The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of a new (3D) diffusion code DALTON, which is coupled to the existing thermal-hydraulics code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5.

The behavior of pebble-bed reactors during a Loss Of Forced Cooling (LOFC) transient is of particular interest, since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature.

For validation purposes calculation results of normal operation, a LOFC transient and a Control Rod Withdrawal transient without SCRAM have been compared with experimental data obtained in the HTR-10. The code system has been applied to the 400 MW(thermal) Pebble Bed Modular Reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system.

It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) size pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100 ⁰C) as compared to its nominal value in the HTR-10, but reaches a maximum value of 1648 ⁰C during the Depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976 ⁰C) during normal operation.

For more information, please contact j.l.kloosterman@tudelft.nl.