Structural Integrity for Nuclear Power Plant Against Excessive Load on Design Extension Condition

2013 ◽  
Author(s):  
Daigo Watanabe ◽  
Kiminobu Hojo
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Factors ◽  
Design Code ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Current Design ◽  
Factor Design ◽  
Excessive Load ◽  
Load And Resistance Factor

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

Overview of Recent NUGENIA Activities Particularly in Relation to Structural Integrity

2015 ◽  
Author(s):  
John Sharples ◽  
Elisabeth Keim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Research ◽  
Joint Research ◽  
Non Profit ◽  
Integrity Assessment ◽  
Industry Research ◽  
Implementation And Dissemination ◽  
Technical Area

NUGENIA, an international non-profit association founded under Belgian legislation and launched in March 2012, is dedicated to nuclear research and development (R&D) with a focus on Generation II and III power plants. NUGENIA is the integrated framework between industry, research and safety organisations for safe, reliable and competitive nuclear power production, and is aimed at running an open innovation marketplace, to promote the emergence of joint research and to facilitate the implementation and dissemination of R&D results. The technical scope of NUGENIA consists of eight technical areas. One of these areas, Technical Area 4, is associated with the structural integrity assessment of systems, structures and components. A brief overview of recent NUGENIA activities in general is provided in this paper and a specific focus is given on developments in relation to Technical Area 4.

Assessing the Target Reliability Index for Nuclear Piping

2016 ◽  
Author(s):  
Kleio Avrithi
Keyword(s):  
Reliability Index ◽  
Historical Data ◽  
Allowable Stress ◽  
Safety Factors ◽  
Code Calibration ◽  
Target Reliability Index ◽  
Consistent Manner ◽  
Factor Design ◽  
Load And Resistance Factor ◽  
Allowable Stress Design

Previous research developed Load and Resistance Factor Design (LRFD) equations for Class 2 and 3 nuclear piping for different reliability levels and load combinations. The LRFD equations consider separate safety factors for each load and for the strength of steel in opposition to the Allowable Stress Design (ASD) equations used in the ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Div. 1, where only one safety factor is considered. In order to use the developed LRFD equations for the design of nuclear piping, specific reliability levels or else acceptable probabilities of failure need to be assigned to each Code equation. The paper discusses the available methods for evaluating the target reliability index, such as historical data of piping failures, expert-opinion elicitation, and Code calibration. Code calibration is the method of determining the existing level of reliability in the Code equations and assigning the same reliability to the developed LRFD equations in a consistent manner. Code Calibration is explained to be the more appropriate method of assigning reliability levels to the LRFD equations. The other methods can supplement the analysis results.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

Load and Resistance Factor Design (LRFD) of Nuclear Straight Pipes for Loads That Cause Primary Stress

2010 ◽  
Vol 132 (2) ◽  
Author(s):  
Kleio Avrithi ◽  
Bilal M. Ayyub
Keyword(s):  
Resistance Factor ◽  
Safety Factors ◽  
Primary Stress ◽  
Safety Margins ◽  
Division 1 ◽  
Partial Safety Factors ◽  
Factor Design ◽  
Load And Resistance Factor ◽  
Different Levels ◽  
Allowable Stress Design

Class 2 and 3 nuclear piping is designed according to the allowable stress design (ASD) method used in the ASME Boiler and Pressure Vessel (B&PV) code, Sec. III, Division 1, NC and ND-3600 according to which safety factors applied to the strength of steel (resistance) provide acceptable safety margins for the piping design. This paper describes the development of design equations according to the load and resistance factor design (LRFD) method for loads that cause primary stress such as sustained weight, internal pressure, and earthquake for different levels of piping operation. The LRFD method differs from the ASD since multiple factors, applied separately to each load and the strength of steel, provide safety margins that correspond to a known and acceptable probability of failure for the piping. Load combinations are provided, statistical properties for the variables under consideration are presented and the partial safety factors are moreover illustrated for different values of the target reliability index.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

Inspection and Evaluation Guidelines for Light Water Reactor Core Internals in Japan

2002 ◽  
Author(s):  
Kimihito Takeuchi ◽  
Naoto Iizuka ◽  
Masashi Kameyama ◽  
Haruo Fujimori ◽  
Yuichi Motora ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Integrity Assessment ◽  
Engineering Society ◽  
Safety Function ◽  
Internal Component

There have been many cracking experiences of light water reactor (LWR) core internals worldwide in the past. Thermal and Nuclear Power Engineering Society in Japan (TENPES) has organized a committee to prepare technically reasonable and appropriate inspection and evaluation guidelines (I&E guidelines) for core internals. This committee consists of scholars and representatives from electric utilities and nuclear plant vendors in Japan. I&E guidelines, which cover a rational inspection plan, structural integrity assessment and repair methods, have been developed considering nuclear safety function and structural strength of each core internal component. For BWR reactors, the development of I&E guidelines cover major core internal components like shroud support, core shroud, top guide, core plate, ICM and CRD housing, core spray piping and sparger, jet pump etc. For PWR reactors, the development of I&E guidelines cover baffle former bolts, barrel former bolts, core barrel weld, bottom mounted instrumentation, etc. The I&E guidelines will be completed by the end of March 2002. The basic concept of the guidelines, and a guideline for shroud support of a BWR as an example, are shown in this paper.

Reliability-Based Derivation of Partial Safety Factor for Structural Integrity Assessment

2008 ◽  
Author(s):  
Shuo Pan ◽  
Jianping Zhao
Keyword(s):  
Safety Factor ◽  
Fracture Mode ◽  
Structural Integrity ◽  
Research Work ◽  
Random Variables ◽  
Assessment Procedure ◽  
Safety Factors ◽  
Integrity Assessment ◽  
Partial Safety Factor ◽  
Partial Safety Factors

When there are uncertainties in the input random variables, or scatter in the material properties, probabilistic assessment is a useful tool for decision making in the field of safety analysis. The partial safety factor (PSF) method was aimed on ensuring that the failure probability did not exceed a target value. In order to be conservative the input value for each random variable during the assessment procedure should be multiplied by the partial safety factors. So it is essentially a deterministic assessment using conservative values of the input random variables and a relatively simple and independent method of assessing failure probabilities using R6 failure assessment diagram. The application of partial safety factors is an important breakthrough of assessment in structures containing defects. In recent years, sets of PSFs for load, defect size, fracture toughness and yield stress had been given in two standards, BS7910 and API579. However, the recommended PSFs in both standards were larger than the original PSFs in PD6493 which was replaced by BS7910. It is therefore a new method of calculating PSFs should be found to prove which is more appropriate and convenient for engineering application. In the case of the partial safety factor method target reliabilities in the range from 0.001 to 0.00001 were considered and new series of PSFs were derived from the results of reliability analysis for the linear elastic fracture mode and elastic-plastic fracture mode. After comparing with the PSFs in BS7910 and API 579, it is concluded that the partial safety factors were generally conservative compared to our research work.

Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Components in ASME Code Section XI

2003 ◽  
Author(s):  
Russell C. Cipolla ◽  
Guy H. DeBoo ◽  
Warren H. Bamford ◽  
Kenneth K. Yoon ◽  
Kunio K. Hasegawa
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Primary Objective ◽  
Surveillance Program ◽  
Acceptance Criteria ◽  
Evaluation Procedures ◽  
Asme Code ◽  
Nuclear Component ◽  
American Society ◽  
Plant Components

The primary objective of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI is to provide the rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Pressure boundary integrity in terms of assuring resistance to sudden and catastrophic failure has been an essential objective of the ASME Code since its inception in 1914. These objectives are especially important in ASME Section XI since maintaining pressure boundary integrity of components has a crucial role in ensuring safe and reliable operation in nuclear operating plants. The purpose of this paper is to describe the evaluation procedures, methods, and acceptance criteria for flaws detected in plant components during implementation of in-service inspection surveillance program. For nuclear plant components, pressure boundary integrity includes both leak integrity (no leakage from the reactor coolant system) and structural integrity (no rupture or burst of the pressure boundary). The evaluation requirements in ASME Section XI provide specific rules for assessing the acceptance limits for flaw indications that may be detected during the service life of a nuclear component. In addition to describing current flaw evaluation procedures, details of recent Code developments and improvements are discussed.

Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Zhou Gengyu ◽  
Liang Shuhua ◽  
Sun Lin ◽  
Lv Feng
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Load Capacity ◽  
Operating Conditions ◽  
Integrity Assessment ◽  
Inner Radius ◽  
Main Steam

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users. This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.

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