Numerical analysis of blockage effects in a liquid metal-cooled wire-wrapped fuel bundle

SCK•CEN Mentor

Van Tichelen Katrien,, +32 (0)14 33 80 06

Expert group

LBE Components and Experiments

SCK•CEN Co-mentor

Keijers Steven , , +32 (0)14 33 34 87

Short project description

Flow blockages in fuel bundles are a postulated accidental scenario with no-negligible probability. As the cooling in one or more of the bundle subchannels is impaired by the blockage, the cladding temperature increases locally and, in extreme cases, this might lead to fuel failure. For the safety evaluation of a reactor, it is paramount to guarantee the integrity of the fuel cladding in all scenarios, including the most-demanding full-power nominal operating conditions.

In the context of heavy liquid metal-cooled reactors, a first-of-its kind experimental campaign has been performed at the Karlsruhe Liquid Metal Laboratory (KALLA) of KIT in Germany, in collaboration with SCK•CEN and NRG in the frame of the FP7 MAXSIMA project. This work considered the experimental investigation of a partially blocked 19-rod bundle with wire spacers cooled by LBE, representative of the conditions expected in the MYRRHA reactor during nominal operation.

Based on experimental data from literature, the following features of the blockages can be listed as the most relevant ones for determining the thermal-hydraulic effects:

  • material properties, in particular the thermal conductivity. Moreover, depending on the origin of the particles, it should be noted whether heat is produced (e.g. by fuel debris) or not (e.g. Fe2O3 or PbO);

  • porosity is expected in the blockages, as they are generally formed by the accumulation of particles;

  • geometry: the column-like structure of blockages in wire-wrapped fuel bundles is represented by two parameters: the length and cross-sectional area, which can cover one or more subchannels;

  • location within the bundle, e.g. in internal or edge subchannel(s).

On top of these, the flow conditions (velocity, heat flux, ...) should be added. The KALLA experiments have been performed on a limited set of features following a conservative approach and considering the experimental capabilities (within the constraints of time and budget):

  • low thermal conductivity material representive of PbO was used. Heated blockages were not considered;

  • solid blockages elements were assumed;

  • 1 of 6 subchannels were blocked over a length of 1/6th of the wire pitch;

  • the 1-subchannel blockage was positioned in an internal and an edge subchannel.

The experiments allowed to establish the global and local temperature increase in the bundle as a function of flow velocity and heat flux. From this, one could conclude that 1-subchannel blockages seem to be acceptable from a safety perspective, while 6-subchannels blockages lead to local failure of cladding in full power conditions. However, it should be noted that these conclusions are based on the aggravating assumptions of low thermal conductivity and 0% porosity. In particular the latter is excessively conservative and strongly depends on the blockage formation mechanism. These mechanisms are studied in current research projects at SCK•CEN and at KIT. The assumption of non-heating blockages, on the other hand, is not conservative and needs to be considered.

Numerical simulations using a Computational Fluid Dynamics code of the unblocked and blocked experimental cases have been performed by NRG in the frame of the FP7 SEARCH and MAXSIMA projects. Comparison with the experimental data shows rather large deviations. Although these deviations are consistently conservative, their reduction through an extensive sensitivity analysis is mandatory. At the same time, the extension of the simulations to porous and self-heating blockages is of great value for the safety analysis.

The goal of the post-doc is to:

  • Develop and validate a CFD model of the MYRRHA fuel bundle with and without blockages on the basis of the experimental data available from the KALLA experiments. The work by NRG and other researchers can be used as a starting point, but needs a thorough analysis of sensitivity with respect to modeling choices (e.g. wire contact point, turbulence modeling, low Prandtl number heat transfer etc.) and experimental uncertainties (e.g. possible leakage paths, thermocouple positioning, etc.)

  • Extend and use this model to analyze the thermal hydraulic behavior of the fuel bundle in blocked and unblocked conditions in different operational states of MYRRHA, including porous and self-heating blockages. For this purpose, porous blockages need to be implemented in the model. Their geometrical features and likely location can be based on the ongoing research on blockage formation. Moreover, heat generation and conduction in the blockage material needs to be implemented as well.

The final objective is to provide safety analysts with data on fuel clad temperatures in blockage conditions with a clear indication of their degree of conservatism.

The post-doc candidate should have a strong background in nuclear thermal hydraulics and extensive expertise in the use of computational fluid dynamics codes in a research framework. Familiarity with the Ansys CFD software is mandatory; experience in using OpenFOAM is a strong plus.

Experience in the development and application of uncertainty quantification and sensitivity analysis  methodologies will be a large added value for SCK•CEN.

The minimum diploma level of the candidate needs to be


The candidate needs to have a background in

Electromechanics , Physics
Before applying, please consult the guidelines for application for Post-doc.