Supplementing fuel behaviour analyses via coupled Monte Carlo neutronics and fission product solution

The accuracy of power distribution solution and burnup algorithms in standalone fuel behavior codes is limited. For improved accuracy, fuel behavior codes can be coupled with external reactor physics solvers. A coupled calculation system between the Monte Carlo neutronics and burnup calculation solver Serpent and the fuel behaviour solver TRANSURANUS developed in earlier work was further improved by developing the capability to transfer and utilize Serpent calculated nuclide compositions in TRANSURANUS within the coupled calculations. Previously the data exchanged between the solvers included fuel pin axial and radial power and temperature distributions, the axial fast flux distribution in the cladding and the changes in the pin radii. Additionally, support for corrector type burnup algorithms in the coupled solution was now implemented for increased accuracy of the nuclide solution. The new functionalities were accomplished via modifications to the coupled calculation driver program and Serpent and via development of a new data import interface in TRANSURANUS, which was also verified to function as intended. The new capabilities were demonstrated with a coupled single rod burnup calculation utilizing power and coolant history data from the Loviisa nuclear power plant. The demonstrations showed small but visible differences when compared to cases that were run without the nuclide transfer capability. The developed capabilities will now facilitate import of further fission product nuclides to be taken into account in TRANSURANUS.

RINTALA Ville;  SUIKKANEN Heikki;  SCHUBERT Arndt;  VAN UFFELEN Paul; 

2022-02-15

ELSEVIER SCIENCE SA

JRC126220

0029-5493 (online),   

https://www.sciencedirect.com/science/article/pii/S002954932200022X,    https://publications.jrc.ec.europa.eu/repository/handle/JRC126220,   

10.1016/j.nucengdes.2022.111668 (online),