scholarly journals Development and Testing of a Methodology for Assessing of the Correlation Velocity Measurements’ Accuracy for the Hydrodynamic Investigations of the Turbulent Coolant Flow in Nuclear Reactor Elements

2021 ◽  
Vol 12 (1) ◽  
pp. 67-74
Author(s):  
I. A. Konovalov ◽  
A. A. Chesnokov ◽  
A. A. Barinov ◽  
S. M. Dmitriev ◽  
A. E. Khrobostov ◽  
...  
Keyword(s):  
Confidence Interval ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Spectral Power ◽  
Correlation Method ◽  
Spectral Power Density ◽  
Coolant Flow ◽  
Measuring System ◽  
Transport Time ◽  
Passive Impurity

The correlation method of the coolant flow measuring is widely used in research practice including for studying of turbulent coolant flows in scale models of elements of nuclear power plants. The aim of this work was to develop a technique for assessing the effect of noise recorded by a measuring system on the flow rate readings obtained using the correlation method.A technique to assess the effect of noise as well as the relative position and acquisition period of sensors is presented. An insignificant concentration of a salt solution (NaCl or Na2SO4 ) is used as a passive impurity which creates a conductivity gradient of the medium recorded by a conductometric system. Turbulent pulsations at the interface between two concurrent isokinetic flows in a channel with a square cross section are used as the signal source for the correlational algorithm.Paper presents the values of  the  turbulence′s  transport  time  between  spatial  conductometers, the results of estimating the spectral power density and band of the recorded signal and also the signalto-noise ratios of the measuring system obtained on their basis which are subsequently used to estimate the confidence interval of the transport time.As a result of measurements the relationship between the confidence  interval  value  and  the signal length were obtained. The measurements which were carried out at different relative positions of conductometers make it  possible  to  make  a  conclusion  about  an  increase  in  the  spectral  width of the signal and, as a consequence, a decrease in the length of the confidence interval with increasing of distance between sensors.The presented work is an approbation of this approach for its application as part of an experimental model of a nuclear reactor in order to determine per-channel flow rates in the channels of the core simulator using mesh conductometric sensors taking into account the effect of noise.

scholarly journals Application of the Correlation Velocity Measurements for Hydrodynamic Investigations of Turbulent Coolant Flow in Nuclear Reactor Elements

2020 ◽  
Vol 11 (3) ◽  
pp. 196-203
Author(s):  
I. A. Konovalov ◽  
A. E. Khrobostov ◽  
M. A. Legchanov ◽  
D. N. Solncev ◽  
A. A. Barinov ◽  
...  
Keyword(s):  
Flow Rate ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Correlation Method ◽  
Coolant Flow ◽  
Wire Mesh ◽  
Coolant Flow Rate ◽  
Test Facility ◽  
Test Model ◽  
Measured Velocity

The method of correlation measurement of the coolant flow rate, widely used for operational diagnostics of nuclear power plants, can be extensively used in research practice. The aim of this work was to apply a correlation method based on the conductometric measurement system with wire-mesh sensors for measuring a coolant flow rate.Insignificant concentration of a salt solution (NaCl or Na2SO4 ) creates a gradient of the conductivity in the flow, which is used as a passive scalar measured by the system. Authors used turbulent pulsations at the interface of two concurrent flows with identical velocities in a square channel as a signal source for the correlation method. The paper presents the methodology of the tests, test facility description, signalto-noise ratio estimation, the results of digital signal processing and comparison of the measured velocities in the model with the flowrate‒averaged velocity determined by the use of flowmeters. The measured velocity values give acceptable agreement for the turbulent flow modes. It was shown that the measurement accuracy drops sharply for low-Reynolds flows.The obtained results were used for flowrate measurements in core-imitator channels of the nuclear reactor test model. The presented paper is an approbation of this approach for its application as part of an test model of a nuclear reactor in order to determine the each duct flow rates in the channels of the core simulator using wire mesh sensors.

scholarly journals Application of the Correlation Measurement Method for Reconstructing of the Velocity Profile with Spatial and Temporal Discretization in Studies of the Hydrodynamics of Turbulent Flows Based on the Matrix Conductometry Method

2021 ◽  
Vol 12 (4) ◽  
pp. 292-300
Author(s):  
S. M. Dmitriev ◽  
A. E. Khrobostov ◽  
D. N. Solncev ◽  
A. A. Barinov ◽  
A. A. Chesnokov ◽  
...  
Keyword(s):  
Data Collection ◽  
Turbulent Flows ◽  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Correlation Method ◽  
Measuring System ◽  
Reliability Of Results ◽  
The Matrix ◽  
Influence Of Noise

The correlation method for measuring of the coolant fl rate is used in the operation of nuclear power plants and is widespread in research practice including study of turbulent fl    hydrodynamics. However the question of its applicability and possibilities in studies using the matrix conductometry method remains open. Earlier the algorithm for determining of the correlation fl rate using a conductometric measuring system was highlighted and the error of the results obtained was estimated and the dependence of the influence of noise and the time of data collection on the reliability of results was investigated. These works were carried out using two independent mesh sensors and the issue of the resolution of local velocity components was not covered. The purpose of this work was to test the correlation method for measuring velocity with temporal and spatial sampling using two-layer mesh conductometric sensors.As the result velocity cartograms were obtained over the cross-section of the experimental model with quasi-stationary mixing and the value of the average flow rate is in good agreement with the values obtained from the standard flow meters of the stand. Also measurements were carried out at a non-stationary setting of the experiment and realizations of the flow rate and velocity components of the flow at the measuring points were obtained.Analysis of the obtained values allows to conclude about the optimal data collection time for correlation measurements, as well as the reliability of results.

scholarly journals Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

2008 ◽  
Vol 2008 ◽  
pp. 1-9 ◽  
Author(s):  
Enrico Zio ◽  
Francesco Di Maio
Keyword(s):  
Confidence Interval ◽  
Order Statistics ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Fuel Cladding ◽  
Complete Group ◽  
Bootstrap Technique ◽  
Safety Margins ◽  
Group Distribution

In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

scholarly journals The Changes of Ergonomic Engine Vibroacoustic Response Regarding Their Development

Energies ◽  
2021 ◽  
Vol 14 (14) ◽  
pp. 4215
Author(s):  
Radosław Wróbel ◽  
Lech Sitnik ◽  
Monika Andrych-Zalewska ◽  
Łukasz Łoza ◽  
Radostin Dimitrov ◽  
...  
Keyword(s):  
Human Health ◽  
Power Density ◽  
High Speed ◽  
Internal Combustion Engines ◽  
Spectral Power ◽  
Spectral Power Density ◽  
Steering Wheel ◽  
Combustion Engines ◽  
Head Restraint ◽  
Vibroacoustic Response

The article presents the results of research on the vibroacoustic response of internal combustion engines mounted in a vehicle. The vehicles studied belong to popular models, which became available in successive versions. Each group included vehicles of the same model of an older generation (equipped with a naturally aspirated engine) and of a newer generation, including downsized (and turbocharged) engines. Tests in each group were carried out under repeatable conditions on a chassis-load dynamometer. The vibrations were measured using single-axis accelerometers mounted on the steering wheel, engine, and driver’s head restraint mounting. The primary purpose of the study was to verify whether the new generations of vehicles equipped with additional high-speed elements (compressors) generate additional harmonics (especially those within the range potentially affecting travel comfort and human health) and whether there are significant changes in the distribution of spectral power density in the new generations. As the study showed, new generations of vehicles are characterized by a different vibroacoustic response, and the trend of change is the same in each of the families studied.

Measurement of the Pulverized Coal Concentration by Measuring the Time Delay of Ultrasonic with Digital Correlation Method

2013 ◽  
Vol 333-335 ◽  
pp. 359-364
Author(s):  
Ke Wang
Keyword(s):  
Time Delay ◽  
Power Plant ◽  
Correlation Method ◽  
The State ◽  
Pulverized Coal ◽  
Measuring System ◽  
Digital Correlation ◽  
Coal Boiler ◽  
Key Parameter

The pulverized coal concentration in blast pipe in front of firebox is one key parameter in coal boiler used in a power plant; it will affect the state of burning in firebox. This paper proposes a new digital correlation method to measure the time delay of ultrasonic for measuring the pulverized coal concentration. The principle of measurement is discussed in detail and the measuring system is designed.

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.

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

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.

scholarly journals Assessment of water hammer effects on boiling water nuclear reactor core dynamics

2007 ◽  
Vol 22 (1) ◽  
pp. 18-33 ◽  
Author(s):  
Anis Bousbia-Salah
Keyword(s):  
Pressure Wave ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Pressure Wave Amplitude ◽  
Sensitivity Studies ◽  
Rapid Transients ◽  
Complex Phenomena

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

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