Search

International audience; The prospective energy mix scenarii generally imply a large contribution of renewable energy. The increasing use of solar and wind energies, which are intrinsically intermittent, actually constitutes a source of uncertain-ties and fragilities for the electrical grid. As nuclear energy produces heat before being converted into elec-tricity, a step of heat storage prior to the heat conversion step might be effective to remedy this intermittence in order to ensure the grid reliability and flexibility without involving large variations of the nuclear core power. Depending on the daily scenarii, the nuclear core could even stay at its maximal power all day long. Following this approach, a smaller core is then able to cope with the same peaks of demand than a larger one without heat storage systems.In this paper, a preliminary heat storage architecture coupled to a Sodium Fast Reactor is proposed in order to highlight the benefits of such a storage technology. The technical design based on two tanks respectively containing hot and cold fluids is inspired by current solar power technologies. The sizing of this system is carried out with a Thermodynamic Cycle Optimization tool (CYCLOP) and preliminary transients are simulated with the system thermal-hydraulics code CATHARE3. Even if some architecture improvements are still necessary, especially for safety related reasons, this study enables to draw the main benefits of such an electricity production strategy. In particular, it is shown that a variable electricity production while operating the reactor at base load is possible in load following condi-tions, thus enabling to optimize the plant profitability. As the impact on the primary circuit is shown to be negligible in terms of temperature evolutions, thermomechanical loading constraints in the vessel may also be drastically relaxed.

International audience; For some years now, CEA, namely the French Alternative Energies and Atomic Energy Commission, leads researches and developments to improve safety and availability of the GEN-IV Sodium-cooled Fast Reactors (SFR). One of these items of development concerns the core control and monitoring of the SFR Instrumentation and control system (IandC system). The Core Control and Monitoring System (CCMS) must be able to operate in all core conditions. The CCMS involves measurements, estimations and tracking of core control and monitoring parameters. It also commands the core control actions performed by the neutron absorber rods and primary system pumps. In this way, the CCMS uses all the means available to keep the operator informed of the state of the core components and core operation, while being able to detect events that could lead to the deterioration of core components. The measurements system of the CCMS comprises five main sub-systems devoted to take continuous measurements to insure a high level of both nuclear safety and power plant availability. Those five measurements sub-systems are the Neutronic Measurement System (NMS), the Thermal Measurement System (TMS), the Hydraulic Measurement System (HMS), the cladding Leaktightness Measurement System (LMS) and the core Geometry Measurement System (GMS). The five measurements sub-systems following features can be distinguished-NMS provides neutronic indications such as neutronic power and reactivity and relies on in-vessel high-temperature fission chambers. -TMS provides thermal indications such as inlet core or outlet subassembies coolant temperatures and relies on short response time thermal devices. -HMS provides hydraulic indications such as inlet core or outlet subassembies coolant flow rate and relies on miniaturized eddy current flowmeters. -LMS provides fuel cladding leaktightness defect indications (cladding rupture) and fuel cladding leaktightness defect location indications (ruptured sub-assembly). -GMS provides dynamic core geometry deformation indications (abnormal movements or vibrations of a fuel sub-assembly) and relies on high temperature ultrasonic transducers, high temperature strain gauges and accelerometer.This paper describes the definition and design of three of these five measurements sub-systems (TMS, HMS and GMS) which has been conducted in accordance to the methodology of systems engineering for the basic design of an innovative GEN-IV nuclear-fission power demonstrator called ASTRID. Moreover, in order to comply with their key operational requirements, the measurement sub-systems must be reliable and maintainable, that means capable respectively of achieving a continuous service and undergoing modifications and repairs. These dependability requirements completed with other operating safety requirements such as the inability of leading to damages to other vessel components (called innocuity in systems engineering). Then furthermore, this paper highlights how the measurements system development fulfils all the key GEN IV SFR requirements.

International audience; The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project reached in 2015 the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phenix and SuperPhenix Sodium Fast Reactors. The CFV core of ASTRID demonstrates an intrinsically safe behaviour with a negative sodium void worth achieved thanks to a new fuel sub-assembly design. This one comprises (U,Pu)O$_2$ and UO$_2$ axially heterogeneous fuel pins, a large cladding versus small spacer wire bundle, a sodium plenum above the fuel pins, and an upper neutron shielding with both enriched and natural boron carbide. The upper shielding also maintains a low secondary sodium activity level and is made removable on-line through the sub-assembly head for washing compatibility. Calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core compaction events. Concerning the radial shielding sub-assemblies surrounding the fuel core, heavy iterative studies have been performed in order to fulfill ASTRID requirements of minimising the secondary sodium activity level and maximising the in-core life-time. Evaluated options were reflectors sub-assemblies made of steel or MgO rods, and radial neutron shielding sub-assemblies made of B$_4$C or borated steel, with different configurations in the design and in the core layout. This paper describes the design of the fuel and radial shielding sub-assemblies for the ASTRID CFV v4 core at the end of the Conceptual Design phase. Focus is placed on innovations and specificities in the design compared with former French SFRs.

International audience; This paper highlights some recent achievements carried out at French Atomic and Alternative EnergiesCommission (CEA) in the framework of the R&D on the supercritical CO$_2$ (sCO$_2$) cycle for SodiumCooled Fast Reactors (SFRs). We discuss the sCO$_2$ cycle thermodynamic potential and in-depthinvestigations of some connected thermal-hydraulic issues regarding compression performancemodeling and cavitation phenomenology nearby the critical point. Outlines of these numerical studiesare the followings:– Net efficiency of a condensing cycle could be up to 44% for a turbine inlet operating at (515°C,25MPa).– A revised approach for compression performance maps representation is developed on a 2 inputparameters basis which complies with current treatment by system codes. This methodology isshown, using CFD, to succeed in modeling the impact of the fluid compressibility change oncompressor performance. This finding should also ease engineering work by reducing therequired number of component qualification tests. In support, ability of CFD to provide arelevant database has been validated by confrontation with some experimental data from a testcompressor.– Through analytical consideration of a characteristic parameter and further dynamic simulations,thermal regime is balanced to be the driving mechanism of bubble collapse near the criticalpoint due to combined effects of liquid-like high density and low thermal diffusivity. A veryslow contraction could therefore be foreseen, leading to the absence of noticeable pressure rise,in line with some experimental observations from the literature. Again, this outcome couldsupport future engineering work on cycle thermodynamic and compressor thermal-hydraulicdesigns.