The present article describes the preliminary design studies for PETALE (Programme d’Etude en Transmission de l’Acier Lourd et ses Eléments), an oncoming experimental program in the CROCUS reactor. Within the framework of the Venus-Eole-Proteus collaboration, PETALE continues the nuclear data validation efforts required for modeling GEN-III pressurized water reactors with heavy steel reflectors. The inelastic scattering cross sections at around 1 MeV of iron-56, as well as nickel and chromium isotopes, will be studied separately. The water reflector will be replaced successively by sheets of stainless steel alloy and pure metals—iron, nickel, and chromium. Data will be extracted from two sources: the measured neutron flux attenuation using adequate dosimetry and possibly fission chambers in the metal reflector and from the criticality effects of these reflectors. PETALE will also be used with nuclear data adjustment methods because, as a separated and elemental integral experiment, it allows the limiting of compensation effects in the nuclear data adjustments. A parametric study has been carried out with MCNPX for assessing the optimal configuration and the feasibility of the experiments. This study is the first step toward optimizing the global sensitivity of the experiments to the reactions in the energy range of interest, thus assessing the measurements’ target uncertainties and preparing further use of the program results.
The objective of this paper is to quantify the coupling effect on the power distribution of sodium-cooled fast reactors (SFRs), specifically the European SFR. Calculations are performed with several state-of-the-art reactor physics and Multiphysics codes (TRACE/PARCS, DYN3D, WIMS, COUNTHER and GeN-Foam) to build confidence in the methodologies and validity of results. Standalone neutronics calculations were generally in excellent agreement with a reference Monte Carlo-calculated power distribution (from Serpent). Next, the impact of coolant density and fuel temperature Doppler feedback was calculated. Reactivity coefficients for perturbations in the inlet temperature, flow rate and core power were shown to be negative with values of around -0.5 pcm/°C, -0.3 pcm/°C and -3.5 pcm/% respectively. Fuel temperature and coolant density feedback was found to introduce a roughly -1%/+1% in/out power tilt across the core. Calculations were then extended to axial expansion for cases where fuel is linked and unlinked to the clad. Core calculations are in good agreement with each other. The impact of differential fuel expansion is found to be larger for fuel both linked and unlinked to the clad, with the in/out power tilt increasing to around -4%/+2%. Thus, while broadly confirming the known result that standalone physics calculations give good results, the expansion coupling effect is perhaps more than anticipated a priori. These results provide a useful benchmark for the further development of Multiphysics codes and methodologies in support of advanced reactor calculations.
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