A whole core thermal-hydraulic analysis program ACT was developed for the purpose of evaluating detailed in-core thermal-hydraulic phenomena of sodium cooled fast reactors under various reactor operation conditions. ACT consists of four kinds of calculation modules, i.e., fuel-assembly, inter-wrapper gap (core barrel), upper plenum and heat transport system modules. The latter two modules give proper boundary conditions for the reactor core thermal-hydraulic analysis. These four modules are coupled with each other by using MPI and calculate simultaneously on a cluster workstation. ACT was applied to analyzing a sodium experiment performed at JNC, which simulated the natural circulation decay heat removal under PRACS and DRACS operation condition. In the experiment, not only inter-wrapper flows but also reverses flows in the fuel assemblies were observed. ACT succeeded in simulating such complicated phenomena.

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