Vibration Testing of Compression Joints

2017 ◽  
Author(s):  
Paul Hirschberg ◽  
Mark Sindelar ◽  
Maher Kassar ◽  
Ron Haupt
Keyword(s):  
Small Diameter ◽  
Operating Experience ◽  
Hydraulic Control ◽  
Damage Resistance ◽  
Premature Failure ◽  
Vibration Load ◽  
Piping Systems ◽  
Input Acceleration ◽  
Vibration Problems ◽  
Industry Experience

The use of compression joints in ASME Nuclear Class 2 and 3 small diameter piping systems has become increasingly popular because their installation does not require welding, and therefore saves time, money, and radiation dose. An important question is whether these types of joints can be a practical alternative to socket welds in piping systems subject to vibration. There have been numerous operating experience events where socket welds have developed cracks due to high cycle fatigue. On the other hand, parts of compression joints plastically deform the pipe to grip and create their sealing connection; the application of a high cycle vibration load to an already plastically strained pipe might lead to premature failure. It is desired to know whether at least one type of compression joint would perform better or worse than socket welds in such an environment. In this paper, a testing methodology is described, in which one supplier’s coupling joint design was tested for vibration loading in tubing assemblies of varying sizes. The intended application for these joints is in an Electro-Hydraulic Control system at a northeastern Boiling Water Reactor plant. Industry experience reports have identified past vibration problems in this system at other plants. A test setup was devised, in which multiple specimens could be tested simultaneously by adjusting specimen natural frequency, shake table speed, and input acceleration. Fatigue Strength Reduction Factors were derived, allowing the resistance to fatigue failure to be quantified. Both compression joints and socket welds were tested using the same procedures, in order that the fatigue damage resistance could be compared between the two types of joints.

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

2021 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Thermal Transients ◽  
Piping Systems ◽  
Class 1 ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Regulatory Experience of Leak-Before-Break (LBB) Technology to the High Energy Piping Systems in Korea

2010 ◽  
Author(s):  
Yeon-Ki Chung ◽  
Jin-Su Kim ◽  
Hae-Dong Chung ◽  
Young-Hwan Choi
Keyword(s):  
Material Properties ◽  
Operating Experience ◽  
Leak Detection ◽  
High Energy ◽  
Leak Rate ◽  
Regulatory Review ◽  
Screening Criteria ◽  
Piping Systems ◽  
Crack Stability ◽  
Leak Before Break

The application of the leak-before-break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved for the several high energy piping systems inside containment in Korea. The main purpose of the LBB application for these systems at the design stage is the removal of the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to the elimination of the pipe whip restraints and jet impingement barriers so as to increase the access the inspections. LBB technology is based on the low probability of pipe ruptures in the candidate piping systems using fracture mechanics and the insights from the state-of-the-art technology including operating experience. The procedures for LBB application is fundamentally based on the Unite States Nuclear Regulatory Commission (US NRC) requirements as detailed in the standard review plan (SRP) 3.6.3. However, a number of the additional requirements and issues are not specified in the review procedure during regulatory review were imposed and addressed during the review process. The regulatory review is focused on the confirmation on the methods for the elements in the screening criteria and several technical concerns on the determination of material properties, the validation of crack evaluation methods and leak rate estimation in the LBB evaluation considering the adequate margin. Although the application of the LBB has been approved by the safety authority for some high energy systems, the validation of LBB is continuously maintained in consideration of operating experience. In this paper, the regulatory positions for LBB application are described for the areas of screening criteria, leak rate estimation including the capability of leak detection system, material properties, load combination, crack stability methods, and margins in the crack stability evaluations. The issues encountered during the regulatory review such as the dynamic fracture test to consider the dynamic strain aging (DSA) of carbon and low alloy steel, thermal stratification and striping in the pressurizer surge line, water/steam hammer in main steam lines, and estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry are also introduced. In addition, several regulatory actions to improve the reliability in the capability of leak detection systems and to clarify the screening criteria such as the corrosion resistance is provided.

The Treatment of ISI Uncertainty in Extremely Low Probability of Rupture

2011 ◽  
Author(s):  
Patrick G. Heasler ◽  
Scott E. Sanborn ◽  
Steven R. Doctor ◽  
Michael T. Anderson
Keyword(s):  
Epistemic Uncertainty ◽  
Probability Distributions ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Piping Systems ◽  
Primary Water ◽  
Low Probability ◽  
Regulatory Commission ◽  
Leak Before Break

The U.S. Nuclear Regulatory Commission (NRC) in cooperation with the nuclear industry is constructing an improved probabilistic fracture model for piping systems that in the past have not been susceptible to known degradation processes that could lead to pipe rupture. Recent operating experience with primary water stress corrosion cracking (PWSCC) has challenged this prior position of leak-before-break and which has now become known as “extremely Low Probability of Rupture” (xLPR). This paper focuses on the xLPR model’s treatment of uncertainty for in-service inspection. In the xLPR model, uncertainty is classified as either aleatory or epistemic, and both types of uncertainty are described with probability distributions. Earlier PFM models included aleatory, but ignored epistemic, uncertainty, or attempted to deal with epistemic uncertainty by use of conservative bounds. Thus, inclusion of both types of uncertainty in xLPR should produce more realistic results than the earlier models. This work shows that by including epistemic uncertainty in the xLPR ISI module, there can be a significant effect on rupture probability; however, this depends upon the specific scenarios being studied. Some simple scenarios are presented to illustrate those where there is no effect and those having a significant effect on the probability of rupture.

Excitation Tests on Elbow Pipe Specimens to Investigate Failure Behavior Under Excessive Seismic Loads

2015 ◽  
Author(s):  
Izumi Nakamura ◽  
Naoto Kasahara
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Safety Management ◽  
Dominant Frequency ◽  
Failure Behavior ◽  
Seismic Loads ◽  
Energy Agency ◽  
Piping Systems ◽  
Input Acceleration

After the accident at Fukushima Dai-ichi Nuclear Power Plant in the 2011 Great East Japan Earthquake, the International Atomic Energy Agency (IAEA) requires to consider the design extension conditions (DEC) for the safety management of nuclear power plants (NPPs). In considering DEC, it is necessary to clarify the possible failure modes of the structures and their mechanism under the extreme loadings. Because piping systems are one of the representative components of NPP, and there is a possibility to failure at seismic events, the authors conducted an experimental investigation on failure modes and their mechanisms of piping systems under excessive seismic loads. The experiments are categorized into the fundamental plate tests and pipe component tests. In this paper, the results of the pipe component tests would be described. In the pipe component tests, the authors used piping specimens constituted with one steel elbow and a weight. Though the input acceleration level was much over the allowable level to prevent collapse failure by the seismic design, the failure mode obtained by the excitation tests were mainly the fatigue failure. The reduction of the dominant frequency and the increase of the hysteresis damping were clearly observed in the high-level input acceleration due to the plastic deformation, and they affected the specimens’ vibration response greatly.

Deficient Materials in Hot Reheat Seam Welded High-Energy Piping

2017 ◽  
Author(s):  
Kent Coleman ◽  
Stan Rosinski ◽  
Jude Foulds
Keyword(s):  
Filler Metal ◽  
Damage Mechanism ◽  
High Energy ◽  
Management Program ◽  
Life Assessment ◽  
Thick Wall ◽  
Premature Failure ◽  
The World ◽  
Piping Systems ◽  
American Society

Although failures of seam welded high-energy piping date to about 1970, the utility industry concern about premature failure of longitudinal seam welded piping has been of the utmost importance since the mid 1980’s driven primarily by well publicized piping failures at that time. The major concern is with hot reheat piping, some of which failed catastrophically resulting in substantial costs and personnel injury. Many utilities manage their high energy piping integrity through a combination of engineering analysis and periodic inspections. At many utilities around the world, however, high energy piping has not received the attention that normally occurs after a major failure because the perception is that there have been few failures. EPRI has compiled a database of over 50 failures and large areas of damage in utility piping systems around the world and it does not include the entire utility experience but still demonstrates the need for a diligent high energy piping integrity management program. The investigation of a recent failure in a 42 in. (1066 mm) diameter, 2 in. (50.8 mm) thick wall, ASTM A155 (American Society of Testing and Materials) seam welded hot reheat pipe demonstrated a first of its kind damage mechanism which determined that inappropriate welding filler metal was utilized for at least some of the weld passes resulting in a weldment that was weak in creep. Due to the placement of the incorrect welding filler metal, the failure occurred as a leak instead of a rupture however, the damage on the inside surface of the pipe extended for over 9ft. (2743 mm). This paper presents the results of the failure analysis and life assessment work performed and provides guidance for the rest of the utility fleet.

A Simplified Method to Obtain Subway Vibration Load Based on Analysis in Frequency Domain

2012 ◽  
Vol 256-259 ◽  
pp. 1195-1200
Author(s):  
Jing Bo Liu ◽  
Xiao Bo Zhang ◽  
Dong Dong Zhao ◽  
Wen Hui Wang
Keyword(s):  
Numerical Simulation ◽  
Frequency Domain ◽  
Ground Vibration ◽  
Domain Analysis ◽  
Frequency Domain Analysis ◽  
Vibration Test ◽  
Vibration Load ◽  
Simplified Method ◽  
Vibration Problems

To obtain reasonable subway vibration load is the key to many subway vibration problems. This paper uses a simplified method to determine subway vibration load in frequency domain analysis, based on ground vibration test induced by subway traveling and numerical simulation of soil-subway model. The operating steps are explained and one example for calculating subway vibration load is given. The result shows that this method is a certain degree of effective in analysis of ground vibration problems, since that it relies on ground vibration test. The operating steps of this method are simple, and also the method can reflect characteristic of randomness of subway vibration.

Excitation Tests on Elbow Pipe Specimens to Investigate Failure Behavior Under Excessive Seismic Loads

2017 ◽  
Vol 139 (6) ◽  
Author(s):  
Izumi Nakamura ◽  
Naoto Kasahara
Keyword(s):  
Nuclear Power ◽  
Failure Modes ◽  
Dominant Frequency ◽  
Failure Behavior ◽  
Seismic Loads ◽  
Experimental Stress ◽  
Safe Management ◽  
Piping Systems ◽  
Input Acceleration ◽  
High Level

The accident at the Fukushima Dai-ichi Nuclear Power Plant (NPP) resulting from the 2011 Great East Japan Earthquake raised awareness as to the importance of considering Beyond Design Basis Events (BDBE) when planning for safe management of NPPs. In considering BDBE, it is necessary to clarify the possible failure modes of structures under extreme loading. Because piping systems are one of the representative components of NPPs, an experimental investigation was conducted on the failure of a pipe assembly under simulated excessive seismic loads. The failure mode obtained by excitation tests was mainly fatigue failure. The reduction of the dominant frequency and the increase of hysteresis damping were clearly observed in high-level input acceleration due to plastic deformation, and they greatly affected the specimens’ vibration response. Based on the experimental results, a procedure is proposed for calculating experimental stress intensities based on excitation test so that they can be compared with design limitations.

scholarly journals High Vibration Problem in Compressor Piping Systems: A Case Study

1997 ◽  
Author(s):  
G. Vijaya Kumar ◽  
S. Raghava Chary ◽  
A. Rajamani
Keyword(s):  
Natural Frequencies ◽  
Excitation Frequency ◽  
Mode Shapes ◽  
Piping System ◽  
Mechanical Responses ◽  
Normal Limits ◽  
Piping Systems ◽  
Vibration Problems ◽  
Excitation Frequencies

High vibration problems resulting in damage to supports, instrument stubs etc. have been experienced in many compressor piping systems installed at different fertilizer plants. Investigations aimed at a solution to the problem included vibration measurements on the suction and discharge piping, and mathematical modeling of the piping. The measurements indicated presence of an excitation frequency in the range of 30–35% of the compressor running speed. Dynamic analysis of the piping system showed the presence of natural frequencies coinciding with or very near to the excitation frequencies. This has been further confirmed by impact tests. Analytical mode shapes clearly show that the antinodes match with high vibration zones observed at the site. The mathematical models were used to determine optimum configurations which would separate mechanical responses from excitation frequencies. These modifications have been implemented at site and the piping vibrations are within normal limits.

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