scholarly journals A Method Using Optical Contactless Displacement Sensors to Measure Vibration Stress of Small-Bore Piping

2013 ◽  
Vol 136 (1) ◽  
Author(s):  
Akira Maekawa ◽  
Takashi Tsuji ◽  
Tsuneo Takahashi ◽  
Michiyasu Noda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Strain Gauges ◽  
Curvature Radius ◽  
Light Emitting ◽  
Displacement Sensors ◽  
Piping Systems ◽  
Vibration Stress ◽  
Vibration Tests ◽  
Prototype Instrument

In nuclear power plants, vibration stress of piping is frequently evaluated to prevent fatigue failure. A simple and fast measurement method is attractive to evaluate many piping systems efficiently. In this study, a method to measure the vibration stress using optical contactless displacement sensors was proposed, the prototype instrument was developed, and the instrument practicality for the method was verified. In the proposed method, light emitting diodes (LEDs) were used as measurement sensors and the vibration stress was estimated by measuring the deformation geometry of the piping caused by oscillation, which was measured as the piping curvature radius. The method provided fast and simple vibration estimates for small-bore piping. Its verification and practicality were confirmed by vibration tests using a test pipe and mock-up piping. The stress measured by both the proposed method and an accurate conventional method using strain gauges were in agreement, and it was concluded that the proposed method could be used for actual plant piping systems.

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.

Noncontact Measurement Method of Vibration Stress Using Optical Displacement Sensors for Piping Systems in Nuclear Power Plants

2015 ◽  
Vol 1 (3) ◽  
Author(s):  
Akira Maekawa ◽  
Tsuneo Takahashi ◽  
Takashi Tsuji ◽  
Michiyasu Noda
Keyword(s):  
Nuclear Power ◽  
Measurement Method ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Vibration Measurement ◽  
Piping System ◽  
Noncontact Measurement ◽  
Displacement Sensors ◽  
Optical Displacement ◽  
Vibration Stress

In nuclear power plants, vibration stress of piping is frequently measured to prevent the occurrence of fatigue failure. A simpler and more efficient measurement method is desired for rapid integrity evaluation of piping. In this study, a method to measure vibration stress in a noncontact manner using optical displacement sensors is presented and validated. The proposed method estimates vibration-induced stress of small-bore piping directly using noncontact sensors based on a light-emission diode. First, the noncontact measurement method was proposed, and the measurement instrument based on the proposed method was developed for the validation. Next, vibration measurement experiments using the instrument were conducted for a mock-up piping system and an actual piping system. The measurement results were compared with the values measured by the conventional method of known accuracy using strain gauges. From this comparison, the proposed noncontact measurement method was demonstrated to be able to provide sufficient accuracy for practical use.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

Effects on Pipe Vibrations of Cavitation in an Orifice and in Globe-Style Valves

2006 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Frequency ◽  
Flow Regimes ◽  
Piping System ◽  
Low Frequencies ◽  
Choked Flow ◽  
Single Hole ◽  
Piping Systems ◽  
Large Amplitude Vibrations

A piping system of French nuclear power plants displays large amplitude vibrations in particular flow regimes. These troubles are attributed to cavitation generated by single-hole orifices in depressurized flow regimes. Real scale experiments on high pressure test rigs and on-site tests are then conducted to explain the observed phenomenon and to find a solution to reduce pipe vibrations. The first objective of the present paper is to analyze cavitation-induced vibrations in the single-hole orifice. It is then shown that the orifice operates in choked flow with supercavitation, which is characterized by a large unstable vapor pocket. One way to reduce pipe vibrations consists in suppressing the orifices and in modifying the control valves. Three technologies involving a standard trim and anti-cavitation trims are tested. The second objective of the paper is to analyze cavitation-induced vibrations in globe-style valves. Cavitating valves operate in choked flow as the orifice. Nevertheless, no vapor pocket appears inside the pipe and no unstable phenomenon is observed. The comparison with an anti-cavitation solution shows that cavitation reduction has no impact on low frequency excitation. The effect of cavitation reduction on pipe vibrations, which involve essentially low frequencies, is then limited and the first solution, which is the standard globe-style valve installed on-site, leads to acceptable pipe vibrations. Finally, this case study may have consequences on the design of piping systems. First, cavitation in orifices must be limited. Choked flow in orifices may lead to supercavitation, which is here a damaging and unstable phenomenon. The second conclusion is that the reduction of cavitation in globe-style valve in choked flow does not reduce pipe vibrations. The issue is then to limit cavitation erosion of valve trims.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

Computational Study of Conjugate Heat Transfer in T-Junctions

2009 ◽  
Author(s):  
Simon Kuhn ◽  
Bojan Nicˇeno ◽  
Horst-Michael Prasser
Keyword(s):  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Water ◽  
Computational Study ◽  
Solid Wall ◽  
Turbulent Viscosity ◽  
Life Time ◽  
Scale Model ◽  
Piping Systems

Thermal fatigue is a relevant problem in the context of life-time extension of nuclear power plants (NPP). In many piping systems in NPPs hot and cold water is mixed, which leads to high temperature fluctuations in the region close to the solid wall and resulting thermal loads on the pipe walls that can cause fatigue. One of the relevant geometric test cases for thermal fatigue is the mixing in T-junctions. In this study we apply large–eddy simulations (LES) to the mixing of hot and cold water in a T-junction. We perform a set of simulations by using different formulations of the LES subgrid scale model, i.e. standard Smagorinsky and dynamic procedure, to identify the influence of the modelled subgrid scales on the simulation results. The results exhibit a large difference between the models, which is caused by the use of turbulent viscosity wall–damping functions when applying the standard model.

Wave Propagation Analysis of Piping Structures

2009 ◽  
Author(s):  
Akemi Nishida
Keyword(s):  
Wave Propagation ◽  
Dynamic Behavior ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Beam Theory ◽  
Spectral Element ◽  
Analysis Tool ◽  
Propagation Analysis ◽  
Piping Systems

It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.

Failure Probability Assessment of Piping Systems Affected by SCC and Pipe Wall Thinning for LBB Analysis

2011 ◽  
Author(s):  
Hiromasa Chitose ◽  
Hideo Machida ◽  
Itaru Saito
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Probability Assessment ◽  
Wall Thinning ◽  
Piping Systems ◽  
Leak Before Break ◽  
Thinning Rate ◽  
Service Inspection

This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.

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