Page 112 - Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

Design of experimental liquid-metal facilities 85

– HLM pump reliability
– Instrumentation
– Advanced fuels and irradiation testing
– Neutronics

3.2.1.1 HLM pool thermal-fluid-dynamic

According to Tarantino et al. (2015), the objectives of the activities in relation to pool
thermal fluid dynamics are twofold: (1) gathering experimental data in geometry and
with boundary conditions that may improve the knowledge of phenomena/processes
at component and system levels and (2) generating databases for supporting the devel-
opment and demonstrating the capability of computer codes to predict phenomena/
processes relevant for the design and safety.
The following, not exhaustive list of topics, is identified as relevant at component
and system levels:

l Flow patterns in forced convection, including
l mixing,
l stratification (inducing thermal stresses),
l stagnant zones,
l surface-level oscillations
l Transition to buoyancy-driven flow
l Natural convection flow including
l pressure drop
l surface-level oscillations
l Fluid structure interaction
l Thermal fatigue issue
l Sloshing due to seismic event tests

3.2.1.2 Fuel assembly thermal-fluid-dynamic

Thermal fluid dynamics of nuclear fuel assemblies has the objective to develop such
geometry of the assembly with spacers or wire, which will provide optimal conditions
for heat transfer between fuel rods and coolant (Di Piazza et al., 2016). Moreover, the
fuel assembly should demonstrate the capability to withstand irradiation, high temper-
atures, mechanical loads, and corrosion environment with minimal changes in stiffness
characteristics and geometry. The areas of investigations include the fuel assembly
thermal hydraulics and hydrodynamics for a wide range of operating conditions.
For design purposes, it is important to test the fuel assembly on the basis of thermal-
hydraulic parameters (i.e., pressure losses, flow distribution, velocity field, and clad
wall temperature distribution) and the geometric features, such as rod bundle lattice,
subchannel geometry, and spacer grids.
The following (incomplete) list of topics should be experimentally investigated for
supporting lead fast reactor system development:
l Heat transfer in forced and natural convection (including transition)
l Subchannel flow distribution