A Blind, Numerical Benchmark Study on Supercritical Water Heat Transfer Experiments in a 7-Rod Bundle

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
M. Rohde ◽  
J. W. R. Peeters ◽  
A. Pucciarelli ◽  
A. Kiss ◽  
Y. F. Rao ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Real Life ◽  
Predictive Ability ◽  
Test Facility ◽  
Test Case ◽  
Significant Deterioration ◽  
Rod Bundle ◽  
Sst Turbulence Model ◽  
Water Reactors

Heat transfer in supercritical water reactors (SCWRs) shows a complex behavior, especially when the temperatures of the water are near the pseudocritical value. For example, a significant deterioration of heat transfer may occur, resulting in unacceptably high cladding temperatures. The underlying physics and thermodynamics behind this behavior are not well understood yet. To assist the worldwide development in SCWRs, it is therefore of paramount importance to assess the limits and capabilities of currently available models, despite the fact that most of these models were not meant to describe supercritical heat transfer (SCHT). For this reason, the Gen-IV International Forum initiated the present blind, numerical benchmark, primarily aiming to show the predictive ability of currently available models when applied to a real-life application with flow conditions that resemble those of an SCWR. This paper describes the outcomes of ten independent numerical investigations and their comparison with wall temperatures measured at different positions in a 7-rod bundle with spacer grids in a supercritical water test facility at JAEA. The wall temperatures were not known beforehand to guarantee the blindness of the study. A number of models have been used, ranging from a one-dimensional (1-D) analytical approach with heat transfer correlations to a RANS simulation with the SST turbulence model on a mesh consisting of 62 million cells. None of the numerical simulations accurately predicted the wall temperature for the test case in which deterioration of heat transfer occurred. Furthermore, the predictive capabilities of the subchannel analysis were found to be comparable to those of more laborious approaches. It has been concluded that predictions of SCHT in rod bundles with the help of currently available numerical tools and models should be treated with caution.

SCWR Rod Bundle Thermal Analysis by a CFD Code

2008 ◽  
Author(s):  
M. Sharabi ◽  
W. Ambrosini ◽  
N. Forgione ◽  
S. He
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Power Distribution ◽  
Transport Model ◽  
Rod Bundles ◽  
Fuel Rod ◽  
Reynolds Stress Transport Model ◽  
Rod Bundle ◽  
Transfer Behavior ◽  
Water Reactors

The present paper describes the results of the application of the FLUENT code in the analysis of rod bundle configurations proposed for high pressure supercritical water reactors. The model considers a 1/8 slice of a rod bundle. The details from CFD calculations offer predictions of the circumferential clad surface temperature and of the effect of axial power distribution on the mass exchange between subchannels and on the maximum surface rod temperature. Geometry and boundary conditions are adopted from a previous work that made use of subchannel programs, allowing for a direct comparison between the two techniques. Both the standard k-ε model and the Reynolds stress transport model are used. Conclusions are drawn about the present capabilities in predicting heat transfer behavior in fuel rod bundles proposed for supercritical water reactors.

Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle

2009 ◽  
Author(s):  
V. G. Razumovskiy ◽  
Eu. N. Pis’mennyy ◽  
A. Eu. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Annular Channel ◽  
Heat Fluxes ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Temperature Regimes ◽  
Outside Diameter

The results of heat transfer to supercritical water flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of the tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 800 to 3000 kg/m2·s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 (heat flux rate up to 2.5 kJ/kg). Temperature regimes of the annular channel and 3-rod bundle were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles.

Nonuniform heat transfer of supercritical water in a tight rod bundle – Assessment of correlations

2017 ◽  
Vol 110 ◽  
pp. 570-583 ◽  
Author(s):  
Han Wang ◽  
Qincheng Bi ◽  
Linchuan Wang ◽  
Laurence K.H. Leung
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Rod Bundle

Validation of Trio_U: Numerical Simulations of a ROCOM Buoyancy Driven Test Case

2004 ◽  
Author(s):  
T. Ho¨hne ◽  
U. Bieder ◽  
S. Kliem ◽  
H.-M. Prasser
Keyword(s):  
Turbulent Flows ◽  
Natural Circulation ◽  
Density Difference ◽  
Test Facility ◽  
Test Case ◽  
Pressurized Water ◽  
Constant Flow Rate ◽  
Sensor Position ◽  
Nuclear Systems ◽  
Water Reactors

A generic investigation of the influence of density differences between the primary loop inventory and the ECC water on the mixing in the downcomer was made at the ROCOM Mixing Test Facility at Forschungszentrum Rossendorf (FZR)/Germany. ROCOM is designed for experimental coolant mixing studies over a wide variety of possible scenarios. It is equipped with advanced instrumentation, which delivers high-resolution information characterizing either temperature or boron concentration fields in the investigated pressurized water reactor. For the validation of the Trio_U code an experiment with 5% constant flow rate in one loop (magnitude of natural circulation) and 10% density difference between ECC and loop water was taken. Trio_U is a CFD code developed by the CEA France, aimed to supply an efficient computational tool to simulate transient thermal-hydraulic single-phase turbulent flows encountered in the nuclear systems as well as in the industrial processes. For this study a LES approach was used for mesh sizes according to between 300000–2 million control volumes. The results of the experiment as well as of the numerical calculations show, that a streak formation of the water with higher density is observed. At the upper sensor, the ECC water covers a small azimuthal sector. The density difference partly suppresses the propagation of the ECC water in circumferential direction. The ECC water falls down in an almost straight streamline and reaches the lower downcomer sensor position directly below the affected inlet nozzle. Only later, coolant containing ECC water appears at the opposite side of the downcomer. The study showed, that density effects play an important role during natural convection with ECC injection in pressurized water reactors. Furthermore it was important to point out, that Trio_U is able to cope the main flow and mixing phenomena.

NUMERICAL INVESTIGATION OF HEAT TRANSFER TO SUPERCRITICAL WATER IN A 2 × 2 ROD BUNDLE WITH TWO CHANNELS

2018 ◽  
Vol 49 (2) ◽  
pp. 103-118
Author(s):  
Ibrahim Tahir ◽  
Waseem Siddique ◽  
Inamul Haq ◽  
Kamran Qureshi ◽  
Anwar Ul Haq Khan
Keyword(s):  
Heat Transfer ◽  
Numerical Investigation ◽  
Supercritical Water ◽  
Rod Bundle

Comparison of Three-Rod Bundle Data With Existing Heat-Transfer Correlations for Bare Vertical Tubes

2010 ◽  
Author(s):  
Krysten King ◽  
Amjad Farah ◽  
Sahil Gupta ◽  
Sarah Mokry ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Annular Channel ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Nuclear Systems ◽  
Heat Transfer Correlations ◽  
Vertical Tubes ◽  
Bare Tube

Many heat-transfer correlations exist for bare tubes cooled with SuperCritical Water (SCW). However, there is very few correlations that describe SCW heat transfer in bundles. Due to the lack of extensive data on bundles, a limited dataset on heat transfer in a SCW-cooled bundle was studied and analyzed using existing bare-tube correlations to find the best-fit correlation. This dataset was obtained by Razumovskiy et al. (National Technical University of Ukraine “KPI”) in SCW flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within a range from 800 to 3000 kg/m2s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2. The objective of this study is to compare bare-tube SCW heat-transfer correlations with the data on 1- and 3-rod bundles. This work is in support of SuperCritical Water-cooled Reactors (SCWRs) as one of the six concepts of Generation-IV nuclear systems. SCWRs will operate at pressures of ∼25MPa and inlet temperatures of 350°C.

Numerical Analysis of Supercritical Water Heat Transfer in Vertical Tube With Different Obstacles

2009 ◽  
Author(s):  
Bo Zhang ◽  
Jianqiang Shan ◽  
Jing Jiang
Keyword(s):  
Heat Transfer ◽  
Critical Point ◽  
Supercritical Water ◽  
Transfer Characteristic ◽  
Vertical Tube ◽  
Bulk Temperature ◽  
Transfer Enhancement ◽  
Heat Transfer Characteristics ◽  
Water Reactors ◽  
The Heat Transfer Coefficient

Supercritical Water Reactors (SCWRs) are essentially water reactors operating at pressure and temperature above critical point. The heat transfer coefficient is relative low when the bulk temperature is above the pseudo-critical point due to the properties of vapor-like fluid. To obtain better heat transfer characteristics, increasing the fluctuation using obstacles is the conventional method. Heat transfer characteristic in vertical tube with different obstacles is numerically investigated under supercritical condition. Numerical simulation is carried out with commercial CFD code Fluent 6.1 and adaptive grid. The results show that The RNG k-ε model with enhanced wall treatment can obtain a reliable result; the blockage ratio and the local temperature have large influence on the heat transfer enhancement. The influence region and decay trend of obstacles are also studied and compared with existing correlations.

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