Page 59 - Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

34 Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

State of the art
Presently, the adopted technology, applied to experimental facilities, mainly consists of
heating elements (i.e., heating cables) placed on the external surface of vessels and pip-
ing. High reliability is not mandatory for experimental loops/pools, while it is relevant
for LMFRs. As consequence, different systems will be proposed and adopted to
improve this feature. For ALFRED, for which the present issue is much more important
than ASTRID or MYRRHA, internal heating systems are under evaluation (i.e., electric
cartridge), to be placed in dummy assemblies surrounding the core. This solution is
mainly considered for the refueling or maintenance phase, while for the start-up, the
primary system is supposed to be heated-up with hot gas and/or adopting as inner source
term the steam generators and decay heat removal heat exchangers. The possibility to
use the reactor cavity to heat-up the whole system is also considered, even if it seems
to be not suitable for large pools.
Development needs
Experimental tests of heating elements placed in the dummy assemblies are required to derive
components performances as input for operational and safety evaluations. Experimental tests
are also needed to study long-term reliability of such components. Modeling is required for
addressing heating up by hot gas and/or by heating up through the reactor cavity.

Off-normal operation:
One-dimensional code improvements and natural circulation stability (see also Chapter 4)
l
Challenge
The large plena present in most LMFRs tend to transition to stratified states as the coolant
flow rate decreases and are usually completely stratified in natural convection states. In this
state, their behavior is difficult to describe with the 0-D and 1-D models available in existing
system thermal-hydraulic codes. Heat exchanges with neighboring structures (such as the
inner and outer vessels) are also hard to predict.
State of the art
With respect to natural circulation stability, both experimental and numerical work are being
performed as described by Roelofs et al. (2015b). These studies will obviously also contrib-
ute to improvements and further validation of one-dimensional system codes, subchannel
codes, CFD, and/or coupled multiscale thermal-hydraulic codes.

Development needs
While the use of finer models (CFD meshing or 3-D system-scale meshes) seems unavoidable
to model the transient behavior of reactor pools in the general case, simple models have been
implemented successfully in system codes to describe, for instance, the behavior of a
completely stratified hot pool. Cases for these models include the calculation, at reasonable
6
cost, of long-term cooling transients ( 10 s) for which employing a complex, multiscale
approach (see below) would be prohibitive.
l Multiscale thermal hydraulics (see also Chapter 7)
Challenge
The transition of an LMFR to natural convection may be affected by a number of com-
plex 3-D phenomena; jet dynamics and stratification in the hot and cold pool, rec-
irculation loops inside and between subassemblies in the core, and recirculation
loops in the interwrapper region. These phenomena are hard to model at the system
thermal-hydraulic scale.