CFD in Supercritical Water-cooled Nuclear Reactor (SCWR) with - PDF document

The 13 th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) N13P1043 Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009. CFD in Supercritical Water-cooled Nuclear Reactor (SCWR) with Horizontal Tube Bundles Zhi Shang Kingston University Faculty of Engineering, London SW15 3DW, UK shangzhi@tsinghua.org.cn Simon Lo CD-adapco Trident House, Basil Hill Road, Didcot OX11 7HJ, UK simon.lo@uk.cd-adapco.com ABSTRACT The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k- ε turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculating region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR. KEYWORDS CFD, supercritical water reactor, turbulence model, heat transfer, tube bundle 1 / 14

The 13 th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) N13P1043 Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009. 1. INTRODUCTION Although heat transfer in supercritical water (SCW) flows has been studied for decades, the subject continues to receive great attention from the research community partly because of its scientific interest and partly because of the interest in using supercritical water in nuclear reactors. The Atomic Energy of Canada Limited (AECL) has announced a long-term plan to develop the supercritical water reactor (SCWR) CANDU [1] as their core nuclear reactor technology. Comparing with the current light water reactors (LWR) of generation III and III+ [1], the next generation (Generation IV) nuclear reactors will have much higher thermal efficiency. SCWR is a candidate for the Generation IV reactor and several designs have been studied by Oka et al and Mori et al [2-5]. Figures 1 and 2 show the two basic designs of SCWR. They can be divided into the vertical and horizontal flow types. The study of flow and heat transfer in SCWR must therefore consider these two flow orientation systems. The vertical system has already been studied by several researchers [8-11], the horizontal system is considered in this paper. Fig. 1 Vertical flow system in Japanese and European designs 2 / 14

The 13 th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) N13P1043 Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009. Multiple products are key to sustainable future and competitive designs Sustainable Electric power Electric power Fuel input Hydrogen and process heat Hydrogen and process heat Drinking Drinking water water Pump Generator Core Heat for Co- H.P Turbine Generation or T3, T3, T3, P3 P3 P3 IP/LP Turbines T2, T2, T2, P2 P2 P2 H.P. H.P. Turbine S S CONDENSER CONDENSER T1, T1, T1, P1 P1 P1 Brine Industrial isotopes Industrial isotopes Fig. 2 Horizontal flow system in the Canadian design Supercritical water has some very specific thermal-physical properties that lead to special heat transfer characteristics. At the supercritical pressures there is no phase change from liquid to vapor and the thermal-physical properties have sharp changes in the vicinity of the pseudo- critical temperature, see Figure 3. -4 1000 0.75 125 1.25x10 Molecular viscosity Density Thermal conductivity -4 800 0.60 100 1.00x10 Specific heat Thermal conductivity (W/m K) Molecular viscosity (Pa s) Specific heat (kJ/kg K) Density (kg/m 3 ) -5 600 0.45 75 7.50x10 -5 5.00x10 400 0.30 50 -5 200 0.15 25 2.50x10 0.00 0 0.00 0 0 0 0 0 200 200 200 200 400 400 400 400 600 600 600 600 800 800 800 800 1000 1000 1000 1000 1200 1200 1200 1200 o C) Bulk Temperature ( Fig. 3 Variations of thermal-physical properties of SCW under 25MPa In this paper, the thermal-physical properties are calculated based on the latest data published by the International Association of Properties of Water and Steam (IAPWS). According to the IAPWS data the thermal conductivity, density, molecular viscosity and specific heat are coded in the appropriate user subroutines in STAR-CD for calculations. 3 / 14

The 13 th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) N13P1043 Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009. 2. PHYSICAL PROBLEM AND COMPUTATIONAL PARAMETERS 2.1. Horizontal Tube Horizontal tube and tube bundle are studied in this paper. In a horizontal tube flow, buoyancy affects both the flow and heat transfer. This combined effect often leads to different wall temperatures and heat transfer coefficients on the top and bottom surfaces of the tube. The horizontal tube flow is fully three-dimensional (3D) and cannot be simplified to 2D or quasi- 3D flow as often done for vertical tube flows [8-11]. Figure 4 shows the geometry of the test case studied. The calculation parameters and boundary conditions are listed in Table 1 [7]. The flow enters the tube with uniform mass flux, is heated by the wall with uniform heat flux in Table 1 and flows out the tube from the outlet. STAR-CD can automatically choose the suitable outflow boundary conditions for the numerical simulations. Fig. 4 Geometry of the horizontal tube Table 1 Parameters of the horizontal tube test case Parameters (boundary conditions) Value Units kg/m 2 s 340 Mass flux (inlet) kW/m 2 300 Heat flux (wall) 24.4 MPa Pressure (reference) 4 / 14

The 13 th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) N13P1043 Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009. 2.2. Horizontal Tube Bundle in a Cylinder Figure 5 shows the geometry of the horizontal tube bundle in a cylinder. Table 2 shows the calculation parameters and boundary conditions used [11]. Again fully developed flow is assumed at the outlet. Figure 6 shows the computational mesh at a cross section. Finer meshes are employed in the near wall regions to resolve the flow and thermal boundary layers at the wall. Fig. 5 Geometry of the horizontal tube bundle in a cylinder Fig. 6 Mesh distribution at a cross section Table 2 Parameters of the horizontal tube bundle test case Parameters (boundary conditions) Value Units kg/m 2 s 1050 Mass flux (inlet) kW/m 2 600 Heat flux (wall) 25 MPa Pressure (reference) 5 / 14