Single- and Two-Phase Flow Modeling for Coupled Neutronics/Thermal-Hydraulics Transient Analysis of Advanced Sodium-Cooled Fast Reactors

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major R&D related issue. The current doctoral research aims at the development of a computational tool for the in-depth understanding of SFR core behavior during accidental transients, particularly those including boiling of the coolant. An accurate modeling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. Models for the representation of sodium two-phase flow are not present in most of the thermal-hydraulics codes used currently, and these have specifically been focused upon. The particular contributions of the present research are: (1) implementation of sodium two-phase flow models into the thermal-hydraulics code TRACE, which forms part of the FAST code system at PSI and, as such, can easily be coupled to the spatial neutron kinetics code PARCS; (2) validation of the TRACE sodium single- and two-phase flow modeling using out-of-pile sodium boiling experiments; (3) validation of the coupled TRACE/PARCS code system on the basis of experimental reactor data; and (4) application of the developed new tool for the core behavior analysis of an advanced SFR during a transient with boiling onset. The extension of the TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. A review has first been performed of the models available in the open literature for the representation of the interfacial and wall-to-fluid transfer mechanisms. The different correlations have then been implemented as options in the extended TRACE code. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the liquid film after dryout onset. The validation of the extended TRACE code has been achieved through the successful simulation of out-of-pile experiments