Annual Report

port (RST) or Large Eddy Simulation (LES) models
- have to be used to predict these corner
zones where recirculation and corner vortices
occur. In order to improve the degree of confidence
with which these models can be used, it
is essential to conduct experimental measurements
in order to ensure that the assumptions
inherent in these models are well-founded.
It is possible to carry out very accurate measurements
within experimental ducts by using
transparent panels through which lasers can
penetrate. Particle Tracking Velocimetry (PTV)
or Particle Image Velocimetry (PIV) are modern
measurement techniques that present considerable
advantages over the more invasive Pitot
Tube and Hot-wire Anemometry methods used
in the past. The resolution in space and time
that can be achieved is more than sufficient to
capture the degree of detail needed to validate
numerical simulations.
Hybrid Methods for Thermal Hydraulic Analysis
of Natural Circulation Environments
Researcher: Morgan Cowper
Supervisor: Dr Michael Bluck
Sponsor: Rolls-Royce
Thermal hydraulic analysis of nuclear reactors
is largely performed with software known as
“system codes”. As a crude but reasonably accurate
shorthand, nuclear engineering systems
codes model the desired scenario in the form
of a series of 1D pipes , in which conservation
equations for mass, momentum and energy are
solved. For circumstances when this is a reasonable
approximation, systems codes are remarkably
effective, reflecting the very extensive
development to which they have been subject
over the past half-century. However, there are
many circumstances in which the environment
cannot adequately be represented as a piece of
pipe and the detail cannot be covered by a 1D
representation. CFD packages are capable of
providing a detailed three-dimensional model
of flow yet this comes at a great expense in the
form of computational time.
Therefore there is a growing interest in the
technique of coupling of the two modelling systems;
where a systems code can be used where
it is adequate with the more detailed CFD being
used where necessary. Although work has
been done on this subject for a number of years
it is still an embryonic aspect of computational
physics and there is vast room for improvement
and refining.
Advanced Component-scale CFD Modelling of
Nucleate Boiling
Researcher: Ronak Thakrar
Supervisor: Dr Simon Walker
Sponsor: Rolls-Royce
In a nuclear reactor, uncontrolled boiling can
lead to the hazardous condition often referred
to as the “critical heat flux” (or CHF) which can
result in the fuel clad no longer being wetted
and the integrity of the fuel being compromised;
this condition defines the upper limit
for safe reactor operation and therefore there
is an essential requirement for the engineering
capability to predict the onset of this condition
(and hence to accurately predict the behaviour
of boiling flows). Today’s state-of-the-art
in boiling modelling in CFD embodies a significant
degree of empiricism by way of using
correlations to define the overwhelming majority
of closure terms in the modelling formulation.
In addition to this, the formulation is itself
unrepresentative of the many interacting heat
transfer mechanisms at work during the boiling
process. These factors impose an upper
bound on the predictive capability (and range
of application) of the CFD code, which limits its
Centre for Nuclear Engineering Annual Report 2014-2016 44