable parent table, a primary child table in the database system represents a particular U.S. EPR™ NI Safety-Related structure in its entirety while a secondary child table a group of slabs or walls of the structure. Each secondary table is comprised of database fields that are representative of various structural demands, section capacities, reinforcing ratios, and demand-to-capacity ratios for three reinforced concrete design conditions (i.e., combined axial force and bending design, in-plane shear design, and out-of-plane shear design). The complete database system with fully populated tables is a central repository where all analysis and design results for the U.S. EPR™ NI Common Basemat structures are stored and sorted. To facilitate data queries from the developed backend database system, this paper introduces a user-friendly frontend interface program developed using Visual Basic Application (VBA) with Excel. Potential benefits of the developed database system are demonstrated with simple application examples involving simple data queries only followed by complex engineering tasks that require a more advanced form of data queries.

scholarly journals Method of Seismic Capacity Analysis of Steel Structure in Nuclear Power Plant

2021 ◽  
Vol 21 (2) ◽  
pp. 111-117
Author(s):  
Dongwon Lee ◽  
Namheoyng Lim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Steel Structure ◽  
Related Structure ◽  
Capacity Analysis ◽  
Seismic Capacity ◽  
Design Earthquake ◽  
External Forces

A steel structure in a nuclear power plant is typically constructed next to major safety-related structures. Accordingly, the structural integrity of the steel structure must be achieved until the safety-related building is damaged by external forces. Consequently, the steel structure should have seismic capacity while maintaining the structural integrity of the surrounding safety-related structure. An optimized method for the seismic capacity against the beyond design earthquake was developed to reflect this capacity concept.

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