FINE GEOMETRICAL MODELING OF THE NEUTRON TRANSPORT IN THE NPP KRŠKO'S FUEL, USING THE SERPENT MONTE CARLO TRANSPORT CODE
Abstract
The impact of some specific design features, such as grids, nozzles and in-core instrumentation thimbles on the neutron flux distribution was evaluated for the NPP Krško fuel. Fine geometrical models have been developed for the Serpent Monte Carlo code. The results obtained are compared to the CORD-2 calculations. The comparison has shown local variations of the neutron flux near the modelled material nonhomogeneities, which are however of minor importance for the global reactor results. There is no need for explicit geometrical treatment of the analysed fuel features in the CORD-2 system.
Downloads
References
J. Leppänen, PSG2 / Serpent – a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code, VTT Technical Research Centre of Finland, November 6, 2009.
A. Trkov, M. Ravnik, CORD-2 Package for PWR Nuclear Core Design Calculations, Proceedings of the International Conference on Reactor Physics and Reactor computations, Tel-Aviv, 23-26. Jan. 1994, Beer-Sheva, Ben-Gurion University of the Negev Press, (1994).
M. Kromar, A. Trkov, Nuclear Design Calculations of the NPP Krško core, Journal of Energy Technology, Volume 2, Issue 4, 2009, pp. 41–50.
