Experimentally-Benchmarked Computational Fluid Dynamics Simulations of an Array of Heated Rods Within a Square-Cross-Section Helium-Filled Pressure Vessel

2016 ◽  
Author(s):  
Dilesh Maharjan ◽  
Mustafa Hadj-Nacer ◽  
Narayana R. Chalasani ◽  
Miles Greiner
Keyword(s):  
Cross Section ◽  
Heat Generation ◽  
Pressure Vessel ◽  
Radiation Heat Transfer ◽  
Used Nuclear Fuel ◽  
Nuclear Fuel Assembly ◽  
Experimental Apparatus ◽  
Induced Motion ◽  
Dynamics Simulations ◽  
Square Array

An experimental apparatus was constructed, consisting of an 8×8 array of electrically-heated rods held in a square array by stainless-steel spacer plates near their ends. The rod/plate assembly was enclosed within a square-cross-section helium-filled aluminum pressure vessel and the rods were oriented vertically. The apparatus simulates the region between two consecutive spacer plates of a used nuclear fuel assembly within a vertical dry storage canister. Rod, spacer plate, and enclosure wall temperatures were measured using thermocouples in a matrix of nine experiments with total rod heat generation rates of 100, 300, and 500 W, and nominal helium pressures of 1, 2, and 3 atm. Steady-state simulations representing the experiment were performed, which include heat generation within the rods, conduction within the solid elements, as well as buoyancy-induced motion within, and natural convection and radiation heat transfer across, helium-filled regions. These were compared to the experimental results to assess the accuracy of the computational model for a range of boundary conditions. The comparison between the simulated and measured data showed that the simulations systematically under predict the hotter rod temperatures and over predict the cooler ones. Linear regression showed that 95% of the simulated temperatures are within 4.26°C of the correlation values.

Experimentally Benchmarked Computational Fluid Dynamics Simulations of a 7×7 Array of Heated Rods Within a Square-Cross-Section Enclosure Filled With Rarefied Helium

2017 ◽  
Author(s):  
Dilesh Maharjan ◽  
Mustafa Hadj-Nacer ◽  
Miles Greiner
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Cross Section ◽  
Heat Generation ◽  
Maximum Temperature ◽  
Helium Pressure ◽  
Nuclear Fuel Assembly ◽  
Jump Model ◽  
Computational Fluid Dynamics Simulations ◽  
Dynamics Simulations

Computational fluid dynamics simulations of a 7×7 array of heated rods within a square-cross-section enclosure filled with rarefied helium are performed for heat generation rates of 50 W and 100 W and various helium pressures ranging from 105 to 50 Pa. The model represents a section of nuclear fuel assembly between two consecutive spacer plates inside a nuclear canister subjected to during vacuum drying process. A temperature jump model is applied at the solid-gas interface to incorporate the effects of gas rarefaction at low pressures. The temperature predictions from simulations are compared to measured temperatures. The results showed that when helium pressure decreased from 105 to 50 Pa, the maximum temperature of the heater rod array increased by about 14 °C. The temperatures of the hottest rod predicted by simulations are within 4°C of the measured values for all pressures. The random difference of simulated rod temperatures from the measured rod temperatures are 3.33 °C and 2.62 °C for 100 W and 50 W heat generation rate.

Temperature Measurement of an Array of Heated Rods Subjected to Vacuum Drying Conditions

2017 ◽  
Author(s):  
Dilesh Maharjan ◽  
Mustafa Hadj-Nacer ◽  
Miles Greiner
Keyword(s):  
Temperature Jump ◽  
Maximum Temperature ◽  
Vacuum Drying ◽  
Helium Pressure ◽  
Used Nuclear Fuel ◽  
Experimental Apparatus ◽  
Computational Fluid Dynamics Simulations ◽  
Dynamics Simulations ◽  
Drying Conditions ◽  
Low Pressures

During vacuum drying of used nuclear fuel canister, helium pressure is decreased to as low as 67 Pa to promote evaporation and removal of water remaining in the canister following draining operation. At low pressures associated with vacuum drying, there is a temperature jump (thermal resistance) between the solid surfaces and helium in contact with them. This temperature jump increases as the pressure decreases (rarefied condition), which contributes to the fuel assembly’s temperature increase. It is important to keep the temperature of the fuel assemblies below 400°C during vacuum drying to ensure their safety for transport and storage. In this work, an experimental apparatus consisting of a 7×7 array of electrically heated rods maintained between two spacer plates and enclosed inside a square cross-section stainless steel pressure vessel is constructed to evaluate the temperature of the heater rods at different pressures. This geometry is relevant to a BWR fuel assembly between two consecutive spacer plates. Thermocouples are installed in each of the 49 heater rods, spacer plates and enclosure walls. They provide a complete temperature profile of the experiment. Different pressures and heat generation relevant to vacuum drying conditions are tested. The results showed that the maximum temperature of the heater rods increases as the pressure decreases. The results from these experiments will be compared to computational fluid dynamics simulations in a separate work.

Natural Convection/Radiation Heat Transfer Simulations of an Enclosed Array of Vertical Rods

2006 ◽  
Author(s):  
N. R. Chalasani ◽  
Miles Greiner
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heat Generation ◽  
Spent Nuclear Fuel ◽  
Radiation Heat Transfer ◽  
Three Dimensional ◽  
Radiation Heat ◽  
Simulation Techniques ◽  
Computational Fluid Dynamics Simulations ◽  
Dynamics Simulations

Experiments performed by others measured the temperature of twelve heated vertical rods within a constant temperature, internally finned cylindrical enclosure. Measurements were performed for a range of air and helium pressures and a range of rod heat generation rates. In the current work, three-dimensional computational fluid dynamics simulations of natural convection and radiation heat transfer within this domain were conducted to benchmark the simulation techniques. These calculations accurately reproduced the local and average temperatures when the heat generation rate was sufficiently low that the velocity field is steady. Future simulations will be used to design experiments that model spent nuclear fuel within non-isothermal cells of storage packages.

Benchmark of Computational Fluid Dynamics Simulations Using Temperatures Measured Within Enclosed Vertical and Horizontal Array of Heater Rods

2010 ◽  
Author(s):  
N. R. Chalasani ◽  
Miles Greiner
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Heat Generation ◽  
Spent Nuclear Fuel ◽  
Radiation Heat Transfer ◽  
Storage Conditions ◽  
Measured Temperature ◽  
Horizontal Orientation ◽  
The Difference ◽  
Dynamics Simulations

Experiments and computational fluid dynamics/radiation heat transfer simulations of an 8×8 array of heated rods within an aluminum enclosure are performed with nitrogen and helium as backfill gases in both horizontal and vertical orientations. This configuration represents a region inside the channel of a boiling water reactor fuel assembly between two consecutive spacer plates. The rods can be oriented horizontally or vertically to represent transport or storage conditions. The measured and simulated rod temperatures are compared for three different rod heat generation rates to assess the accuracy of the simulation technique. Simulations show that temperature gradients in the air are much steeper near the enclosure walls than they are near the center of the rod array. The measured temperatures of rods at symmetric locations are not identical, and the difference is larger for rods close to the wall than for those far from it. Small but uncontrolled deviations of the rod positions away from the design locations may cause these differences. The simulations reproduce the measured temperature profiles. For nitrogen experiment in horizontal orientation and a total rod heat generation rate of 500 W, the maximum rod-to-enclosure temperature difference is 138°C. The maximum measured heater rod and enclosure wall temperatures 375°C and 280°C, are measured in 2-inch insulated, nitrogen backfill vertical experiment for 1 atm internal pressure. Linear regression shows that the simulations slightly but systematically under predict the hotter rod temperatures but accurately predict the cooler ones. For all rod locations, heat generation rates, nitrogen and helium backfill gases, and apparatus orientations, 95% of the simulated temperatures are within 11°C of the correlation values. These results can be used to assess the accuracy of using simulations to design spent nuclear fuel transport and storage systems.

2D Natural Convection and Radiation Heat Transfer Simulations of a PWR Fuel Assembly Within a Constant Temperature Support Structure

2006 ◽  
Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Fuel Assembly ◽  
Heat Generation ◽  
Temperature Difference ◽  
Wall Temperature ◽  
Radiation Heat Transfer ◽  
Radiation Heat ◽  
Fuel Rods ◽  
Water Reactor

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14×14 pressurized water reactor (PWR) spent nuclear fuel assembly within a square basket tube of a typical transport package were conducted using a commercial computational fluid dynamics package. The assembly is composed of 176 heat generating fuel rods and 5 larger guide tubes. The maximum cladding temperature was determined for a range of assembly heat generation rates and uniform basket wall temperatures, with both helium and nitrogen backfill gases. The results are compared with those from earlier simulations of a 7×7 boiling water reactor (BWR). Natural convection/radiation simulations exhibited measurably lower cladding temperatures only when nitrogen is the backfill gas and the wall temperature is below 100°C. The reduction in temperature is larger for the PWR assembly than it was for the BWR. For nitrogen backfill, a ten percent increase in the cladding emissivity (whose value is not well characterized) causes a 4.7% reduction in the maximum cladding to wall temperature difference in the PWR, compared to 4.3% in the BWR at a basket wall temperature of 400°C. Helium backfill exhibits reductions of 2.8% and 3.1% for PWR and BWR respectively. Simulations were performed in which each guide tube was replaced with four heat generating fuel rods, to give a homogeneous array. They show that the maximum cladding to wall temperature difference versus total heat generation within the assembly is not sensitive to this geometric variation.

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