Investigation and Assessment of the CFD for Horizontal Flow in the VHTR Core

Abstract EN

A nuclear power station using gas as a cooling medium has attracted so much attention because it offers high efficiency and greater safety.

For a nuclear station that operates at a very high temperature, a gas-cooled reactor is fueled by uranium, moderated by graphite, and customarily cooled by helium.

Nevertheless, throughout the operation, the bypass flow might be a result of a change in graphite shape that is caused by neutron damage.

Core bypass and cross flows are significant elements to consider since the cross gap set hurdles to the flow field that are capable of diverting sufficient amount of coolant from reactor core location and initiating a possible fuel overheating.

However, there is a great need to sufficiently validate this method by carrying out a thorough evaluation based on working environment analysis.

Comparing the computed results with the existing data from Groehn’s NHDA PMR-200 study was the only way to validate data.

A model simulation was performed on a two-prismatic fuel block with a cross gap to examine the gaping size effect.

Finally, the prediction methods for horizontal flow phenomena using a CFD technique and the field investigation results from the VHTR core were verified, and the identification of the horizontal flow behavior played a vital role in investigating the coolant velocity and pressure distribution in the horizontal gap.