The aim of this work was to develop a methodological approach based on Monte Carlo code for the neutronic analysis of the first core configuration of the TRIGA Mark II research reactor of the University of Pavia and the validation of the results by benchmarking with the experimental and operational data exported in the reactor First Criticality Final Report. The neutron analysis was realised by means of the 3-D continuous-energy Monte Carlo code MCNP4C. The reactor model implemented represents in good detail all core components and allows the description of different core and fuel configurations with good versatility. Continuous-energy cross section data from ENDF/B-VI and ENDF/B-V and S(a, b) scattering functions from the ENDF/B-VI library were used. The results of the MCNP simulations and the experimental values show a very good agreement proving the goodness of the model.
Borio di Tigliole, A., Cammi, A., Clemenza, M., Memoli, V., Pattavina, L., Previtali, E. (2010). Benchmark evaluation of reactor critical parameters and neutron fluxes distributions at zero power for the TRIGA Mark II reactor of the University of Pavia using the Monte Carlo code MCNP. PROGRESS IN NUCLEAR ENERGY, 52(5), 494-502 [10.1016/j.pnucene.2009.11.002].
Benchmark evaluation of reactor critical parameters and neutron fluxes distributions at zero power for the TRIGA Mark II reactor of the University of Pavia using the Monte Carlo code MCNP
CLEMENZA, MASSIMILIANO;PATTAVINA, LUCA MARIA;Previtali, E.
2010
Abstract
The aim of this work was to develop a methodological approach based on Monte Carlo code for the neutronic analysis of the first core configuration of the TRIGA Mark II research reactor of the University of Pavia and the validation of the results by benchmarking with the experimental and operational data exported in the reactor First Criticality Final Report. The neutron analysis was realised by means of the 3-D continuous-energy Monte Carlo code MCNP4C. The reactor model implemented represents in good detail all core components and allows the description of different core and fuel configurations with good versatility. Continuous-energy cross section data from ENDF/B-VI and ENDF/B-V and S(a, b) scattering functions from the ENDF/B-VI library were used. The results of the MCNP simulations and the experimental values show a very good agreement proving the goodness of the model.
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