Numerical Analysis of the Flow and Heat Transfer in the Sub-Channel of Supercritical Water Reactor

Author(s):  
Ajoy Debbarma ◽  
K. M. Pandey

Research activities are ongoing for High performance light water reactor (HPLWR) with square double rows fuel assembly to develop nuclear power plants with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, there is still a big deficiency in understanding and prediction of heat transfer in supercritical fluids. This paper evaluates three-dimensional turbulent flow and convective heat transfer in a single-phase and steady-state sub-channel of HPLWR by using general computational fluid dynamics code, Ansys 14 Fluent. The major concern using supercritical water as work fluid is the heat transfer characteristics due to large variations of thermal properties of supercritical water near pseudo-critical line. In order to ensure the safety of operation in High performance light water reactor (HPLWR), heat transfer deterioration (HTD) must be avoided. Numerical results prove that the RNG k-e model with the enhanced near-wall treatment obtained the most satisfactory prediction and lead to satisfactory simulation results. The HPLWR Square fuel assembly has many square-shaped water rods, Out of four types of sub-channels; three sub-channels SC-1, SC-2 and SC-3 are investigated (adjacent to the side of the moderator flow channels (SC-1) (moderator tube and assembly gap), central sub-channels formed by four fuel rods (SC-2), adjacent to the corner of the moderator tube (SC-3). Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and water rod, fuel rod pitch to diameter ratio 1.1–1.4 with 8mm diameter are considered for simulation. Sub-channel analysis clarifies that coolant flow distribution becomes uniform when the gap width is set to 1.0 mm. was less than 620°C. Effects of various parameters, such as boundary conditions and pitch-to-diameter ratios, on the mixing phenomenon in sub-channels and heat transfer are investigated. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) in a sub-channel, it was found that HTC increases with P/D 1.1 first and then decreases with increasing P/D ratio. Apart from the basic geometry sub-channel, a square sub-channel with a wire-wrapped rod inside has been chosen to investigate the “wire effect”.

Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14×14 pressurized water reactor (PWR) spent nuclear fuel assembly within a square basket tube of a typical transport package were conducted using a commercial computational fluid dynamics package. The assembly is composed of 176 heat generating fuel rods and 5 larger guide tubes. The maximum cladding temperature was determined for a range of assembly heat generation rates and uniform basket wall temperatures, with both helium and nitrogen backfill gases. The results are compared with those from earlier simulations of a 7×7 boiling water reactor (BWR). Natural convection/radiation simulations exhibited measurably lower cladding temperatures only when nitrogen is the backfill gas and the wall temperature is below 100°C. The reduction in temperature is larger for the PWR assembly than it was for the BWR. For nitrogen backfill, a ten percent increase in the cladding emissivity (whose value is not well characterized) causes a 4.7% reduction in the maximum cladding to wall temperature difference in the PWR, compared to 4.3% in the BWR at a basket wall temperature of 400°C. Helium backfill exhibits reductions of 2.8% and 3.1% for PWR and BWR respectively. Simulations were performed in which each guide tube was replaced with four heat generating fuel rods, to give a homogeneous array. They show that the maximum cladding to wall temperature difference versus total heat generation within the assembly is not sensitive to this geometric variation.


Author(s):  
Mohsen Modirshanechi ◽  
Kamel Hooman ◽  
Iman Ashtiani Abdi ◽  
Pourya Forooghi

Convection heat transfer in upward flows of supercritical water in triangular tight fuel rod bundles is numerically investigated by using the commercial CFD code, ANSYS Fluent© 14.5. The fuel rod with an inner diameter of 7.6 mm and the pitch-to-diameter ratio (P/D) of 1.14 is studied for mass flux ranging between 550 and 1050 kg/m2s and heat flux of 560 kW/m2 at pressures of 25 MPa. V2F eddy viscosity turbulence model is used and, to isolate the effect of buoyancy, constant values are used for thermo-physical properties with Boussinesq approximation for the density variation with temperature in the momentum equations. The computed Nusselt number normalized by that of the same Reynolds number with no buoyancy against the buoyancy parameter proposed by Jackson and Hall’s criterion. Mentioned results are compared with V2F turbulence model whereas strong nonmonotonic variation of the thermo-physical properties as function of temperature have been applied to the commercial CFD code using user defined function (UDF) technique. A significant decrease in Nusselt number was observed in the range of 10-6<Grq/Reb3.425Prb0.8<5×10-6 before entering a serious heat transfer deterioration regime. Based on an analysis of the shear-stress distribution in the turbulent boundary layer and the significant variation of the specific heat across the turbulent boundary layer, it is found that the same mechanism that leads to impairment of turbulence production in concentric annular pipes is present in triangular lattice fuel rod bundles at supercritical pressure.


2013 ◽  
Vol 59 ◽  
pp. 211-223 ◽  
Author(s):  
Emmanuel Ampomah-Amoako ◽  
Edward H.K. Akaho ◽  
Benjamin J.B. Nyarko ◽  
Walter Ambrosini

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