Investigation of ATHLET System Code for Supercritical Water Applications

2014 ◽  
Author(s):  
J. Samuel ◽  
G. Lerchl ◽  
G. D. Harvel ◽  
I. Pioro
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
High Pressures ◽  
Computer Code ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Water Reactors ◽  
First Time

SuperCritical Water-cooled Reactors (SCWRs) are one of six Generation-IV nuclear-reactor concepts. They are expected to have high thermal efficiencies within the range of 45–50% owing to the reactor’s high pressures and outlet temperatures. Efforts have been made to study the supercritical phenomena both analytically and experimentally. However, codes that have been used to study the phenomena analytically have not been validated for supercritical water. The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid- and transport-properties packages, thus enabling a transition from subcritical to supercritical fluid states. This extension needs to be validated using experimental data. In this work, the applicability of ATHLET code to predict supercritical-water behaviour in various heat-transfer conditions is assessed. Several well-known heat-transfer correlations for supercritical fluids are added to the code and applied for the first time in ATHLET simulations of experiments. A numerical model in ATHLET is created to represent an experimental test section and results for the heat transfer coefficient, bulk fluid temperature, and the tube inside-wall temperature are compared with the experimental data. The results from the ATHLET simulations are promising in the Normal and Enhanced Heat-Transfer Regimes. However, important phenomena such as Deteriorated Heat Transfer are currently not accurately predicted. While ATHLET can be used to develop preliminary design solutions for SCWRs, a significant effort in analysis of experimental work is required to make further advancements in the use of ATHLET for SCW applications.

Experimental Study of Heat Transfer in a 7-Element Bundle Cooled With Supercritical Freon-12

2012 ◽  
Author(s):  
G. Richards ◽  
J. Samuel ◽  
A. S. Shelegov ◽  
P. L. Kirillov ◽  
I. L. Pioro ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Mass Flux ◽  
Test Section ◽  
High Pressures ◽  
Natural Circulation ◽  
Heat Fluxes ◽  
Bulk Fluid ◽  
Test Loop

Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are technical difficulties in testing and experimental costs at high pressures, temperatures and heat fluxes. Also, there are only a few SCW experimental setups currently in the world capable of providing data. SuperCritical Water-cooled nuclear Reactors (SCWRs), as one of the six concepts of Generation IV reactors, cannot be designed without such data. Therefore, a preliminary approach uses modeling fluids such as carbon dioxide and refrigerants instead of water are practical. In particularly, experiments in supercritical refrigerant-cooled bundles can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) with the critical pressure of 4.136 MPa and the critical temperature of 111.97°C. A set of experimental data obtained at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) in a vertically-oriented bundle cooled with supercritical R-12 was analyzed. This dataset consisted of 20 runs. The test section was 7-element bundle installed in a hexagonal flow channel with 3 grid spacers. Data was collected at pressures of approximately 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values of mass flux were ranged from 400 to 1320 kg/m2s and inlet temperatures ranged from 72 to 120°C. The test section consisted of fuel-element simulators that were 9.5 mm in OD with the total heated length of about 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 thermocouples. Analysis of the data has confirmed that there are three distinct heat-transfer regimes for forced convention in supercritical fluids: 1) Normal heat transfer; 2) Deteriorated heat transfer characterized with higher than expected wall temperatures; and 3) Enhanced heat transfer characterized with lower than expected wall temperatures. It was also confirmed that the effects of spacers are evident which was previously observed in sub-critical experimental data. Further analysis needs to be conducted on the deteriorated heat transfer phenomena for low mass flux cases which represent accident scenarios. This can be done by designing a natural circulation experimental test loop.

Investigation of Characteristic-Temperature Approaches for Heat-Transfer Correlations at Supercritical Conditions

2013 ◽  
Author(s):  
Sarah Mokry ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Wall Temperature ◽  
Characteristic Temperature ◽  
Fluid Temperature ◽  
Transfer Coefficients ◽  
Supercritical Conditions ◽  
Bulk Fluid ◽  
Heat Transfer Coefficients ◽  
Bare Tube

It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) and high coolant temperatures (350–625°C). In support of the development of SuperCritical Water-cooled Reactors (SCWRs), research is currently being conducted for heat-transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare-tube data can be used as a conservative approach. A number of empirical generalized correlations, based on experimentally obtained datasets, have been proposed to calculate Heat Transfer Coefficients (HTCs) in forced convective heat transfer for various fluids, including water, at supercritical pressures. These bare-tube-based correlations are available in various literature sources. There have been a number of methods applied to correlate heat transfer data. The most conventional approach, which accounts for property variations in the data, is to modify the classical Dittus-Boelter equation for forced convection. However, analysis and comparison of these correlations has shown that differences in HTC values can be up to several hundred percent. In general, the familiar correlations of Dittus-Boelter and Bishop et al. have used the bulk-fluid temperature approach for characteristic temperature properties evaluations. However, at high heat fluxes, fluid near the tube-wall will have a temperature close to that of the wall temperature. This might be significantly different from the bulk-fluid temperature. Therefore, another approach can be used based on the wall temperature as the characteristic temperature. The Swenson et al. correlation is based upon this approach. Finally, a third approach has been considered in which the film-temperature is used as the characteristic temperature (Tf = (Tw+Tb) / 2). McAdams et al. based their correlation for annuli on this approach. Therefore, the objective of this paper is to evaluate the three characteristic temperature approaches, (1) Bulk-fluid temperature approach; (2) Wall-temperature approach; and (3) Film-temperature approach, and determine which characteristic temperature method can most accurately predict supercritical water heat transfer coefficients. Both classical correlations and more recently developed correlations are considered in this investigation.

Development of Supercritical Water Heat-Transfer Correlation for Vertically-Upward Internally-Ribbed Tubes

2014 ◽  
Author(s):  
Weiqiang Zhang ◽  
Huixiong Li ◽  
Qing Zhang ◽  
Yifang Zhang ◽  
Tai Wang
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Large Deviation ◽  
Experimental Studies ◽  
Great Influence ◽  
Dimensionless Numbers ◽  
Experimental Database ◽  
Heat Transfer Correlations

The investigation on the heat transfer characteristics for supercritical pressure water (SCW) is of value for the development of the supercritical water-cooled nuclear reactor (SCWR). As an important heat transfer enhancement element, heat transfer for SCW in internally-ribbed tubes was still not solved, though lots of experimental studies have been published and a great many heat transfer correlations were proposed. This paper presented an analysis of heat transfer in the internally-ribbed tubes, through comparing heat transfer correlations for SCW gained from different internally-ribbed tubes under the same operating condition. It was found that all existing heat transfer correlations reported could not been well applied for various internally-ribbed tubes with large deviation between prediction results and experimental values, because rib geometry had a great influence on heat transfer of internally-ribbed tubes. On the basis of experimental data collected from open literature for internally-ribbed tubes, a new general calculation correlation of heat transfer coefficient for SCW was developed for various internally-ribbed tubes by combining an optimized empirical correlation for vertically-upward smooth tubes and four dimensionless numbers of rib geometry. The results show that the calculated values of the new present correlation is in reasonable agreement with available experimental data collected. Moreover, the new correlation was verified well by experiment data of two new-type internally-ribbed tubes performed beyond the above experimental database.

Experimental Heat Transfer in an Annular Channel and 3-Rod Bundle Cooled With Upward Flow of Supercritical Water

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
V. G. Razumovskiy ◽  
Eu. N. Pis’mennyi ◽  
Kh. Sidawi ◽  
I. L. Pioro ◽  
A. Eu. Koloskov
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Annular Channel ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Experimental Heat ◽  
Rod Bundle ◽  
Fuel Bundle ◽  
Experimental Heat Transfer ◽  
Heated Length

There have been relatively few publications detailing heat transfer to supercritical water (SCW) flowing through a channel with a bundle or just with a single rod (annular channel) as compared to heat transfer to SCW in bare tubes. In the present paper, results of experimental heat transfer to SCW flowing upward in an annular channel with a heated rod equipped with four helical ribs and a 3-rod bundle (rods are also equipped with four helical ribs) are discussed. The experimental results include bulk-fluid-temperature, wall-temperature, and heat-transfer-coefficient (HTC) profiles along the heated length (485 mm) for these flow geometries. Data obtained from this study could be applicable as a reference estimation of heat transfer for future fuel-bundle designs.

Simulation of the Natural Convection Test of Phenix Reactor Using Modified ATHLET

2012 ◽  
Author(s):  
Chong Zhou ◽  
Klaus Huber ◽  
Xu Cheng
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Natural Convection ◽  
Fast Reactor ◽  
Atomic Energy Commission ◽  
Computer Code ◽  
Light Water Reactors ◽  
Light Water ◽  
Water Reactors ◽  
Alternative Energies

The sodium-cooled fast reactor (SFR) is gathering worldwide attention for its features of the fast-spectrum reactor and closed fuel recycle system. This paper presents the modification of the ATHLET code for application to SFRs. The thermal-hydraulic computer code ATHLET is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) in Germany for light water reactors analysis. In this paper, a sodium properties package is implemented in to the ATHLET code, and heat transfer correlations for sodium are also added for heat transfer prediction. To evaluate the capability of the modified code, the Phenix reactor, a SFR operated by French Alternative Energies and Atomic Energy Commission (CEA) from 1973 to 2009, is modeled. The scenario of transient from forced to natural convection is simulated and analyzed. The results are compared with the experimental data of the natural convection ultimate test of the Phenix facility. Results achieved so far indicate good applicability of the modified ATHLET code.

Assessment of FLUENT Code as a Tool for SCW Heat Transfer Analysis

2013 ◽  
Author(s):  
Amjad Farah ◽  
Glenn Harvel ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Fluid Dynamics ◽  
Supercritical Water ◽  
High Pressures ◽  
Stokes Equations ◽  
Heat Transfer Analysis ◽  
Model Parameters ◽  
Transfer Coefficients ◽  
Bulk Fluid ◽  
Heat Transfer Coefficients

Generation-IV SuperCritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45–50% owing to the reactor’s high pressures and outlet temperatures. The behavior of supercritical water however, is not well understood and most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations which do not capture the multi-dimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. Computational Fluid Dynamics (CFD) is a numerical approach to model fluids in multidimensional space using the Navier-Stokes equations and databases of fluid properties to arrive at a full simulation of a fluid dynamics and heat transfer system. In this work, the CFD code, FLUENT-12, is used with associated software such as Gambit and NIST REFPROP to predict the Heat Transfer Coefficients at the wall and corresponding wall temperature profiles inside vertical bare tubes with SuperCritical Water (SCW) as the cooling medium. The numerical results are compared with experimental data and 1-D models represented by existing empirical correlations. Analysis of the individual heat-transfer regimes is conducted using an axisymmetric 2-D model of tubes of various lengths and composed of different nodalizations along the heated length. Wall temperatures and heat transfer coefficients were analyzed to select the best model for each region (below, at and above the pseudocritical region). Two turbulent models were used in the process: k-ε and k-ω, with variations in the sub-model parameters such as viscous heating, thermal effects, and low-Reynolds number correction. Results of the analysis show a fit of ±10% for the wall temperatures using the SST k-ω model in the deteriorated heat transfer regime and less than ±5% for the normal heat transfer regime. The accuracy of the model is higher than any empirical correlation tested in the mentioned regimes, and provides additional information about the multidimensional effects between the bulk-fluid and wall temperatures.

Preliminary Investigation of Heat-Transfer Correlation for Upward Flow of CO2 at Supercritical Pressure

2013 ◽  
Author(s):  
Eugene Saltanov ◽  
Igor Pioro ◽  
Glenn Harvel
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Preliminary Investigation ◽  
Convection Heat Transfer ◽  
Theoretical Models ◽  
Supercritical Pressure ◽  
Operational Parameters ◽  
Bulk Fluid ◽  
Wide Range

The development of Generation IV nuclear reactors is an ongoing worldwide interdisciplinary effort. Canada is involved with the development of the SuperCritical Water-cooled Reactor (SCWR). One of the numerous engineering challenges associated with this development is to ensure intensive and stable heat transfer to SuperCritical Water (SCW) in a core. It is very important to develop sophisticated theoretical models to improving understanding of physical processes behind forced-convective heat transfer and its deterioration in near-critical region. However, not all turbulent models implemented in Computational Fluid Dynamics (CFD) codes are applicable to heat transfer at supercritical pressures. Additionally these codes should be first tuned on the basis of experimental data and, after that, used in similar calculations. Therefore, there is still a great reliance on 1D heat-transfer correlations for preliminary calculations. Performing experiments in SCW is extremely expensive due to the high temperatures and pressures involved. Therefore, it is reasonable to study the general properties of SuperCritical Fluids (SCF) by running experiments with modelling fluids; for example, carbon dioxide, which is widely used as a modeling fluid. In view of this, there is a need to compile a large database that will include experimental data on forced-convection heat-transfer to SuperCritical (SC) CO2 within a wide range of operational parameters. To date, a part of this database has been produced. It contains the bulk-fluid and wall temperatures of CO2 flowing upward in a vertical bare tube. CO2 is at supercritical pressure and the bulk-fluid and wall temperatures are below or above the pseudocritical temperature at the inlet of the test-section. Therefore, the objectives of this paper are: 1) to propose a correlation in a standard non-dimensional form, which will generalize data within ±25%; 2) to compare this correlation with the most recent, as well as previous 1D correlations for SC CO2 and SCW (appropriate scaling is performed).

Study on Deteriorated Heat Transfer in Upward Flow of Supercritical Water in a 1-M Vertical Bare Tube

2021 ◽  
Author(s):  
N. Dort-Goltz ◽  
I. Pioro ◽  
J. McKellar
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Fluid Temperature ◽  
Upward Flow ◽  
Transfer Coefficients ◽  
Safe Operation ◽  
Bulk Fluid ◽  
Heat Transfer Coefficients ◽  
Heated Length

Abstract SuperCritical Water-cooled Reactors (SCWRs) represent potential improvements over traditional water-cooled reactors in many respects, including thermal efficiency. These reactors are still under development, however, thermalhydraulics data needed for their design are lacking. Experimentation is complex and costly. In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller hydraulic-equivalent diameters. As a first step, tests have been conducted by others on heat transfer in short, vertical bare tubes cooled with the upward flow of SCW. The objective of this work is to analyze that collected data with particular attention to the Deteriorated Heat Transfer (DHT) regime. The DHT regime is characterized by reduced Heat Transfer Coefficients (HTCs) and consequently increased wall temperatures. As such, it represents a hazard to the safe operation of a Nuclear Power Plant (NPP). The results of this analysis indicate that DHT did occur in each of the tests analyzed, often seen as a gradual decrease in HTC along the heated length, but occasionally as a sharp “dip”. The DHT can occur along the heated length, when the bulk-fluid temperature is close to or within the pseudocritical region. The results also confirmed that the Dittus-Boelter correlation does not adequately predict HTCs within the pseudocritical region. Two other applied correlations (Gupta et al. and Mokry et al.) performed better, but neither was able to predict the occurrence of the DHT. The results of this analysis will be of use to designers and developers of SCWRs, and can help to plan future experiments.

Improvement of Supercritical Water Heat-Transfer Correlations for Vertical Bare Tubes

2014 ◽  
Author(s):  
Sarah Mokry ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Operating Conditions ◽  
Entrance Region ◽  
Fluid Temperature ◽  
Supercritical Conditions ◽  
Bulk Fluid ◽  
Wide Range ◽  
Heat Transfer Correlations ◽  
Region Effect

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. These high temperature, high pressure reactors will have much higher operating parameters compared to current Nuclear Power Plants (NPPs) (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. In support of developing these SCWRs, studies are currently being conducted for heat transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer at supercritical conditions from power-reactor fuel bundles to a coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare-tube data can be used as a conservative approach. A number of empirical generalized correlations, based on experimentally obtained datasets, have been proposed to calculate Heat Transfer Coefficients (HTCs) in forced convective heat transfer for various fluids, including water at supercritical pressures. There have been a number of methods applied to correlate heat transfer data. The most conventional approach is to modify the classical Dittus-Boelter correlation for forced convection. The Bishop et al. correlation is an example of this type modification with an addition of an entrance-region term. The Mokry et al. correlation (2009) was developed as a Dittus-Boelter-type correlation with thermophysical properties taken at a bulk-fluid temperature. The derived correlation has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in normal and improved heat-transfer regimes. This correlation has an uncertainty of about ±25% for HTC values and about ±15% for calculated wall temperature. However, this correlation does not take into account the entrance-region effect. The objective of this paper is an investigation of the entrance-region effect to be incorporated into the proposed Mokry et al. correlation (2009) in an attempt to further improve its accuracy.

Oscillatory Heat Transfer in a Pipe Subjected to a Laminar Reciprocating Flow

1996 ◽  
Vol 118 (3) ◽  
pp. 592-597 ◽  
Author(s):  
T. S. Zhao ◽  
P. Cheng
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Heat Flux ◽  
Nusselt Number ◽  
Numerical Study ◽  
Temperature Cycle ◽  
Uniform Heat Flux ◽  
Fluid Temperature ◽  
Cycle Space ◽  
Reciprocating Flow

An experimental and numerical study has been carried out for laminar forced convection in a long pipe heated by uniform heat flux and subjected to a reciprocating flow of air. Transient fluid temperature variations in the two mixing chambers connected to both ends of the heated section were measured. These measurements were used as the thermal boundary conditions for the numerical simulation of the hydrodynamically and thermally developing reciprocating flow in the heated pipe. The coupled governing equations for time-dependent convective heat transfer in the fluid flow and conduction in the wall of the heated tube were solved numerically. The numerical results for time-resolved centerline fuid temperature, cycle-averaged wall temperature, and the space-cycle averaged Nusselt number are shown to be in good agreement with the experimental data. Based on the experimental data, a correlation equation is obtained for the cycle-space averaged Nusselt number in terms of appropriate dimensionless parameters for a laminar reciprocating flow of air in a long pipe with constant heat flux.

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