Development and Application of a UTSG Thermal-Hydraulic Analysis Code

2014 ◽  
Author(s):  
Tenglong Cong ◽  
Guanghui Su ◽  
Wenxi Tian ◽  
Suizheng Qiu
Keyword(s):  
Steam Generator ◽  
Structural Integrity ◽  
Tube Bundle ◽  
Flow Boiling ◽  
Three Dimensional ◽  
Hydraulic Parameters ◽  
Two Phase ◽  
Drift Flux ◽  
Thermal Hydraulic Analysis ◽  
Secondary Side

Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics analysis code, named STAF, is developed based on FLUENT. With STAF code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This code is developed based on the porous media theory. In this code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF code. The calculated results are compared with a general steam generator thermohydraulic analysis code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.

CFD Analysis of Subcooled Wall Boiling at Shell Side of Steam Generator With TSP

2014 ◽  
Author(s):  
Chenglong Wang ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Wenxi Tian ◽  
Guanghui Su
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Steam Generator ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Two Phase ◽  
Pressurized Water ◽  
Support Plate ◽  
Coupled Heat Transfer ◽  
Secondary Side

This paper addresses the numerical simulation of two-phase flow heat transfer among the tube bundles with tube support plate (TSP) of an integral type pressurized water reactor steam generator using RPI wall boiling model. The subcooled nucleate boiling phenomenon and the coupled heat transfer between the SG primary side and secondary side were obtained. Also, the effects of tube support plate (TSP) and the different inlet subcooling on the thermal-hydraulic characteristics of SG were studied. From the results of the present numerical simulation, it reasonably revealed the subcooled flow boiling occurred in the SG secondary side and the distributions of key parameters around TSP, elucidating that this model can provide useful information to the design of the steam generator.

Numerical Simulation of Subcooled Flow Boiling Heat Transfer in Helical Tubes

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Jong Chull Jo ◽  
Woong Sik Kim ◽  
Chang-Yong Choi ◽  
Yong Kab Lee
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Steam Generator ◽  
Two Phase Flow ◽  
Flow Boiling ◽  
Heat Transfer Analysis ◽  
Normal Operation ◽  
Phase Flow ◽  
Two Phase ◽  
Side Flow

This paper addresses the numerical simulation of two-phase flow heat transfer in the helically coiled tubes of an integral type pressurized water reactor steam generator under normal operation using a computational fluid dynamics code. The shell-side flow field where a single-phase fluid flows in the downward direction is also calculated in conjunction with the tube-side two-phase flow characteristics. For the calculation of tube-side two-phase flow, the inhomogeneous two-fluid model is used. Both the Rensselaer Polytechnic Institute wall boiling model and the bulk boiling model are implemented for the numerical simulations of boiling-induced two-phase flow in a vertical straight pipe and channel, and the computed results are compared with the available measured data. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. Both the internal and external turbulent flows are simulated using the standard k-ε model. From the results of the present numerical simulation, it is shown that the bulk boiling model can be applied to the simulation of two-phase flow in the helically coiled steam generator tubes. In addition, the present simulation method is considered to be physically plausible in the light of discussions on the computed results.

Vibration Excitation Forces in a Rotated Triangular Tube Bundle Subjected to Two-Phase Cross Flow

2010 ◽  
Author(s):  
H. Senez ◽  
N. W. Mureithi ◽  
M. J. Pettigrew
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Cross Flow ◽  
Fretting Wear ◽  
Experimental Program ◽  
Phase Flow ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Fluid Forces ◽  
Vibration Excitation

Two-phase cross flow exists in many shell-and-tube heat exchangers. Flow-induced vibration excitation forces can cause tube motion that will result in long-term fretting wear or fatigue. Detailed flow and vibration excitation force measurements in tube bundles subjected to two-phase cross flow are required to understand the underlying vibration excitation mechanisms. Studies on this subject have already been done, providing results on flow regimes, fluidelastic instabilities, and turbulence-induced vibration. The spectrum of turbulence-induced forces has usually been expected to be similar to that in single-phase flow. However, a recent study, using tubes with a diameter larger than that in a real steam generator, showed the existence of significant quasi-periodic forces in two-phase flow. An experimental program was undertaken with a rotated-triangular array of cylinders subjected to air-water cross-flow, to simulate two-phase mixtures. The tube bundle here has the same geometry as that of a real steam generator. The quasi-periodic forces have now also been observed in this tube bundle. The present work aims to understand turbulence-induced forces acting on the tube bundle, providing results on drag and lift force spectra and their behaviour according to flow parameters, and describing their correlations. Detailed experimental test results are presented in this paper. Comparison is also made with previous measurements with larger diameter tubes. The present results suggest that quasi-periodic fluid forces are not uncommon in tube arrays subjected to two-phase cross-flow.

Flow-Induced Vibration Responses of U-Tube Bundle in Air-Water Flow

2007 ◽  
Author(s):  
In-Cheol Chu ◽  
Heung June Chung ◽  
Chang Hee Lee ◽  
Hyung Hyun Byun ◽  
Moo Yong Kim
Keyword(s):  
Steam Generator ◽  
Two Phase Flow ◽  
Tube Bundle ◽  
Damping Ratio ◽  
Geometrical Parameters ◽  
Phase Flow ◽  
Elastic Instability ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Vibration Responses

In the present study, a series of experiments have been performed to investigate a fluid-elastic instability of a nuclear steam generator U-tube bundle in an air-water two-phase flow condition. A total of 39 U-tubes are arranged in a rotated square array with a pitch-to-diameter ratio of 1.633. The diameter and other geometrical parameters of U-bend region are the same to those of an actual steam generator, but the vertical length of U-tubes are reduced to 2-span in contrast to 9-span of an actual steam generator. The following parameters were experimentally measured to evaluate a fluid-elastic instability of U-tube bundles in a two-phase flow: a general tube vibration response, a critical gap velocity, a damping ratio and a hydrodynamic mass. Based on the experimental measurements, the instability factor, K, of Connors’ relation was preliminary assessed with some assumptions on the velocity and density profiles of the two-phase flow.

Numerical Simulation of Subcooled Flow Boiling Heat Transfer in Helical Tubes

2006 ◽  
Author(s):  
Jong Chull Jo ◽  
Woong Sik Kim ◽  
Chang-Yong Choi ◽  
Yong Kab Lee
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Steam Generator ◽  
Two Phase Flow ◽  
Flow Boiling ◽  
Heat Transfer Analysis ◽  
Normal Operation ◽  
Phase Flow ◽  
Two Phase ◽  
Pressurized Water

This paper addresses the numerical simulation of two phase flow heat transfer in the helically coiled tubes of an integral type pressurized water reactor steam generator under normal operation using a CFD code. The single phase flow which flow downward direction in the shell side is also calculated together. For the calculation of tube side two-phase flow the inhomogeneous two-fluid model is used. Both the RPI (Rensselaer Polytechnic Institute) wall boiling model and the bulk boiling model are implemented for the numerical simulation and the computed results are compared with the available measured data. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. Both the internal and external turbulent flows are simulated using the standard k-ε model From the results of present numerical simulation, it is shown that the bulk boiling model can be applied to the simulation of two-phase flow in the helically coiled steam generator tubes. The results also show that the present simulation method is considered to be physically plausible when the computed results are compared with available previous experimental and numerical studies.

Three-dimensional study on steady thermohydraulics characteristics in secondary side of steam generator

2014 ◽  
Vol 70 ◽  
pp. 188-198 ◽  
Author(s):  
Tenglong Cong ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu ◽  
Yongcheng Xie ◽  
...  
Keyword(s):  
Steam Generator ◽  
Three Dimensional ◽  
Secondary Side

Investigations of In-Plane Fluidelastic Instability in a Multi-Span U-Bend Tube Bundle: Tests in Air Flow

2017 ◽  
Author(s):  
Paul Feenstra ◽  
Teguewinde Sawadogo ◽  
Bruce Smith ◽  
Victor Janzen ◽  
Helen Cothron
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Air Flow ◽  
Cross Flow ◽  
Test Rig ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Fluidelastic Instability ◽  
Air Blower ◽  
R 134A

The tubes in the U-bend region of a recirculating type of nuclear steam generator are subjected to cross-flow of a two-phase mixture of steam and water. There is a concern that these tubes may experience flow-induced vibration, including the damaging effects of fluidelastic instability. This paper presents an update and results from a series of flow-induced vibration experiments performed by Canadian Nuclear Laboratories for the Electric Power Research Institute (EPRI) using the Multi-Span U-Bend test rig. In the present experiments, the main focus was to investigate fluidelastic instability of the U-tubes subjected to a cross-flow of air. The tube bundle is made of 22 U-tubes of 0.5 in (12.7 mm) diameter, arranged in a rotated triangular configuration with a pitch-over-diameter ratio of 1.5. The test rig could be equipped with variable clearance flat bar supports at two different locations to investigate a variety of tube and support configurations. The primary purpose of the overall project is to study the effect of flat bar supports on ‘in plane’ (‘streamwise’) instability in a U-tube bundle with realistic tube-to-support clearances or preloads, and eventually in two-phase flow conditions. Initially, the test rig was designed for tests in air-flow using an industrial air blower. Tests with two-phase Freon refrigerant (R-134a) will follow. This paper describes the test rig, experimental setup, and the challenges presented by simulating an accurate representation of current steam generator designs. Results from the first series of tests in air flow are described.

Experimental Investigation on Natural Circulation in a Thermosyphon Reboiler

2010 ◽  
Author(s):  
Xing Luo ◽  
Yanming Gao ◽  
Stephan Kabelac
Keyword(s):  
Heat Transfer ◽  
Tube Bundle ◽  
Flow Boiling ◽  
Natural Circulation ◽  
Liquid Product ◽  
Pilot Scale ◽  
Vertical Tube ◽  
Transfer Coefficients ◽  
Two Phase ◽  
Set Up

Thermosyphon reboilers are widely used in refineries, petrochemical industries and other chemical processes. The liquid product stream coming from the bottom of the vapor-liquid separator is heated in an evaporator consisting of a vertical tube or tube bundle. When the evaporation occurs, the specific volume of the two-phase fluid increases. The upward buoyancy force caused by the density difference between the evaporator and down-flow pipe drives the fluid flowing through the evaporator in to the separator and forms a natural circulation. The experiments were conducted in a pilot scale thermosyphon system in which the evaporator consists of 7 steel tubes (outside diameter 30 mm, wall thickness 2 mm, length 4 m). A mathematical model was set up to simulate the heat transfer and pressure drop, in which empirical equations from literature were used. With the help of the simulation, the flow boiling heat transfer coefficients inside the tubes can be evaluated from a few measured local wall temperatures.

scholarly journals ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

2016 ◽  
Vol 2016 ◽  
pp. 1-15
Author(s):  
Takeshi Takeda ◽  
Akira Ohnuki ◽  
Daisuke Kanamori ◽  
Iwao Ohtsu
Keyword(s):  
Steam Generator ◽  
Safety System ◽  
Injection System ◽  
Test Facility ◽  
Gas Release ◽  
Dissolved Gas ◽  
Nitrogen Gas ◽  
Two Phase ◽  
Pressurized Water ◽  
Secondary Side

Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

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