Multiscale thermal hydraulic analysis of fuel assembly and system of SFR

As one of the innovative 4th generation nuclear reactors, the sodium cooled fast reactor (SFR) has unique advantages on heat transfer efficiency and breeding ratio. Due to the design requirement of a high fuel-coolant ratio, the fuel rods are usually arranged in triangular array. Helical wire spacers are widely used in the sodium cooled fuel assemblies. Another important feature is the pool type design of SFR. The most famous demonstration of SFR is Phenix reactor, which was closed in 2009. The upgraded future design version of Phenix is ASTRID [Le Coz et al. (2011)]. Comparing with water reactors which have a long term commercial operating experience, the special features of SFR result several unique detailed thermal hydraulic issues of fuel assembly and system. The engineering interested several points are listed as below. 1. Thermal hydraulic issues of the fuel assembly a) Sweeping ow induced by the special wire spacers b) Local heat transfer due to the triangular fuel rod arrangement and spacers c) Sub-channel analysis of fuel assembly by considering the above two points 2. Thermal hydraulic issue of the reactor system a) The local detailed 3D phenomena inside the reactor pool, which is coupled with the system behaviour. The thermal hydraulic of SFR fuel assembly is mainly influenced by its helical spacers and fuel rod arrangement pattern. One of the main effects of the wire spacers is enhancing the inter-channel cross flow in the triangular arrayed rod bundles. To study the characteristic of sweeping flow rate across the gaps between the sub-channels, the computational fluid dynamics (CFD) method is used in the present study to analyze the flow field in a 19-rod bundle with wire wraps. This CFD work only considers the fluid zone of one section of the rod bundle. The height of the CFD model with a periodic inlet and outlet boundary condition is one wire pitch. The k - $\omega$ SST model is used as the turbulent model. The wire effect on the velocity distribution of the flow across the interface between two sub-channels is investigated in the CFD study. Based on the force balance equation between the wire wake force and the surface friction force of rods in the circumferential direction, a theoretical model for estimating the average swirling flow rate driven by one single wire is proposed. The simulation shows that the ratio of rod pitch (P) to rod diameter (D) has a strong effect on the velocity of sweeping flow. The normalized sweeping flow velocity increases with the P/D ratio. The simulation agrees well with the proposed model. In addition, the effect of the ratio of wire pitch (H) to rod diameter has also an influence on the cross flow rate: A small H/D ratio leads to higher cross flow velocity. A correlation as a function of both P/D ratio and H/D ratio is proposed to predict the sweeping flow velocity by the present study. Relating to the sweeping flow and the triangular arrangement pattern, the circumferential heat transfer of fuel rod may have an uneven distribution rather than the uniform distribution assumed in most of the sub-channel approaches. Since the local flow field in the coolant flow passage is influenced by the wires, which induce the diversion flow, the circumferential heat transfer of the rod is strongly affected by both the arrangement and the wire spacer. To study the local heat transfer behavior of wire-wrapped rod bundle, the CFD code OpenFOAM is used in present study to figure out the detailed temperature and heat flux distributions on the surface of rod and wire. In present study, two CFD models for both bare rod bundle and wire wrapped 19-rod bundle are created. For the bare rod model, a periodical boundary condition is used to simulate the heat transfer of developed flow. For the wire wrapped 19-rod bundle case, the normal inlet outlet boundary is imposed to the fluid region with a length of two wire pitches. In the CFD studies, the models are split into solid and fluid regions. The k-omega SST model is used for turbulent simulation. The simulation result shows a cosine like local heat transfer distribution in the circumferential direction of the rod. Besides the detailed investigation of the triangular arrayed fuel assembly, the sub-channel approach code MATRA is improved for the coolant of sodium and local heat transfer calculation. The thermal properties functions for sodium are implemented. The pressure drop correlation, heat transfer correlation and turbulent mixing coefficient for sodium coolant from the literatures are also implemented. An automatic nodalization function is created for generating the input deck. Besides that, the proposed local heat transfer model is implemented for local cladding temperature and heat transfer coefficient output. Finally, the modified MATRA code is used for the fuel assembly analysis of ASTRID reactor. The thermal hydraulic of SFR reactor system is mainly influenced by its pool type design. The coolant loops of the sodium cooled reactor involve a strong local 3D phenomena in hot plenum and cold plenum, which are coupling with the system behavior. In present study, the system thermal hydraulic code ATHLET and CFD code OpenFOAM are coupled to simulate the dissymmetric test of the Phenix reactor. The simulation results are compared with experimental data. The main achievements of present study includes: (1) A new physical model is proposed for the sweeping flow induced by the wire spacers; (2) A new local heat transfer correlation is proposed for the triangular arrayed fuel rods; (3) The code MATRA is improved for simulating the fuel assemblies of the sodium cooled reactor; (4) The ATHLET and OpenFOAM coupling scheme is applied to analyze the dissymmetric test of the Phenix reactor.