The universal fuel performance code interface in serpent 2

The increasing interest in high fidelity modeling of power and test reactors drives the development of new tools for multi-physics modeling as well as the coupling capabilities of the current, well established, codes. The simultaneous modeling of fuel performance and core neutronics enables solving the coupled problem between fission power, fuel temperature and mechanical behavior of fuel elements at different parts of the system. To make such considerations easier, an universal fuel performance code interface has been implemented in the Monte Carlo reactor physics code Serpent 2. The interface allows to take in account the temperature and strain distributions, calculated by fuel performance codes, in the neutron transport simulations. Information is transmitted to Serpent 2 via a separate file containing 2D axisymmetric temperature and strain distributions in the fuel elements. The output in the coupled calculation mode includes 2D axisymmetric power distributions and axial fast fluence distributions in different lattice positions. Results from an initial test problem involving the depletion of a short 3D VVER assembly are presented and compared to a similar calculation without the thermal and mechanical feedback