• heat
• plate-type
• transfer
• transient

Aim of this work is to investigate transient heat transfer phenomena of interest for
the thermal-hydraulic analysis of nuclear reactor core during the course of a postulated
reactivity insertion accident (RIA). An instantaneous extraction of control rods may cause the so called prompt jump, if
the reactivity insertion is larger than the reactor β, usually referred to as reactor dollar.
Associated to the prompt jump of the neutron flux, an exponential increase of the thermal
power produced within the fuel rods is observed. Consequently, the heat flux from the fuel
to the coolant also increases, making the coolant to heat up and eventually boiling. The
thermal-hydraulic feedbacks caused by the heating and, most of all, the formation of voids,
cause an insertion of negative reactivity that tends to stabilize the reactor core power. It may happen however, that before the thermal-hydraulic feedbacks produce a stabi-
lizing effect on the neutron flux, the energy stored within the fuel is large enough to cause
the fuel rod to melt down. Then, the fuel melted and released in the bulk of the coolant
may lead to the undesired phenomenon of steam explosion.
Whether the thermal power is stabilized to a safe level before fuel melting could occur
or not depends on the growth rate of the thermal power (and therefore the reactivity
inserted) and the time delay between the production of the thermal energy within the
fuel and the transfer of this energy to the coolant. Indeed, due to finite heat capacities
and thermal conductivities within the different structures (fuel and claddings), the heat
transferred to the coolant can be significantly lower than the heat released in the fuel at
the same moment, leading to an accumulation of energy within the fuel.
In order to contribute to the understanding and the prediction of power excursion
transient, a fundamental analysis of transient heat transfer phenomena has been performed.
Aim of this analysis is to provide a clear quantification of the rate of energy transfer
between the fuel and the coolant during the transient, as a function of the subchannel
geometry, material properties and the characteristic time scale of the exponential power
excursion (the so called period).
A multi-slices configuration is considered, representative of plate fuel type subchannels,
typical of experimental assemblies, like those of the SPERT and the BORAX reactors
described in chapter 2.
A two-step analysis is proposed. In a first instance, the analysis of purely conductive
systems is carried out. Then, the effects of coolant convection are investigated.
In the purely conductive case, both the numerical (CFD) and the analytical solutions
of the heat transfer equation have been obtained. The analytic solution has provided the
exact formulation for the spatial temperature profile all along the transient. Basing on the
temperature profile, other relevant quantities have been resumed, i.e. the disequilibrium
between the energy produced within the fuel end the energy transferred to the coolant
(defined as R) and the heat transfer coefficient between the solid structures and the coolant.
Last but not least, the mutual verification of the models has been achieved.
The CFD model, verified against the analytic solutions for the purely conductive
systems, has been thus applied to investigate convection effects associated to the coolant
motion.
The boundary conditions of the SPERT-IV experimental campaign have been thus
addressed in order to define the phenomenology of heat transfer associated to the different
tests. This aims at drawing guidelines for the application of correlations for the thermal
disequilibrium and the heat transfer coefficient in system codes and subchannel codes.