scholarly journals THE MEASURING CHANNELS CALIBRATION IN THE INDUSTRY

2021 ◽  
Vol 82 (4) ◽  
pp. 32-35
Author(s):  
Oleg Sulyma ◽  
◽  
Mykola Mykyichuk ◽  
Iryna Zinchuk ◽  
◽  
...  
Keyword(s):  
Nuclear Power ◽  
Measuring Equipment ◽  
Power Plants ◽  
High Accuracy ◽  
Measured Temperature ◽  
Measuring Systems ◽  
Wide Range ◽  
Means Of Measurement ◽  
Accuracy Of Measurement ◽  
Unity Of Measurements

The trend of development of measuring equipment in particular measuring systems and technological processes leads to the need of measuring the temperature without dismantling the primary transducers and stopping the technological process in a very wide range and with high accuracy. The choice of methods and means of measurement depends on such factors as the value of measured temperature, necessary accuracy of measurement, conditions of measurements, and operation of the investigated object. The topicality of the problem is that industrial metrology operates the temperature measuring systems, transducers of which are deployed in such a way that their dismantling is impossible due to the specifics of the process or dangerous for operators’ life (nuclear power plants). Such measuring systems, due to the specifics of their application, are not metrologically provided in the periods between major repairs of controlled objects, thus violating the unity of measurements.

scholarly journals High-Precision Noncontact Guided Wave Tomographic Imaging of Plate Structures Using a DHB Algorithm

2020 ◽  
Vol 10 (12) ◽  
pp. 4360
Author(s):  
Junpil Park ◽  
Jaesun Lee ◽  
Zong Le ◽  
Younho Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Detection Efficiency ◽  
Detection System ◽  
Automatic Detection ◽  
Guided Wave ◽  
Direct Access ◽  
Experimental Conditions ◽  
Plate Structures ◽  
Wide Range

The safety diagnostic inspection of large plate structures, such as nuclear power plant containment liner plates and aircraft wings, is an important issue directly related to the safety of life. This research intends to present a more quantitative defect imaging in the structural health monitoring (SHM) technique by using a wide range of diagnostic techniques using guided ultrasound. A noncontact detection system was applied to compensate for such difficulties because direct access inspection is not possible for high-temperature and massive areas such as nuclear power plants and aircraft. Noncontact systems use unstable pulse laser and air-coupled transducers. Automatic detection systems were built to increase inspection speed and precision and the signal was measured. In addition, a new Difference Hilbert Back Projection (DHB) algorithm that can replace the reconstruction algorithm for the probabilistic inspection of damage (RAPID) algorithm used for imaging defects has been successfully applied to quantitative imaging of plate structure defects. Using an automatic detection system, the precision and detection efficiency of data collection has been greatly improved, and the same results can be obtained by reducing errors in experimental conditions that can occur in repeated experiments. Defects were made in two specimens, and comparative analysis was performed to see if each algorithm can quantitatively represent defects in multiple defects. The new DHB algorithm presented the possibility of observing and predicting the growth direction of defects through the continuous monitoring system.

Direct Numerical Simulation of a Turbulent Flow on an Oscillating Wall Using Cut-Cell Based Immersed Boundary Method

2016 ◽  
Author(s):  
Chiaki Kino
Keyword(s):  
Turbulent Flow ◽  
Nuclear Power ◽  
Power Plants ◽  
Separated Flow ◽  
Immersed Boundary ◽  
Turbulent Structures ◽  
Flow Induced Vibration ◽  
Production Term ◽  
Large Eddy Simulation Model ◽  
Wide Range

The flow-induced vibration of a pipe is an important issue in various engineering fields, and this phenomenon is widely observed in nuclear power plants. Although turbulent structures play important roles in the velocity and pressure fields in a pipe, only a few studies have been conducted on the turbulent flow on an oscillating wall. In this study, direct numerical simulations were conducted to establish a large eddy simulation model for a turbulent flow on an oscillating wall and scrutinize the energy transfer between the grid scale (GS) and sub-grid scale (SGS). Although energy is generally transferred from the GS to SGS (forward scatter), it is likely that energy is transferred from the SGS to GS (backward scatter) under specific conditions. The present numerical results indicated that backward scatter exists in the production term in the case of a static wavy wall. On the other hand, such backward scatter could not be observed in the case of an oscillating wall. It is well known that separated flows and backward flows are generated behind the crest. Stronger backward flows accelerate the main flow and enhance the velocity gradients in a wide range behind the crest. In the case of an oscillating wall, the development of separated flow is immature because the shape of the wall is not fixed. Eventually, the backward scatter is deemed to be suppressed.

Systematic Evaluation of Creep-Fatigue Life Prediction Methods for Various Alloys

2009 ◽  
Author(s):  
Yukio Takahashi ◽  
Bilal Dogan ◽  
David Gandy
Keyword(s):  
Power Generation ◽  
Nuclear Power ◽  
Power Plants ◽  
Strain Range ◽  
Creep Damage ◽  
Small Strain ◽  
Systematic Evaluation ◽  
Wide Range ◽  
Creep Fatigue ◽  
Different Temperatures

Failure under creep-fatigue interaction is receiving increasing interest due to an increased number of start-up and shut-down in fossil power generation plants as well as development of newer nuclear power plants employing low-pressure coolant. These situations have promoted the development of various approaches for evaluating its significance. However, most of them are fragment and rather limited in terms of materials and test conditions they covered. Therefore applicability of the proposed approaches to different materials or even different temperatures is uncertain in many cases. The present work was conducted in order to evaluate and compare the representative approaches used in the prediction of failure life under creep-fatigue conditions as well as their modifications, by systematically applying them to available test data on a wide range of materials which have been used or are planned to be used in various types of power generation plants. The following observations have been made from this exercise. (i) Time fraction model has a tendency to be unconservative in general, especially at low temperature and small strain range. Because of the large scatter of the total damage, this shortcoming would be difficult to cover by the consideration of creep-fatigue interaction in a fixed manner. (ii) Classical ductility exhaustion model showed a common tendency to be overly conservative in many situations, especially at small strain ranges. (iii) The modified ductility exhaustion model based on the re-definition of creep damage showed improved predictability with a slightly unconservative tendency. (iv) Energy-based ductility exhaustion model developed in this study seems to show the best predictability among the four procedures in an overall sense although some dependency on strain range and materials was observed.

Small Diameter Effects on Internal Flow Boiling

2001 ◽  
Author(s):  
Gail E. Kendall ◽  
Peter Griffith ◽  
Arthur E. Bergles ◽  
John H. Lienhard
Keyword(s):  
Heat Transfer ◽  
Thermal Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Instability ◽  
Flow Boiling ◽  
Small Diameter ◽  
Internal Flow ◽  
Two Phase ◽  
Wide Range

Abstract Since the 1950’s, the research and industrial communities have developed a body of experimental data and set of analytical tools and correlations for two-phase flow and heat transfer in passages having hydraulic diameter greater than 6 mm or so. These tools include flow regime maps, pressure drop and heat transfer correlations, and critical heat flux limits, as well as strategies for robust thermal management of HVAC systems, electronics, and nuclear power plants. Designers of small systems with thermal management by phase change will need analogous tools to predict and optimize thermal behavior in the mesoscale and smaller sizes. Such systems include a wide range of devices for computation, measurement, and actuation in environments that range from office space to outer space and living systems. This paper examines important proceses that must be considered when channel diameters decrease, including flow distribution issues in single, parallel, and split flows; flow instability in parallel passages; manufacturing tolerances effects; nucleation processes; and wall conductance effects. The discussion focuses on engineering issues for the design of practical systems.

Experimental Investigation on Heat Transfer for Two-Phase Flow Under Natural Convection

2012 ◽  
Author(s):  
Milan Amižić ◽  
Estelle Guyez ◽  
Jean-Marie Seiler
Keyword(s):  
Heat Transfer ◽  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Liquid Pool ◽  
Severe Accident ◽  
Bottom Plate ◽  
Interfacial Region ◽  
Two Phase ◽  
Wide Range

In the frame of severe accident research for the second and the third generation of nuclear power plants, some aspects of the concrete cavity ablation during the molten corium–concrete interaction are still remaining issues. The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt-through. For the purpose of experimental investigation of thermal-hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched jointly by CEA, EDF, IRSN, GDF-Suez and SARNET. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the smallest pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s. This paper presents some preliminary conclusions deduced from the experiments which involve a liquid pool with the gas injection only from the bottom plate. A comparison with existing models for the assessment of heat transfer has also been carried out.

scholarly journals Power Law Behavior and Self-Similarity in Modern Industrial Accidents

2015 ◽  
Vol 25 (01) ◽  
pp. 1550004 ◽  
Author(s):  
António M. Lopes ◽  
J. A. Tenreiro Machado
Keyword(s):  
Power Law ◽  
Nuclear Power ◽  
Statistical Physics ◽  
Power Plants ◽  
Data Series ◽  
Industrial Accident ◽  
Industrial Accidents ◽  
Wide Range ◽  
Accident Data ◽  
Human Operators

Advances in technology have produced more and more intricate industrial systems, such as nuclear power plants, chemical centers and petroleum platforms. Such complex plants exhibit multiple interactions among smaller units and human operators, rising potentially disastrous failure, which can propagate across subsystem boundaries. This paper analyzes industrial accident data-series in the perspective of statistical physics and dynamical systems. Global data is collected from the Emergency Events Database (EM-DAT) during the time period from year 1903 up to 2012. The statistical distributions of the number of fatalities caused by industrial accidents reveal Power Law (PL) behavior. We analyze the evolution of the PL parameters over time and observe a remarkable increment in the PL exponent during the last years. PL behavior allows prediction by extrapolation over a wide range of scales. In a complementary line of thought, we compare the data using appropriate indices and use different visualization techniques to correlate and to extract relationships among industrial accident events. This study contributes to better understand the complexity of modern industrial accidents and their ruling principles.

Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Load ◽  
Severe Accident ◽  
Steel Columns ◽  
Wide Range ◽  
Buckling Failure ◽  
The Difference ◽  
Accident Conditions

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

Risk Importance Measures for Maintenance in Non-Coherent Systems

2010 ◽  
Author(s):  
Ye Tao ◽  
Lixuan Lu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Weak Links ◽  
Coherent System ◽  
Level Control ◽  
Coherent Systems ◽  
Component Importance ◽  
Wide Range ◽  
Undesired Event ◽  
Importance Analysis

Probabilistic Safety Assessment (PSA) has been an important tool to assist Nuclear Power Plants (NPPs) management over the years. Through PSA, the weak links of the whole system can be identified by component importance measures. The importance measures can be classified according to risk significance and safety significance. They signify the role that each component plays in either causing or contributing to the occurrence of an undesired event. A wide range of importance measures have been developed over the years and most of them are geared towards coherent systems. Importance analysis of non-coherent system is rather limited. In this paper, a set of component importance measures for non-coherent systems is analyzed and investigated. The comparison and the selection of the most informative and appropriate measures for guiding the maintenance of a system are presented. A steam generator level control system of NPP is used to demonstrate the application of the results.

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