Numerical Simulation and Analyses of the Loss of Feedwater Transient at the Unit 4 of Kola NPP

2002 ◽  
Author(s):  
Vladimir D. Stevanovic ◽  
Zoran V. Stosic ◽  
Michael Kiera ◽  
Uwe Stoll
Keyword(s):  
Numerical Simulation ◽  
Steam Generator ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Numerical Results ◽  
Coolant Temperature ◽  
Feed Water ◽  
Steam Generation ◽  
Horizontal Steam Generator ◽  
Mass Inventory

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give complete insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with safety codes.

An adaptive p-FEM for three-dimensional concrete aggregate models

2020 ◽  
Vol 11 (01) ◽  
pp. 2050004
Author(s):  
Ruiqi Guo ◽  
Yingxiong Xiao
Keyword(s):  
Numerical Simulation ◽  
Finite Element ◽  
Three Dimensional ◽  
Poor Quality ◽  
Numerical Results ◽  
Concrete Aggregate ◽  
Material Interfaces ◽  
Reinforced Composite ◽  
Text Version ◽  
Particle Reinforced Composite

Numerical simulation for concrete aggregate models (CAMs) with different shape aggregates usually requires high accuracy and convergence near the material interfaces. But high memory usage will be needed for those traditional finite element methods such as the method by using mesh refinement throughout the domain. Thus, an adaptive [Formula: see text]-version finite element method ([Formula: see text]-FEM) is proposed in this paper for the solution of 3D CAM problems, and meanwhile the resulting adaptive computational algorithm and post-processing program are presented. We firstly focused two typical 3D weak discontinuity problems on the influence of different convergence criterions for the computational results of each point on the interface in order to verify the efficiency and convergence of the resulting [Formula: see text]-FEM, and then this method is successfully applied to the numerical simulation of CAMs with different shape aggregates. In addition, an efficient hybrid realization method which combines ANSYS and Hypermesh software is also presented in order to quickly establish the geometric models of 3D CAMs. The numerical results have been shown that the proposed [Formula: see text]-FEM can efficiently solve the concrete-like particle-reinforced composite problems and more accurate numerical results can be obtained under the case of fewer elements used in simulation of CAMs, even there being some elements with poor quality.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

scholarly journals Sensitivity Evaluation of AP1000 Nuclear Power Plant Best Estimation Model

2017 ◽  
Vol 2017 ◽  
pp. 1-13 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Temperature Variation ◽  
Steam Generator ◽  
Nuclear Power ◽  
Channel Number ◽  
Design Parameters ◽  
Coolant Temperature ◽  
Node Number

The best estimation process of AP1000 Nuclear Power Plant (NPP) requires proper selections of parameters and models so as to obtain the most accurate results compared with the actual design parameters. Therefore, it is necessary to identify and evaluate the influences of these parameters and modeling approaches quantitatively and qualitatively. Based on the best estimate thermal-hydraulic system code RELAP5/MOD3.2, sensitivity analysis has been performed on core partition methods, parameters, and model selections in AP1000 Nuclear Power Plant, like the core channel number, pressurizer node number, feedwater temperature, and so forth. The results show that core channel number, core channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop through the reactor. The feedwater temperature is a sensitive factor to the Steam Generator (SG) outlet temperature and the Steam Generator outlet pressure. In addition, the cross-flow model nearly has no effects on the coolant temperature variation and pressure drop in the reactor, in both the steady state and the loss of power transient. Furthermore, some fittest parameters with which the most accurate results could be obtained have been put forward for the nuclear system simulation.

CFD Analysis of the Primary Side in a Helical-Coil Once-Through Steam Generator

2017 ◽  
Author(s):  
Kai Ye ◽  
Yaoli Zhang ◽  
Jianshu Lin ◽  
Ning Li ◽  
Yinglin Yang ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Rational Design ◽  
Large Scale ◽  
Three Dimensional ◽  
Helical Coil ◽  
Complicated Structure ◽  
Ansys Fluent ◽  
Flow Parameters ◽  
Computational Fluid Dynamics Cfd

The helical-coil once-through steam generator (OTSG) is usually used in the nuclear power plant when the compactness of equipment was taken into consideration. The investigation of flow parameters in the primary side is valuable for the optimization of the OTSG. The purpose of this research is to obtain a further understanding of fluid behaviors in the primary side of the OTSG to achieve a more rational design. Using ANSYS ICEM and ANSYS FLUENT, a three-dimensional (3D) computational fluid dynamics (CFD) model was created and analyzed. Through a series of cases, the velocity profiles and pressure drop through the primary side of the helical-coil OTSG have been calculated, and the influences of different structure designs on the coolant flow parameters have also been tested. Ultimately some pertinent suggestions for improvements were proposed, and insight is obtained into the importance of various modeling considerations in such a model with a complicated structure and large-scale grids.

Study on the Principle of Yarn Splicing of Pneumatic Splicer Using Numerical Simulation

2012 ◽  
Vol 588-589 ◽  
pp. 1355-1358
Author(s):  
Xiao Xing ◽  
Guo Ming Ye
Keyword(s):  
Numerical Simulation ◽  
Flow Field ◽  
Three Dimensional ◽  
Numerical Results ◽  
Simulation Results

During the splicing process of pneumatic splicer, the principle of yarn splicing is closely related to the flow field inside the splicing chamber. This paper presents a numerical simulation of the flow char-acteristics inside the splicing chamber of the pneumatic splicer. A three-dimensional grid and the realizable tur¬bulence model are used in this simulation. The numerical results of veloc¬ity vectors distribution inside the chamber are shown. Streamlines starting from the two air injectors are also acquired. Based on the simulation, the principle of yarn splicing of the pneumatic splicer is discussed. The airflow in the splicing chamber can be divided into three regions. In addition, the simulation results have well sup¬ported the principle of yarn splicing of pneumatic splicer claimed by the splicing chamber makers.

scholarly journals Particle-laden flow around an obstacle in a square pipe: experiments and modeling

2020 ◽  
Vol 21 (5) ◽  
pp. 517
Author(s):  
Ouardia Ait Oucheggou ◽  
Véronique Pointeau ◽  
Guillaume Ricciardi ◽  
Élisabeth Guazzelli ◽  
Laurence Bergougnoux
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Particle Deposition ◽  
Three Dimensional ◽  
Vertical Tube ◽  
Nuclear Industry ◽  
Solid Particles ◽  
Complex Nature

Particle trapping and deposition around an obstacle occur in many natural and industrial situations and in particular in the nuclear industry. In the steam generator of a nuclear power plant, the progressive obstruction of the flow due to particle deposition reduces the efficiency and can induce tube cracking leading to breaking and damage. The steam generator then loses its role as a safety barrier of the nuclear power plant. From a fundamental standpoint, dilute and concentrated particulate flows have received a growing attention in the last decade. In this study, we investigate the transport of solid particles around obstacles in a confined flow. Experiments were performed in a simplified configuration by considering a laminar flow in a vertical tube. An obstacle was inserted at the middle height of the tube and neutrally-buoyant particles were injected at different locations along the tube. We have investigated first the trajectories of individual particles using particle tracking (PT). Then, the particle trajectories were modeled by using the Boussinesq-Basset-Oseen equation with a flow velocity field either measured using particle image velocimetry (PIV) or calculated by the Code_Saturne software in order to account for the three-dimensional (3D) character of the obstacle wake. This paper presents a comparison between the experimental observations and the predictions of the modeling for an obstacle consisting of a rectangular step at a Reynolds number of ≈100 and evidences the importance of accounting for the 3D complex nature of the flow.

A New Eddy Current 3-D Profilometry for a Quantitative Measurement of Geometric Anomaly in Steam Generator Tubes of Nuclear Power Plants

2006 ◽  
Vol 321-323 ◽  
pp. 426-429
Author(s):  
Deok Hyun Lee ◽  
Myung Sik Choi ◽  
Do Haeng Hur ◽  
Jung Ho Han ◽  
Myung Ho Song ◽  
...  
Keyword(s):  
Steam Generator ◽  
Eddy Current ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Three Dimensional ◽  
Accurate Information ◽  
Profile Measurement ◽  
Root Cause ◽  
Steam Generator Tubes

Most of the corrosive degradations in steam generator tubes of nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, an expansion transition, u-bend, ding, dent, bulge, etc. Therefore, accurate information on a geometric anomaly in a tube is a prerequisite to the activity of a non destructive inspection for a precise and earlier detection of a defect in order to prevent a failure during an operation, and also for a root cause analysis of a failure. In this paper, a newly developed eddy current technique of a three-dimensional profilometry is introduced and the proof for the applicability of the technique to a plant inspection is provided. The quantitative profile measurement using a new eddy current probe was performed on steam generator expansion mock-up tubes with various geometric anomalies typically observed in the operating power plants, and the accuracy of the measured data was compared with those from the laser profilometry.

A Heat Recovery Study: Application of Intercooler as a Feed-Water Heater of Heat Recovery Steam Generator

2010 ◽  
Author(s):  
P. Shukla ◽  
M. Izadi ◽  
P. Marzocca ◽  
D. K. Aidun
Keyword(s):  
Steam Generator ◽  
Gas Turbines ◽  
Heat Recovery ◽  
Waste Heat ◽  
Feed Water ◽  
Heat Recovery Steam Generator ◽  
Steam Generation ◽  
Water Heater ◽  
Theoretical Support ◽  
Recovery Study

This paper evaluates the possibility of combining an intercooled gas turbine power cycle with a steam turbine cycle and the application of the intercooler as a feed-water heater for the heat recovery steam generator. In advance gas turbines the intercooler is used to improve the overall efficiency of the simple cycle but a noticeable amount of heat is wasted to the atmosphere. However, this energy can be recovered by using the proposed method in the current study. Accordingly, a thermodynamic study is done to investigate the improvement in efficiency achieved by feed-water heating. First the effect of intercooler parameters on the outlet condition of the water is studied. The bottoming cycle is then studied in details for the effect of feed-water temperature. An estimate of the energy saving by using the proposed method will be reported. The results show that less heat input will be required for the same amount of steam generation. The current study provides a theoretical support for waste heat recovery from the intercooler.

VVER-440 Steam Generator’s Two-Phase Flow Analysis

2016 ◽  
Vol 821 ◽  
pp. 57-62
Author(s):  
Lukas Joch ◽  
Roman Krautschneider
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Flow Analysis ◽  
Three Dimensional ◽  
Reactor Power ◽  
Physical Models ◽  
Two Phase ◽  
Ansys Fluent

The subject of this report is creation of three-dimensional thermal hydraulic model of horizontal steam generator for Dukovany nuclear power plant. A procedure is presented for simulation and analysis of secondary side of PGV-440 steam generator for nominal and increased reactor power. A two-fluid approach is applied for modeling physical processes inside the steam generator. Physical models were implemented in ANSYS Fluent CFD environment using User Defined Functions (UDFs). Results from this thermal hydraulic numerical model can be used for various other subsequent nuclear power plant operations and safety analysis.

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