scholarly journals Revealing deformation of segments and their supports in a hydrostatic segmental bearing

2021 ◽  
Vol 4 (7(112)) ◽  
pp. 33-40
Author(s):  
Vladimir Nazin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Experimental Studies ◽  
Axial Direction ◽  
Working Fluid ◽  
Radial Clearance ◽  
Maximum Deformation ◽  
Feed Pressure ◽  
Starting Point ◽  
Study Results

At present, there are theoretical and experimental studies of such bearings without taking into account the elastic deformation of the bearing segments. The rotor bearings of powerful turbines at nuclear power plants are subjected to loads as high as tens of tons. One of the important issues in designing segmental bearings operating under these conditions consists in taking into account elastic deformations of the segments. A schematic diagram of a segmental hydrostatic bearing was presented and the principle of its operation was described. When determining the deformation of spherical support, a formula of change in volume of a solid steel ball subjected to uniform pressure was applied. To determine the segment deformation in the axial direction, differential equation of bending of the strip beam as the initial one. The basic equation of deformation of rods with a curved axis acting in the plane of curvature was taken as a starting point of determining the segment deformation in the circumferential direction. It was found in the studies that the maximum deformation of the segment is 4.5 % of radial clearance at a feed pressure of 5 MPa and can affect the bearing characteristics. A substantially nonlinear character of deformations along the segment axis was revealed. It was found that the pressure of the working fluid significantly affects the segment thickness. With an increase in feeding pressure from 1 MPa to 10 MPa, the thickness of the steel segment increased more than 2 times and the thickness of the bronze segment increased more than 3 times. It was established that the pressure of the working fluid exceeding 10 MPa substantially affects the deformation of the spherical support and the bearing clearance. The study results will make it possible to determine more accurately the main characteristics of the segmental bearing and design it more efficiently.

Application of Supercritical Fluids in Thermal- and Nuclear-Power Engineering

2021 ◽  
pp. 601-658
Author(s):  
Igor L. Pioro
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Rankine Cycle ◽  
Working Fluid ◽  
Thermal Power Plants ◽  
Power Cycles

Supercritical Fluids (SCFs) have unique thermophyscial properties and heat-transfer characteristics, which make them very attractive for use in power industry. In this chapter, specifics of thermophysical properties and heat transfer of SCFs such as water, carbon dioxide, and helium are considered and discussed. Also, particularities of heat transfer at Supercritical Pressures (SCPs) are presented, and the most accurate heat-transfer correlations are listed. Supercritical Water (SCW) is widely used as the working fluid in the SCP Rankine “steam”-turbine cycle in fossil-fuel thermal power plants. This increase in thermal efficiency is possible by application of high-temperature reactors and power cycles. Currently, six concepts of Generation-IV reactors are being developed, with coolant outlet temperatures of 500°C~1000°C. SCFs will be used as coolants (helium in GFRs and VHTRs, and SCW in SCWRs) and/or working fluids in power cycles (helium, mixture of nitrogen (80%) and helium (20%), nitrogen and carbon dioxide in Brayton gas-turbine cycles, and SCW/“steam” in Rankine cycle).

Efficient Intensity Measures for Probabilistic Seismic Response Analysis of Anchored Above-Ground Liquid Steel Storage Tanks

2016 ◽  
Author(s):  
Hoang Nam Phan ◽  
Fabrizio Paolacci
Keyword(s):  
Ground Motion ◽  
Earthquake Engineering ◽  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Time History ◽  
Response Analysis ◽  
Storage Tanks ◽  
Intensity Measures ◽  
Study Results

Liquid storage tanks are vital lifeline structures and have been widely used in industries and nuclear power plants. In performance-based earthquake engineering, the assessment of probabilistic seismic risk of structural components at a site is significantly affected by the choice of ground motion intensity measures (IMs). However, at present there is no specific widely accepted procedure to evaluate the efficiency of IMs used in assessing the seismic performance of steel storage tanks. The study presented herein concerns the probabilistic seismic analysis of anchored above-ground steel storage tanks subjected to several sets of ground motion records. The engineering demand parameters for the analysis are the compressive meridional stress in the tank wall and the sloshing wave height of the liquid free surface. The efficiency and sufficiency of each alternative IM are quantified by results of time history analyses for the structural response and a proper regression analysis. According to the comparative study results, this paper proposes the most efficient and sufficient IMs with respect to the above demand parameters for a portfolio of anchored steel storage tanks.

scholarly journals Analyses of the Control System Strategies and Methodology for Part Power Control of the Simple and Intercooled Recuperated Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
A. Gad-Briggs ◽  
P. Pilidis ◽  
T. Nikolaidis
Keyword(s):  
Control System ◽  
Inventory Control ◽  
Nuclear Power ◽  
Power Plants ◽  
Control Method ◽  
Control Strategies ◽  
Limiting Factors ◽  
Cycle Performance ◽  
Study Results ◽  
And Performance

An important requirement for Generation IV Nuclear Power Plant (NPP) design is the control system, which enables part power operability. The choices of control system methods must ensure variation of load without severe drawbacks on cycle performance. The objective of this study is to assess the control of the NPP under part power operations. The cycles of interest are the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). Control strategies are proposed for NPPs but the focus is on the strategies that result in part power operation using the inventory control method. First, results explaining the performance and load limiting factors of the inventory control method are documented; subsequently, the transient part power performances are also documented. The load versus efficiency curves were also derived from varying the load to understand the efficiency penalties. This is carried out using a modeling and performance simulation tool designed for this study. Results show that the ICR takes ∼102% longer than the SCR to reduce the load to 50% in design point (DP) performance conditions for similar valve flows, which correlates with the volumetric increase for the ICR inventory tank. The efficiency penalties are comparable for both cycles at 50% part power, whereby a 22% drop in cycle efficiency was observed and indicates limiting time at very low part power. The analyses intend to aid the development of cycles for Generation IV NPPs specifically gas cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs), where helium is the coolant.

Dynamic Processes in Vacuum Contact Devices of Robots for Vertical Motion in the Water Environment

2019 ◽  
Vol 20 (7) ◽  
pp. 417-421
Author(s):  
V. G. Gradetsky ◽  
M. M. Knyazkov ◽  
E. A. Semenov ◽  
A. N. Sukhanov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Experimental Studies ◽  
Vertical Motion ◽  
Contact Device ◽  
Dynamic Processes ◽  
Simplified Models ◽  
Contact Devices ◽  
Air Ejector

The results of experimental investigation intended to improve movement conditions for pneumatic robots on vertical surfaces under water are discussed. Features of the movement of vacuum contact devices for the simulation of mathematical model of the vacuum contact device with surfaces under water are presented. The experimental studies made it possible to obtain additional data on the dynamics of attachment, to obtain transient processes for air-water flow through ejector and to correct the results obtained earlier. For the purpose of analytical study of dynamic processes occurring in the system of vacuum contact devices, and taking into account the complexity of the description of nonlinearities, linearized simplified models of the system "air ejector — contact device — water environment" were developed. Vacuum contact devices are designed to provide guaranteed contact with vertical surfaces, plane slopes or horizontal surfaces on which the underwater robot performs its movement, carrying out the prescribed technological tasks, for example, in dry wells of nuclear power plants, on the surfaces of ship hulls, on the surfaces of underwater structures. The models took into account the forces of adhesion to the surfaces under water — the forces from the pressure drop, the friction force, the contact and vacuum interaction, the elasticity of suction caps. As a result of the solution of the model problem, the values of mechanical parameters, as well as the values of vacuum and flow in the cavity of variable volume as functions of changing the gap between the end of the corrugated membrane and the surfaces are obtained explicitly. As a result of the study of dynamic processes occurring in simplified models of vacuum contact devices "air ejector — contact surface — water environment", the transient characteristics of the change in the operating forces and pressures over time, as well as the dependence of the normal and tangential components of the forces on the depth of immersion in water were obtained. The variants of the designs of vacuum contact devices with surfaces in the water environment are investigated, and the modernization of the laboratory test bench for testing vacuum contact devices under water is carried out.

The Effect of Nanoscale Surface Modification on Boiling Heat Transfer and Critical Heat Flux

2010 ◽  
Author(s):  
Moo Hwan Kim
Keyword(s):  
Heat Transfer ◽  
Boiling Heat Transfer ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Boiling ◽  
Heated Surface ◽  
Pool Boiling ◽  
Working Fluid ◽  
Wetting Ability ◽  
Modified Surface

Recently, there were lots of researches about enormous CHF enhancement with the nanofluid in pool boiling and flow boiling. It is supposed the deposition of nanoparticles on the heated surface is one of main reasons. In a real application, nanofluid has a lot of problems to be used as the working fluid because of sedimentation and aggregation. The artificial surfaces on silicon and metal were developed to have the similar effect with nanoparticles deposited on the surface. The modified surface showed the enormous ability to increase CHF in pool boiling. Furthermore, under flow boiling, it had also good results to increase CHF. In these studies, we concluded that wetting ability of surface; e.g. wettability and liquid spreading ability (hydrophilic property of surface) was a key parameter to increase CHF under both pool and flow boiling. In addition, using wettability difference of surface; e.g. hydrophilic and hydrophobic, we conducted some tests of BHT (boiling heat transfer) enhancement using the oxide silicon which have micro-sized hydrophobic islands on hydrophilic surface. By using both of these techniques, we propose an optimized surface to increase both CHF and BHT. Also, the fuel surface of nuclear power plants is modified to have same effect and the results shows a good enhancement of CHF, too.

Experimental and Numerical Simulation on Hydraulic Behavior of Molten Flow in the Lower Part in Reactor Core

2016 ◽  
Author(s):  
Kota Matsuura ◽  
Hideaki Monji ◽  
Susumu Yamashita ◽  
Hiroyuki Yoshida
Keyword(s):  
Numerical Simulation ◽  
Liquid Film ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Behavior ◽  
Reactor Core ◽  
Video Camera ◽  
Working Fluid ◽  
Simulation Code

In the decommissioning work of nuclear power plants, it is important to grasp the sedimentation place of molten materials. However, the technique to grasp exactly sedimentation place is not established now. Therefore, the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior is developed. In the study, visualization experiment and numerical simulation were performed to validate the applicability of the JUPITER to the hydraulic relocation behavior in core internals. The test section used in this experiment simulated the structure of the core internals, such as a control rod and a fuel support piece, simply. The working fluid is water under the atmospheric pressure. The experiment uses a high-speed video camera to visualize the flow behavior. The behavior and the speed of the liquid film in a narrow flow channel is obtained. For the numerical analysis carried out prior to the experiment, the behavior of flow down liquid was shown. The typical behavior was also observed that the tip of a liquid film flowing down splits into.

scholarly journals Performance Analyses and Evaluation of Co2 and N2 as Coolants in a Recuperated Brayton Gas Turbine Cycle for a Generation IV Nuclear Reactor Power Plant

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Emmanuel O. Osigwe ◽  
Arnold Gad-Briggs ◽  
Theoklis Nikolaidis ◽  
Pericles Pilidis ◽  
Suresh Sampath
Keyword(s):  
Power Plant ◽  
Gas Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Critical Conditions ◽  
Working Fluid ◽  
Closed Cycle ◽  
Gas Turbine Cycle

Abstract As demands for clean and sustainable energy renew interests in nuclear power to meet future energy demands, generation IV nuclear reactors are seen as having the potential to provide the improvements required for nuclear power generation. However, for their benefits to be fully realized, it is important to explore the performance of the reactors when coupled to different configurations of closed-cycle gas turbine power conversion systems. The configurations provide variation in performance due to different working fluids over a range of operating pressures and temperatures. The objective of this paper is to undertake analyses at the design and off-design conditions in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as the working fluid in the cycle. The analysis is demonstrated using an in-house tool, which was developed by the authors. The results show that the choice of working fluid controls the range of cycle operating pressures, temperatures, and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions. The analyses intend to aid the development of cycles for generation IV nuclear power plants (NPPs) specifically gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs).

Experimental Studies of Radiation-Protective Properties of a Modified Titanium Hydride

2022 ◽  
Vol 1049 ◽  
pp. 174-179
Author(s):  
A.A. Karnauhov ◽  
R.N. Yastrebinskii
Keyword(s):  
Gamma Radiation ◽  
Dose Rate ◽  
Titanium Hydride ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Experimental Studies ◽  
Point Sources ◽  
Relaxation Length ◽  
Protective Properties

The results of experimental studies of the protective properties of titanium hydride with respect to neutron and gamma radiation in order to determine the optimal conditions for their use in the composition of the structural radiation protection of the nuclear reactor are presented. The weakening of the basic functionals in the thickness of protection, including the density of fast, intermediate and thermal neutrons, and the dose rate of gamma radiation is established. The functions of weakening the density of neutron flow and the dose rate of gamma radiation are measured in the conditions of "barrier" geometry. Determination of the protective properties of the structure was carried out when the modified titanium hydride fraction was placed in aluminum containers with a filling coefficient of a volume of container 0.63. The relaxation lengths for all neutron groups are close and on average are 9.8 cm. The functions of weakening the dose rate of gamma radiation of point sources Cs-137 and Co-60 are exponential. The weakening of radiation occurs with a constant relaxation length. For energy 0.661 MeV, the relaxation length is 7.1 cm, for energy 1.25 MeV, the relaxation length is equal to 10.1 cm. On the basis of the experimental studies, the high efficiency of the modified fraction of titanium hydride was confirmed during its use in protecting nuclear power plants.

HYDROGEN-OXYGEN STEAM GENERATOR FOR A CLOSED HYDROGEN COMBUSTION CYCLE

2018 ◽  
pp. 68-79
Author(s):  
R. Z. Aminov ◽  
A. N. Egorov
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Combustion Products ◽  
Working Fluid ◽  
Hydrogen Energy ◽  
Energy Complex ◽  
Steam Turbine Plant ◽  
Oxygen Medium

The paper analyzes the problems of combustion hydrogen in an oxygen medium for produce high-temperature steam that can be used to produce electricity at various power plants. For example, at the nuclear power plants, the use of a H2-O2 steam generator as part of a hydrogen energy complex makes it possible to increase its power and efficiency in the operational mode due to steam-hydrogen overheating of the main working fluid of a steam-turbine plant. In addition, the use of the hydrogen energy complex makes it possible to adapt the nuclear power plants to variable electric load schedules in conditions of increasing the share of nuclear power plants and to develop environmentally friendly technologies for the production of electricity. The paper considers a new solution of the problem of effective and safe use of hydrogen energy at NPPs with a hydrogen energy complex.Technical solutions for the combustion of hydrogen in the oxygen medium using direct injection of cooling water or steam in the combustion products have a significant drawback – the effect of “quenching” when injecting water or water vapor which leads to a decrease in the efficiency of recombination during cooling of combustion products that is expressed in an increase fraction of non-condensable gases. In this case, the supply of such a mixture to the steam cycle is unsafe, because this can lead to a dangerous increase in the concentration of unburned hydrogen in the flowing part of the steam turbine plant. In order to solve this problem, the authors have proposed a closed hydrogen cycle and a hydrogen vapor overheating system based on it, and carried out a study of a closed hydrogen combustion system which completely eliminates hydrogen from entering the working fluid of the steam cycle and ensures its complete oxidation due to some excess of circulating oxygen.The paper considers two types of hydrogen-oxygen combustion chambers for the system of safe generating of superheated steam using hydrogen in nuclear power plant cycle by using a closed system for burning hydrogen in an oxygen medium. As a result of mathematical modeling of combustion processes and heat and mass transfer, we have determined the required parameters of a hydrogen-oxygen steam generator taking into account the temperature regime of its operation, and a power range of hydrogen-oxygen steam generators with the proposed combustion chamber design.

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