Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the University of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation.
Borio di Tigliole, A., Bruni, J., Panza, F., Alloni, D., Cagnazzo, M., Magrotti, G., et al. (2012). Preliminary TRIGA Fuel Burn-Up Evaluation by Means of Monte Carlo Code and Computation Based on Total Energy Released During Reactor Operation. In International Conference on the Physics of Reactors 2012 (PHYSOR 2012) Advances in Reactor Physics (pp.976-984).
Preliminary TRIGA Fuel Burn-Up Evaluation by Means of Monte Carlo Code and Computation Based on Total Energy Released During Reactor Operation
CHIESA, DAVIDE;CLEMENZA, MASSIMILIANO;PATTAVINA, LUCA MARIA;Previtali, E;SISTI, MONICA;
2012
Abstract
Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the University of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation.
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