Advanced method for neutronic simulation of control rods in sodium fast reactors: Numerical and experimental validation

A hybrid approach for neutronics calculations in the neutron shielding of sodium fast reactors

Neutron calculations in the neutron shielding of fast neutron reactors are a complex problem as deterministic schemes are usually not suited for such calculations while Monte-Carlo codes have poor computational performance due to the very low flux levels in neutron shields. In this article, both methods are studied, as well as a hybrid scheme on the neutron shielding of the ASTRID fast reactor benchmark. This hybrid scheme uses a fission source calculated by a deterministic code in order to precisely calculate neutron fluxes in the shielding with a Monte-Carlo code using variance reduction techniques. This provides reference results in order to validate deterministic calculations. Comparisons between deterministic codes and this hybrid reference show that large biases are obtained, up to 50%. Further studies are made to reduce the biases, showing that many physical phenomena should be treated, including anisotropy of the scattering law at high energies and spatial self-shielding inside the boron carbide shielding. These improvements reduce the biases to less than 10%. Finally, some applications to designing criteria for the neutron shielding are presented, such as gas production in the neutron shielding and activation of secondary sodium at the intermediate heat exchanger (IHX).

ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.