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International audience; Sensitivity analysis studies the effect of a change in a given parameter to a response function of the system under investigation. In reactor physics, this usually translates into the study of how cross sections and fission spectrum modifications affect the value of the multiplication factor, the delayed neutron fraction or the void coefficient for example. Generalized Perturbation Theory provides a useful tool for the assessment of adjoint weighed functions such as keff and void coefficient sensitivities. In this work, the capability of SERPENT code to perform sensitivity calculation based on GPT is used to study the TRIGA Mark II research reactor installed at L.E.N.A. of University of Pavia. A general sensitivity analysis to the most important reactor’s cross sections has been performed in order to highlight the biggest reactivity contributions. Two numerically challenging tasks related to GPT calculation have been performed thanks to the relatively quick Monte Carlo approach allowed by this reactor: investigating the linearity of the reactivity injection caused by the flooding of the central channel, and calculating the fuel void coefficient sensitivity to the coolant density.

Nuclear fuel plates based on a γU-Mo/Al mixture are proposed for research reactors. In this work their thermal behavior in the [425; 550°C] temperature range has been studied mainly by neutron and high energy X-ray diffraction. Even if complementary studies will be necessary, the kinetics of first the growth of the interaction layer between γU-Mo and Al and second of the γU-Mo destabilization have been accurately measured. This basic work should be helpful for defining manufacturing conditions for fuel plates with optimized composition.

International audience; The aim of this work is to validate experimentally the results of the Monte Carlo Recoil Energy Determination (MC-RED) nuclear physics code used to determine the deposited energy in a silicon volume taking into account the probabilistic approach of the physical phenomenon. A silicon sensor has been used to measure the deposited energy spectrum after an irradiation with a 14 MeV neutron source. Neutrons of 14 MeV were produced by the SAMES accelerator of the Valduc CEA research center. The experimental results were compared with the one obtained by the MC-RED code in the same silicon volume. To compare experiment and simulations, it requires that the complete neutron field be specified. We have to take into account the facility cell environment. Nuclear simulations were performed with the MCNP code to determine with accuracy the energy spectra of neutrons and gamma rays produced by this facility on the sensor. It is shown that the contribution of secondary neutrons is not negligible