Abstract
Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.
Highlights
The Gen-IV design approach for future reactor is based on improving safety, managing natural resources and optimizing economic performance
In order to study the dynamic aspects of the coupling between neutron physics and two-phase flow during an unprotected loss of cooling flow transients (ULOF) transient, an adiabatic kinetic model [3] is proposed completed by a coupling scheme based on the neutron code APOLLO3®[4], used to update periodically the spatial power distribution, and the system thermalhydraulics code CATHARE3 [5]
These negative feedbacks drive a slow global power decrease whereas the sodium quickly overheats until it reaches its saturation temperature (~900°C) which triggers the start of a new phase of the accident with a two-phase flow dynamics interacting with neutron physics
Summary
The Gen-IV design approach for future reactor is based on improving safety, managing natural resources and optimizing economic performance. With respect to safety of sodium cooled fast reactor, the gen-IV French SFR prototype ASTRID-600MWe proposes advanced features such as an in-vessel core catcher, an innovative nitrogen Brayton power conversion cycle and a low sodium void effect core design [1]. The sodium feedback effect is a key parameter to improve the core safety behavior during transients involving sodium concentration variations. In order to study the dynamic aspects of the coupling between neutron physics and two-phase flow during an ULOF transient, an adiabatic kinetic model [3] is proposed completed by a coupling scheme based on the neutron code APOLLO3®[4], used to update periodically the spatial power distribution, and the system thermalhydraulics code CATHARE3 [5]. The results obtained by the use of the coupling platform will be compared to those of decoupled calculations with constant power profile
Sign-in/Register to view highlights & summary Sign-in/Register to access full text options 
Free) 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a call
