Effect of Bending Load on the Failure Pressure of Wall-Thinned Pipe Bends

2010 ◽  
Author(s):  
Jin Weon Kim ◽  
Oon Young Jung
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Operating Conditions ◽  
Accelerated Corrosion ◽  
Failure Pressure ◽  
Bending Load ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Under normal operating conditions, piping systems in nuclear power plants (NPPs) are subject not only to internal pressure but also to bending loads induced by deadweight and thermal expansion [1]. Bending is thus considered to be an important factor in evaluating the integrity of defective piping components. Local wall-thinning due to flow-accelerated corrosion is a main degradation mechanism of carbon steel piping systems in NPPs [2], and the integrity evaluation of wall-thinned piping components has become an important issue [3]. This study investigated the effects of bending load on the failure pressure of wall-thinned pipe bends under internal pressure. Our previous study experimentally evaluated the bending load effects on the failure pressure of wall-thinned elbows under displacement controlled in-plane bending load [4], but the numbers of experimental data were insufficient to determine the effects of bending load on the failure pressure of wall-thinned pipe bends. Therefore, the present study systematically evaluates the effects of bending load on the failure pressure of wall-thinned pipe bends using parametric finite element analyses.

Experimental Evaluation of the Bending Load Effect on the Failure Pressure of Wall-Thinned Elbows

2008 ◽  
Author(s):  
Jin-Weon Kim ◽  
Yeon-Soo Na ◽  
Sung-Ho Lee ◽  
Chi-Yong Park
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Load Effect ◽  
Failure Pressure ◽  
Wall Thinning ◽  
Bending Load ◽  
Piping Systems ◽  
Plane Bending

During normal operating conditions, piping systems in nuclear power plants are subject to internal pressure and to bending loads induced by deadweight, thermal expansion, and internal pressure, and understanding the effect of bending load on the failure of wall-thinned elbows is important to evaluate the failure pressure reliably. This study includes a series of burst tests using real-scale 4-inch schedule 80 elbow specimens with local wall-thinning under combined internal pressure and in-plane bending load applied by displacement control. The results are compared with those tested under simple internal pressure only. In the tests, various circumferential thinning angles (θ/π = 0.125, 0.25, 0.5, 1.0) and thinning locations (intrados, extrados, and full circumference) were considered. Each specimen was initially subjected to an in-plane bending load, closing mode for extrados wall-thinned elbows and opening mode for intrados wall-thinned elbows, and then internal pressure was applied up to point of final failure. The results showed that the effect of in-plane bending on the failure pressure and failure mode was minor under all wall-thinning conditions. In addition, the dependence of failure pressure on the circumferential thinning angle and thinning locations was identical to that observed under simple internal pressure.

Experimental Evaluation of the Bending Load Effect on the Failure Pressure of Wall-Thinned Elbows

2009 ◽  
Vol 131 (3) ◽  
Author(s):  
Jin-Weon Kim ◽  
Yeon-Soo Na ◽  
Sung-Ho Lee
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Failure Behavior ◽  
Load Effect ◽  
Failure Pressure ◽  
Wall Thinning ◽  
Bending Load ◽  
Plane Bending

During normal operating conditions, piping systems in nuclear power plants are subject to internal pressure and bending loads induced by deadweight, thermal expansion, and internal pressure. Thus, understanding the effect of bending load on the failure of wall-thinned elbows is important to understand failure behavior and to evaluate the failure pressure reliably. This study includes a series of burst tests using full-scale 4-in. schedule 80 elbow specimens with local wall-thinning under combined internal pressure and in-plane bending load. The results are compared with those tested under simple internal pressure only. In the tests, various circumferential thinning angles (θ/π=0.125, 0.25, 0.5, 1.0) and thinning locations (intrados, extrados, and full-circumference) were considered. Each specimen was initially subjected to a displacement controlled in-plane bending load, closing mode for extrados wall-thinned elbows, and opening mode for intrados wall-thinned elbows, and then internal pressure was applied up to the point of final failure. The results showed that the effect of in-plane bending on the failure pressure and failure mode was minor under all wall-thinning conditions. In addition, the dependence of failure pressure on the circumferential thinning angle and thinning locations was identical to that observed under simple internal pressure.

Effects of Flow Accelerated Corrosion on Piping Steady-State Vibration Evaluations

2003 ◽  
Author(s):  
Brian J. Voll
Keyword(s):  
Steady State ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Vibration Monitoring ◽  
Accelerated Corrosion ◽  
Monitoring Programs ◽  
Wall Thinning ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Piping steady-state vibration monitoring programs were implemented during preoperational testing and initial plant startup at most nuclear power plants. Evaluations of piping steady-state vibrations are also performed as piping and component failures attributable to excessive vibration are detected or other potential vibration problems are detected during plant operation. Additionally, as a result of increased flow rates in some piping systems due to extended power uprate (EPU) programs at several plants, new piping steady-state vibration monitoring programs are in various stages of implementation. As plants have aged, pipe wall thinning resulting from flow accelerated corrosion (FAC) has become a recognized industry problem and programs have been established to detect, evaluate and monitor pipe wall thinning. Typically, the piping vibration monitoring and FAC programs have existed separately without interaction. Thus, the potential impact of wall thinning due to FAC on piping vibration evaluations may not be recognized. The potential effects of wall thinning due to FAC on piping vibration evaluations are reviewed. Piping susceptible to FAC and piping susceptible to significant steady-state vibrations, based on industry experience, are identified and compared. Possible methods for establishing links between the FAC and vibration monitoring programs and for accounting for the effects of FAC on both historical and future piping vibration evaluations are discussed.

Wall-Thinning Analysis of the Pipelines in the Secondary Section of NPP

2013 ◽  
Author(s):  
Na Ma ◽  
Li Wang ◽  
Jinguang Qin
Keyword(s):  
Carbon Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Xrd Analysis ◽  
Operating Conditions ◽  
Chemical Compositions ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Erosion Corrosion ◽  
Flow Accelerated Corrosion

Wall-thinning investigation of three carbon steel pipe samples from secondary section of nuclear power plants has been carried out in this paper. The operating conditions of the three pipe samples are quite different, which leads to the different wall-thinning reasons and characteristics of the pipes. The chemical compositions of the steel materials, the stereomicroscope examinations, SEM examinations, as well as the XRD analysis are performed. The results show that: The wall-thinning of No.1 elbow was caused by erosion corrosion; the wall-thinning of No.2 elbow was caused by flow accelerated corrosion; the wall-thinning and crevasse of No.3 orifice plate was caused by cavitations. Measures to solve the wall-thinning problems of different pipes are also given in this paper.

Tackling Erosion in Nuclear Piping Systems

2007 ◽  
Author(s):  
Harold M. Crockett ◽  
Jeffrey S. Horowitz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Training Module ◽  
Droplet Impingement ◽  
Damage Mechanisms ◽  
Accelerated Corrosion ◽  
Computerized Training ◽  
Piping Systems ◽  
Research Plan ◽  
Flow Accelerated Corrosion

Various mechanisms degrade power piping in nuclear power plants. The most important mechanism has been flow-accelerated corrosion (FAC). FAC has caused ruptures and leaks and has led to numerous piping replacements. U.S. utilities are using a combination of EPRI software and aggressive inspection programs to deal with FAC. However, current technology does not deal with erosive forms of attack including, cavitation erosion, flashing erosion, droplet impingement, and solid particle erosion. These forms of degradation have caused shutdowns and leaks have become a maintenance issue. To deal with these problems EPRI has begun a series of projects in this area. The first of these was a comprehensive report on erosion in piping systems. This work was followed with a computerized training module designed to educate utility engineers about erosive attack. Further steps are planned to deal with these forms of degradation. The first will be a meeting with knowledgeable EPRI and utility engineers to prioritize the damage mechanisms. From this meeting a research plan will be developed. This paper will present a description of erosive damage mechanisms and describe the planned R&D to deal with these mechanisms.

Erosion in Nuclear Piping Systems

2010 ◽  
Vol 132 (2) ◽  
Author(s):  
Harold M. Crockett ◽  
Jeffrey S. Horowitz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cavitation Erosion ◽  
Solid Particle Erosion ◽  
Particle Erosion ◽  
Droplet Impingement ◽  
Accelerated Corrosion ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Various mechanisms degrade components and power piping in nuclear power plants. The mechanism with the greatest consequence has been flow-accelerated corrosion (FAC). FAC has caused ruptures and leaks and has led to numerous piping replacements. United States utilities use a combination of EPRI guidance, software, and aggressive inspection programs to deal with FAC. However, current technology does not detail guidance for erosive forms of attack including, cavitation erosion, flashing erosion, droplet impingement, and solid particle erosion. These forms of degradation have caused shutdowns, and leaks have become a maintenance issue. This brief will present a description of erosive damage mechanisms found in nuclear power plants.

The Role of CHECWORKS™ in an Effective FAC Program

2014 ◽  
Author(s):  
David Smith ◽  
Jeffrey Horowitz
Keyword(s):  
Water Chemistry ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Power Level ◽  
Carbon Steels ◽  
Operating Conditions ◽  
Inspection Planning ◽  
Accelerated Corrosion ◽  
The Impact

Flow-accelerated corrosion (FAC) is a degradation mechanism that attacks carbon steels under conditions often found in nuclear and fossil power plants. FAC has been responsible for a number of significant accidents in nuclear power plants. The most recent noteworthy accident was at the Mihama Unit 3 (Japan) where the catastrophic failure of a pipe in the condensate system resulted in five fatalities. FAC can affect virtually all of the carbon steel piping and components in the power cycle of nuclear reactors. The presence of wall thinning caused by FAC is determined through the use of non destructive examination (NDE) techniques. For large-bore piping components the most commonly used approach is ultrasonic technique (UT) using an inspection grid applied to the piping components. As FAC is a reasonably well-understood degradation mechanism, a number of computer programs have been developed to help utility engineers determine the inspection locations and managing the data. CHECWORKS™ — the most commonly used computer program for this purpose — is the program discussed in this paper. The use of CHECWORKS™ in utilities’ FAC programs will be described. Particular emphasis will be placed on: inspection planning, handling power uprates and other changes to operating conditions. Inspection planning is one of the most common uses of the CHECWORKS™ software. As there are typically around 5,000 FAC-susceptible components in a reactor system, utility engineers must select the components with the highest FAC-rates for inspection. CHECWORKS™ uses its predictions combined with plant inspection data to provide a best estimate of FAC rates. From these FAC rate predictions and knowledge of the piping schedule and allowable wall thicknesses, inspection locations are determined. Both the water chemistry used and the local operating conditions strongly influence the rate of FAC. CHECWORKS™ is used to study potential changes to water chemistry, system operation or power level to determine the impact of such changes on FAC rates and hence inspection locations. Of particular interest is the use of CHECWORKS™ to determine the impacts of power uprates. Because of the complicated parametric behavior of FAC rates, changing the power level will likely increase the FAC rates in some areas of the plant while other areas will likely see a decrease in FAC rates. This fact requires a pre-uprate analysis to determine how an inspection program will need to be modified. This paper provides a description of how CHECWORKS™ is used in the above applications as well as showing typical examples of its usefulness in these analyses.

A Study on the Cause Investigation of the Local Pipe Wall Thinning in the Downstream of an Orifice Through the UT Data and Numerical Analysis

2011 ◽  
Author(s):  
Kyeong Mo Hwang ◽  
Hun Yun ◽  
Chan Kyoo Lee
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Liquid Droplet ◽  
Degradation Mechanism ◽  
Shock Focusing ◽  
Pump Impeller ◽  
Accelerated Corrosion ◽  
Focusing Effect ◽  
Piping Systems

Piping systems in nuclear power plants are affected by degradation mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, LDI (Liquid Droplet Impingement) and the like. FAC is well-known degradation mechanism. Most nuclear power plants, therefore, implement inspection programs to detect and control damages by FAC. On the other hand, various erosion mechanisms occurring on piping lines and local components lead to wear, leaks, and failures on piping systems. This paper describes intensive wear phenomenon on pipes, elbows, and tees located in upstream of heat exchangers. In the beginning of the study, it appeared to be flow cavitation damages. This is because the orifice is installed in the upstream section of a damaged elbow. After reviewing the damaged surface, it turned out that flow cavitation was not the key issue in this case. It is supposed that shock focusing effect, related to the bubbles that are created by vibration of a pump impeller, is the main reason of this intensive wear. The phenomena of shock focusing effect are known to have resulted from the pressure concentration due to collapsing bubbles in liquid.

On-Site Evaluation of Bending Stresses at the Root of Small Bore Pipes by Accelerometric Measurements

2003 ◽  
Author(s):  
J. Guillou ◽  
L. Paulhiac
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Numerical Models ◽  
Operating Conditions ◽  
Strain Gauges ◽  
Bending Stress ◽  
Piping Systems ◽  
Bending Stresses ◽  
Site Evaluation

Several vibration-induced failures at the root of small bore piping systems occurred in French nuclear power plants in past years. The evaluation of the failure risk of the small bore pipes requires a fair estimation of the bending stress under operating conditions. As the use of strain gauges is too time-consuming in the environmental conditions of nuclear power plants, on-site acceleration measurements combined with numerical models are easier to handle. It still requires yet a large amount of updating work to estimate the stress in multi-span pipes with elbows and supports. The aim of the present study is to propose an alternate approach using two accelerometers to measure the local nozzle deflection, and an analytical expression of the bending stiffness of the nozzle on the main pipe. A first formulation is based on a static deformation assumption, thus allowing the use of a simple analog converter to get an estimation of the RMS value of the bending stress. To get more accurate results, a second method is based on an Euler Bernoulli deformation assumption: a spectral analyzer is then required to get an estimation of the spectrum of the bending stress. A better estimation of its RMS value is then obtained. An experimental validation of the methods based on strain gauges has been successfully performed.

A Procedure for Determining the Safe Operation Time of Equipment and Pipelines Based on Nondestructive Testing Results

Vestnik MEI ◽  
2020 ◽  
Vol 6 (6) ◽  
pp. 11-17
Author(s):  
Dmitriy A. Kuz'min ◽  
◽  
Aleksandr Yu. Kuz'michevskiy ◽  
Artem E. Gusarov ◽  
◽  
...  
Keyword(s):  
High Pressure ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operation Time ◽  
Safe Operation ◽  
Accelerated Corrosion ◽  
Equipment And Pipelines ◽  
Flow Accelerated Corrosion

The reliability of nuclear power plants (NPPs) has an influence on power generation safety and stability. The reliability of NPP equipment and pipelines (E&P), and the frequency of in-service inspections are directly linked with damage mechanisms and their development rates. Flow accelerated corrosion (FAC) is one of significant factors causing damages to E&P because these components experience the influence of high pressure, temperature, and high flow velocity of the inner medium. The majority of feed and steam path components made of pearlitic steels are prone to this kind of wear. The tube elements used in the coils of high pressure heaters (HPH) operating in the secondary coolant circuit of nuclear power plants equipped with a VVER-1000 reactor plant were taken as the subject of the study. The time dependences of changes in the wall thickness in HPH tube elements are studied proceeding from an analysis of statistical data of in-service nondestructive tests. A method for determining the initial state of the E&P metal wall thickness before the commencement of operation is proposed. The article presents a procedure for predicting the distribution of examined objects' wall thicknesses at different times of operation with determining the occurrence probability of damages caused by flow accelerated corrosion to calculate the time of safe operation until reaching a critical state. A function that determines the boundary of permissible values of the HPH wall thickness distributions is obtained, and it is shown that the intervals of in-service inspections can be increased from 6 years (the actual frequency of inspections) to 9 years, and the next in-service inspection is recommended to be carried out after 7.5 years of operation. A method for determining the existence of FAC-induced local thinning in the examined object has been developed. The developed approaches and obtained study results can be adapted for any pipelines prone to wall thinning to determine the frequency of in-service inspections (including an express analysis based on the results of a single nondestructive in-service test), the safe operation time, and quantitative assessment of the critical value reaching probability.

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