Abstract
The cooling of a nuclear reactor depends on a suitable fluid flow pattern among its fuel elements aiming the removal of heat produced in the fuel. In case of light water reactors, an excess of heat drives the fluid to change its phase from liquid to vapor, significantly reducing its capacity to remove heat and leading the reactor to a Loss of Coolant Accident. Numerical simulations using a CFD code is a suitable tool to address this kind of problem and explore the conditions that should be avoided during the reactor operation. The commercial CFD codes had proven to be reliable to simulate with a high accuracy and confidence the thermal-hydraulics of a sort of equipment and systems, avoiding spending efforts and financial resources in the development of new codes that, essentially, perform the same tasks. Despite of it, the CFD codes must be validated, such as against experimental results. To comply with this objective, a benchmark fuel element was purposed and experimentally essayed to provide experimental results for CFD codes calibration. The results of this essay are provided to the four types of subchannels for a 5x5 PWR fuel element, with results provided as density and void fraction. This work presentes the preliminary results obtained with CFD numerical simulations using the ANSYS-CFX® code for the central subchannel with active rods for stead state operation. The results demonstrated that the ANSYS-CFX® is adequate to simulate with high accuracy the flow in this subchannel.
Highlights
In nuclear reactors, the cooling fluid develops a very important task, ensuring that the heat produced by the fissions on nuclear fuel is suitable removed
The geometry refers to the central-typical subchannel of a 5x5 fuel element, with four quarters of active fuel rods located in each corner of the subchannel
Based on these initial results, it was concluded that the diameter that best fits with the experimental results are diameter of 0.5mm and 1.0mm, which were used in the remaining simulations
Summary
The cooling fluid develops a very important task, ensuring that the heat produced by the fissions on nuclear fuel is suitable removed. In Pressurized Water Reactors (PWRs), the cooling fluid should be kept in liquid phase to this, removing the heat efficiently from the fuel elements and its fuel rods. This liquid should flow through the spaces among the fuel rods; this space is named as subchannel [1]. A plug is developed, gradually blocking the flow of the coolant through the fuel element. This effect contributes to build up the heat that should be removed from fuel rods. Despite of PWR project avoids the vapor bubbles formation, it could be formed during transients, such as in case of accidents [1,2,3]
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