Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

Design Strength of Primary Structural Welds in Freestanding Structures

1991 ◽  
Vol 113 (3) ◽  
pp. 471-475
Author(s):  
K. P. Singh ◽  
A. I. Soler ◽  
S. Bhattacharya
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Structural Components ◽  
Rational Analysis ◽  
Free Standing ◽  
Design Strength ◽  
Analysis Technique ◽  
Asme Code

A rational analysis technique to evaluate structural integrity of primary welds in free-standing structures in accordance with the ASME Code is presented. This paper is intended to fill the void in the ASME Code rules for analyzing welds under “faulted” (level D) conditions in nonlinear free-standing structural components used in safety-related applications in nuclear power plants.

Dynamic Response of a Nuclear Power Plant Subjected to External Accident Event

2010 ◽  
Author(s):  
Rosa Lo Frano ◽  
Giuseppe Forasassi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Containment System ◽  
Dynamic Loadings ◽  
Water Reactors ◽  
Set Up

In recent times there is a renewed worldwide interest in the development and application of advanced nuclear power plants (NPPs). Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR in Finland and France, AP1000 in China, etc.) and more are under consideration for licensing in several countries. Innovative NPPs are designed to be built with very broad siting conditions; therefore the safety aspects related to the external events might follow new scenarios and failure modes, different from those well known for the currently operated reactors. In this paper, the intent is evaluating the structural integrity of a nuclear containment system subjected to dynamic loadings due to a Design Base Earthquake and an aircraft impact (large size civilian jets or military aircrafts impact), which represent the two most relevant external accidents that should be considered and investigated as part of the basic design of a NPP in particular a III+ and IV Gens. In fact a suitable safety design of the NPP containment system (according to the international safety and design code guidelines, as NRC or IAEA ones), even if designed to meet other design goal, may represent a “built-in protection” to avoid or mitigate the effects of mentioned dynamic loadings. To the purpose a rather sophisticated numerical methodology, adopting finite element (FEM) approach, is employed for studying the overall dynamic behaviour of nuclear reactor and to determine the structural effects of the propagation of dynamic seismic as well as impulsive loads (containment structure response) up to the relevant nuclear components. Therefore representative three-dimensional FEM models of mentioned NPP containment and aircraft structures were set up, and used, in the performed analyses taking also into account the suitable materials behaviour and their related constitutive laws as well as the seismic excitation (determined according to the NRC rules). Moreover the performed analyses and the carried out response analyses of internal components, to both the ground motion and impact loads, were studied to check the considered NPP containment strength reserve in the case of the considered events. The obtained results seem to confirm the possibility to achieve an optimization of the NPP internal components.

Limit Analysis of a Pipe Section With Non-Uniform Wall Thinning

2002 ◽  
Author(s):  
Keshab K. Dwivedy
Keyword(s):  
Limit Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Pipe Wall ◽  
Limit Load ◽  
Pipe Section ◽  
Wall Thinning ◽  
Uniform Wall

Certain process piping in nuclear and non-nuclear power plants undergo pipe wall thinning due to flow assisted corrosion (FAC). This localized mechanism of corrosion combined with erosion is complex. The potential degradation of the pipe wall depends upon the water chemistry, operating temperature and pressure, flow velocity, piping material and piping configuration. The management of FAC in a power plant is performed in the following basic steps: Identification of potential locations, UT inspection of locations and characterization of pipe wall thinning, and evaluation of wall thinning to establish structural integrity and/or repair/replacement. The section of the pipe is repaired or replaced if the structural integrity cannot be established until next scheduled inspection. In the past 15 years, FAC programs have been established in nuclear power plants. Structural integrity evaluation is a part of the program. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. Pressure design methods are formalized for uniform and non-uniform wall thinning. However, the limit analysis methods for moment loading in the code rules are formulated for uniform thinning of the wall for simplicity. FAC related wall thinning is truly non-uniform, and treating it as non-uniform in the analysis can show additional structural margin compared to analysis conservatively assuming a uniformly thinned wall. This paper has developed simple analytical formulation of limit load carrying capability of a pipe section with non-uniform thinning. The method of analysis is illustrated with examples of actual plant situations. The formulation developed here can be used with the ASME code method to extend remaining life of FAC degraded components until the plant can plan for repair or replacement. Thus the analytical tool can help the plant owners to save resources by performing repair and replacement in a planned manner.

Nuclear Regulatory Commission Staff View of Progress in the Nondestructive Testing of Cast Austenitic Steel Components

2014 ◽  
Author(s):  
Stephen E. Cumblidge
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Degradation Mechanism ◽  
Austenitic Steels ◽  
Inspection Method ◽  
The Past ◽  
Asme Code

Welds in cast austenitic steels (CASS) are very challenging to inspect using the current American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements. Supplement 9 of ASME Boiler and Pressure Vessel Code Section XI, Appendix VIII is still in the course of preparation, requiring inspectors to use ASME Code Section XI, Appendix III, which provides prescriptive ultrasonic testing (UT) requirements that are significantly less rigorous than UT techniques that have been demonstrated under Appendix VIII. The inability of licensees to demonstrate that the welds in CASS components meet ASME Code requirements has been an ongoing area of concern for the NRC staff. The lack of a reliable inspection method for welds in CASS materials has led to hundreds of relief requests over the past four decades. While no degradation mechanism has been found in CASS components to date, there is no guarantee that a new degradation mechanism affecting CASS welds will not emerge as nuclear power plants go beyond forty years of operation. Licenses need qualified procedures and personnel for the inspection of welds in CASS materials in order to put licensees into compliance with ASME Code, meet federal regulations, reduce the number of needed relief requests, and ensure the structural integrity of their welds. Over the past decade there have been significant developments in nondestructive examination (NDE) technology. The use of encoded phased array techniques using low frequency ultrasound has been shown to be able to reliably find flaws greater than 30% through wall in CASS materials with a variety of microstructures. Additionally, an improved understanding of the fracture mechanics of CASS components is being developed that shows the flaw sizes that can be tolerated in CASS components. These advances in NDE techniques and fracture mechanics theory are converging on a path to allow for qualifications of procedures and personnel for the ultrasonic inspections of welds in CASS components. Recent developments in ASME Code includes Code Case N-824, which provides guidance on the examination of CASS materials based on the advances in NDE technology and an improved understanding of the NDE techniques capable of finding flaws in CASS components as well as Code Case N-838 for flaw tolerance evaluations of CASS piping components. Finally, work on ASME Code Section XI Supplement 9 is progressing, with several important issues still to be addressed. The NRC staff sees a clear path forward and is working to ensure that qualified inspections of welds in CASS materials will be possible in the future.

Alternative Characterization Rules for Quasi-Laminar Flaws Based on 3D X-FEM Calculations

2015 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu ◽  
Anne-Sophie Bogaert
Keyword(s):  
Finite Element Method ◽  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Division I ◽  
Extended Finite Element ◽  
Alternative Characterization ◽  
Asme Code ◽  
Alternative Methodology

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the RPVs. In the frame of the Structural Integrity demonstration of these RPVs according to ASME XI principles, ASME XI IWB-3300 article requires combining closely spaced flaws in order to account for their mechanical interactions. However, it appeared early that the characterization rules were adapted neither to quasi-laminar flaws nor to such densities of flaws. Therefore, an alternative methodology to derive characterization rules for quasi-laminar flaws has been developed, implemented and validated. This work, based on 2D eXtended Finite Element Method (X-FEM) calculations and presented during ASME PVP 2014, has led to the proposed ASME Code Case N-848 “Alternative characterization rules for quasi-laminar flaws – Section XI, Division I”. This 2D approach, even though better suited to quasi-laminar flaws, results however in very conservative proximity rules. Therefore, it appeared that more realistic — although still conservative — proximity rules based on 3D X-FEM calculations could be developed.

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.

scholarly journals Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

2020 ◽  
Author(s):  
Evrim Oyguc ◽  
Abdul Hayır ◽  
Resat Oyguc
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Energy Sources ◽  
Finite Element Program ◽  
Design Spectrum ◽  
Reactor Building ◽  
Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

2005 ◽  
Vol 19 (11) ◽  
pp. 1988-1997 ◽  
Author(s):  
June-soo Park ◽  
Ha-cheol Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
Jai-hak Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Interference Fit ◽  
Reactor Coolant

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

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