An advanced computational tool based on Monte Carlo code was developed for the neutronic analysis of the TRIGA Mark II research reactor of the University of Pavia and the results from simulations were benchmarked with the experimental data. The neutron analysis was performed by means of the 3-D continuous-energy Monte Carlo codes MCNP4C and MCNP5. All core components were reproduced in good detail and the model allows the description of different core and fuel configurations with good versatility. In a first step, the results from the simulations with fresh fuel at room temperature were compared with the experimental data about critical core parameters at zero power level, quoted in the reactor First Criticality Final Report; a very good agreement was found and the goodness of the model was proved. Afterwards, neutron flux and reactivity dependence on the fuel and moderator temperature was studied in detail; all the available methods for the simulation of thermal effects were analyzed in order to model the reactor in the stationary operating condition at 250 kW. The model was updated to the present configuration and the use of specific S(α,β) cross sections was found to be of primary importance for a correct simulation of low-energy neutrons scattering within moderator.
Neutronic Analysis of the Pavia University TRIGA Mark II Reactor Core
BORIO DI TIGLIOLE, ANDREA;
2011-01-01
Abstract
An advanced computational tool based on Monte Carlo code was developed for the neutronic analysis of the TRIGA Mark II research reactor of the University of Pavia and the results from simulations were benchmarked with the experimental data. The neutron analysis was performed by means of the 3-D continuous-energy Monte Carlo codes MCNP4C and MCNP5. All core components were reproduced in good detail and the model allows the description of different core and fuel configurations with good versatility. In a first step, the results from the simulations with fresh fuel at room temperature were compared with the experimental data about critical core parameters at zero power level, quoted in the reactor First Criticality Final Report; a very good agreement was found and the goodness of the model was proved. Afterwards, neutron flux and reactivity dependence on the fuel and moderator temperature was studied in detail; all the available methods for the simulation of thermal effects were analyzed in order to model the reactor in the stationary operating condition at 250 kW. The model was updated to the present configuration and the use of specific S(α,β) cross sections was found to be of primary importance for a correct simulation of low-energy neutrons scattering within moderator.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
