scholarly journals SIMULATION OF THERMAL PROCESSES OF A NPP STEAM GENERETOR FOR INFORMATION TECHNOLOGY OPTIMIZED CONTROL

2021 ◽  
pp. 56-61
Author(s):  
Olena Nikulina ◽  
Valerii Severyn ◽  
Nina Kotsiuba ◽  
Anton Bubnov
Keyword(s):  
Heat Transfer ◽  
Information Technology ◽  
Differential Equations ◽  
Steam Generator ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Feed Water ◽  
Thermal Processes ◽  
State Variables ◽  
System Of Differential Equations

Mathematical models of thermal processes in the form of Cauchy in the state space with relative variables of the steam generator PGV-1000 of the power unit of a nuclear power plant with a nuclear reactor VVER-1000 have been developed for the using of models in information technology for optimizing the control of a steam generator. The working thermal processes in the PGV-1000 steam generator associated with the supply of feed water to it from the water treatment system and the coolant from the nuclear reactor and the removal of vapors to the main steam header are considered. The design diagram of the steam generator is presented, which reflects the working processes in it under the evaporation mirror and above it. On the basis of differential equations of the heat balance of the heat carrier in the steam generator and in the metal heat exchange tubes, the simulation of heat transfer from the heat carrier to the feed water in the steam generator is carried out. The heat transfer model in the form of a linear system of differential equations in relative state variables is developed. The processes of vaporization during heating of feed water by the heat transfer surface are considered. Differential equations of material and heat balances of dynamic processes of vaporization in a steam generator are compiled, which are not equations in the Cauchy form. Transformations of the differential equations of material and heat balances in the steam generator to the Cauchy form are carried out. A nonlinear system of differential equations for the balance of vaporization in relative state variables is obtained. The values of the constant parameters of the models for the steam generator PGV-1000 are calculated. The mathematical model of thermal processes in the PGV-1000 steam generator, which is presented in the form of a system of differential equations and includes the processes of heat transfer and steam generation, will make it possible to identify and optimize the steam generator control system with the help of information optimization technology. Keywords: nuclear power plant, steam generator, thermal processes, mathematical model, differential equations, optimization, control, information technology.

MODELS OF WWER-1000 NUCLEAR REACTOR WITH DIVISION INTO ZONES ON VERTICAL AXIS FOR INFORMATION TECHNOLOGY OF CONTROL

2021 ◽  
Vol 4 ◽  
pp. 105-116
Author(s):  
Valeriy Severyn ◽  
◽  
Elena Nikulina ◽  
◽  
Keyword(s):  
Information Technology ◽  
Differential Equations ◽  
Mathematical Models ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Control Object ◽  
Vertical Axis ◽  
State Variables ◽  
Relative State ◽  
Systems Of Differential Equations

Mathematical models of the WWER-1000 nuclear power reactor have been developed with division into zones along the vertical axis in the form of nonlinear systems of differential equations with dimensionless relative state variables. Models in a given number of zones along the vertical axis represent neutron kinetics, gradual heat release, thermal processes in fuel, cladding and coolant, changes in the concentration of iodine, xenon and boron. The parameters of mathematical models have been calculated based on the design and technological parameters of the V-320 series nuclear reactor. A general model of the reactor as a control object with division into zones along the vertical axis, as well as models with control of absorbing rods and boric acid, are obtained. Integration of the obtained systems of differential equations for given initial conditions allows one to obtain changes in all state variables in the reactor zones along the vertical axis. In particular, from the change in power in the zones along the vertical axis, the axial offset is calculated as the relative value of the difference between the powers of the upper and lower halves of the reactor core. The developed reactor models with dimensionless relative state variables use a minimum number of calculations, allow calculating the change in the axial offset, and are included in the information technology for controlling the power units of nuclear power plants to optimize the maneuvering modes of the WWER-1000 V-320 series reactor.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

Thermal Hydraulic Studies on Helical Coil Steam Generator by CFD

2011 ◽  
Author(s):  
Sooyun Joh
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Reactor ◽  
Complex Geometry ◽  
Transport Model ◽  
Cfd Modeling ◽  
Helical Coil ◽  
Power Module ◽  
Flux Boundary Condition ◽  
Nuclear Core

NuScale Power, Inc. is commercializing a 45 Megawatt electric light water nuclear reactor NuScale Power Module (NPM). Each NPM includes a containment vessel, a reactor vessel, a nuclear reactor core, an integral steam generator, and an integral pressurizer. The NuScale Power Module is cooled by natural circulation. The primary coolant in the Reactor Pressure Vessel is heated in the nuclear core, it rises through a central riser, it spills over and encounters the helical coil steam generator, it is cooled as steam is generated inside the steam generator, and it is again heated in the nuclear core. The Steam Generator also must be designed to provide adequate heat transfer, to allow adequate primary reactor coolant flow, and to provide adequate steam flow to produce the required power output. This paper presents the CFD results that describe the transport phenomena on the heat transfer and fluid flow dynamics in helical coil steam generator tubes. The ultimate goal of the CFD modeling is to predict the steam outlet conditions associated with the chosen helical coil tube geometries, solving the primary and secondary flow region together coupled with the helical coil tube. However, current studies are focused on the primary side with the heat flux boundary condition assigned on the outer surface of the helical coil steam generator. In this study, the ANSYS CFX v. 12.1 [1] was used to solve the three-dimensional mass, momentum and energy equations. The helical coil steam generator has complex geometry and modeling entire geometry requires the enormous memory that is beyond our hardware capability and is not practical. Therefore, geometry was limited to 1 degree of the wedge and 5% of the total length in the middle. Only external flow, single phase flow around the helical coils, is simulated using the standard k-ε model and shear stress transport model. From the results of the numerical simulation, the pressure drop and temperature profiles were determined. It is important to understand thermal hydraulic phenomena for the design and performance prediction of the reactor internal.

Modeling and Simulation of a Heat Recovery Steam Generator Using Partially Known Design Point Data

2013 ◽  
Vol 845 ◽  
pp. 596-603
Author(s):  
Mesfin G. Zewge ◽  
T.A. Lemma ◽  
A.A. Ibrahim ◽  
D. Sujan
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Performance Optimization ◽  
Heat Recovery ◽  
Conservation Equation ◽  
Feed Water ◽  
Heat Recovery Steam Generator ◽  
Transfer Coefficients ◽  
Design Point ◽  
Point Data

In a cogeneration or combined heat and power plant, a heat recovery steam generator (HRSG) helps achieve overall thermal efficiency as high as 80%. The purpose of this study is to model and simulate the HRSG given partial design point data. The pinch and approach temperatures are optimized within generally accepted range. In order to satisfy the energy conservation equation, tuning parameters are used for the overall heat transfer coefficients corresponding to the evaporator and economizer. For the off-design simulation, the values of pinch and approach temperatures are adjusted until the modeling error is within a set limit. The effect of mass flow rate on the heat transfer coefficient is accounted for & by employing empirical relations. A 12 Ton/hr natural circulation HRSG was considered as a case study. The validation test on inlet temperatures of the exhaust gas and feed water to the economizer demonstrated relative percentage errors of 0.4246% and 1.8776%, respectively. The model can be used for fault detection and diagnostic system design, performance optimization, and environmental load assessment.

Effects of Various Parameters on Supercritical-Pressure Heat Transfer and Limits for Heat Transfer Correlations Obtained in Vertical Bare Tubes

2017 ◽  
Author(s):  
Hakim Maloufi ◽  
Hanqing Xie ◽  
Andrew Zopf ◽  
William Anderson ◽  
Christian Langevin ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Heat Fluxes ◽  
Operating Conditions ◽  
Transfer Coefficients ◽  
Wide Range ◽  
Heat Transfer Correlations ◽  
Bare Tube

Currently, there is a number of Generation-IV SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts under development worldwide. These high temperature and pressure reactors will have significantly higher operating parameters compared to those of current water-cooled nuclear-power reactors (i.e., “steam” pressures of about 25 MPa and “steam” outlet temperatures up to 625 °C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated, as the steam will be flowing directly to a steam turbine. In support of developing SCWRs studies are being conducted on heat transfer at SuperCritical Pressures (SCPs). Currently, there are very few experimental datasets for heat transfer at SCPs in power-reactor fuel bundles to a coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations developed with bare-tube data can be used as a conservative approach. Selected empirical heat-transfer correlations, based on experimentally obtained datasets, have been put forward to calculate Heat Transfer Coefficients (HTCs) in forced convective in various fluids, including water at SCPs. The Mokry et al. correlation (2011) has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in Normal and Improved Heat-Transfer (NHT and IHT) regimes. However, it is known that a Deteriorated Heat-Transfer (DHT) regime appears in bare tubes earlier than that in bundle flow geometries. Therefore, it is important to know if bare-tube heat-transfer correlations for SCW can predict HTCs at heat fluxes beyond those defined as starting of DHT regime in bare tubes. The Mokry et al. (2011) correlation fits the best SCW experimental data for HTCs and inner wall temperature for bare tubes at SCPs within the NHT and IHT regimes. However, this correlation might have problems with convergence of iterations at heat fluxes above 1000 kW/m2.

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

2018 ◽  
Author(s):  
Jun Huang ◽  
Junli Gou ◽  
Haifu Ma ◽  
Jie Fan ◽  
Jianqiang Shan
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Fluid Model ◽  
Computer Code ◽  
Medium Size ◽  
Hydraulic Characteristics ◽  
Pressurized Water

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

APPLICATION OF INFORMATION TECHNOLOGY FOR MODELING THE CONTROL DYNAMICS OF A NUCLEAR REACTOR ZONING ON THE VERTICAL AXIS

2021 ◽  
Vol 5 ◽  
pp. 45-56
Author(s):  
Valery Severyn ◽  
◽  
Elena Nikulina ◽  
Keyword(s):  
Information Technology ◽  
Nonlinear Systems ◽  
Differential Equations ◽  
Control Systems ◽  
Nuclear Reactor ◽  
Initial Conditions ◽  
Control Object ◽  
Vertical Axis ◽  
Integration Methods ◽  
Systems Of Differential Equations

The structure of information technology for modeling control systems, which includes a block of systems models, a module of integration methods and other program elements, is considered. To analyze the dynamics of control of a nuclear reactor, programs of mathematical models of a WWER-1000 nuclear reactor of the V-320 series and its control systems in the form of nonlinear systems of differential equations in the Cauchy form have been developed. For the integration of nonlinear systems of differential equations, an algorithm of the system method of the first degree is presented. A mathematical model of a WWER-1000 reactor as a control object with division into zones along the vertical axis in relative variables of state is considered, the values of the constant parameters of the model and the initial conditions corresponding to the nominal mode are given. Using information technology for ten zones of the reactor, the system integration method was used to simulate the dynamics of control of a nuclear reactor. Graphs of neutron and thermal processes in the reactor core, as well as changes in the axial offset when the reactor load is dumped under the influence of the movement of absorbing rods and an increase in the concentration of boric acid, are plotted. The analysis of dynamic processes of reactor control is carried out. The programs of integration methods and models of the WWER-1000 reactor of the V-320 series are included in the information technology to optimize the maneuvering modes of the reactor.

Dynamic Characteristics Analysis of Steam Generator Tube of Nuclear Power Plant by Finite Elements

2014 ◽  
Vol 577 ◽  
pp. 149-153
Author(s):  
Shuang Jiang ◽  
Jun Cai ◽  
Jing Wei Zhang ◽  
Qiao Zhi Sun ◽  
Xin Guo ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Modal Analysis ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Primary Circuit ◽  
Harmonic Load ◽  
Heat Transfer Tube ◽  
Finite Element Software ◽  
Transfer Tube

In nuclear power plants, the steam generator heat transfer tube is the weakest part of the primary circuit pressure boundary. Flow induced vibration is one of the main reasons for the failure of the heat transfer tube. In this paper, an ANSYS finite element software is used to carry out the modal analysis of the heat transfer tube, and to simulate the dynamic response of the heat transfer tube in the harmonic load based on the modal analysis.

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