Wang-Kee In

Numerical Computation of Heat Transfer Enhancement of a PWR Rod Bundle with Mixing Vane Spacers

A series of computational fluid dynamics (CFD) simulations has been conducted to analyze the heat transfer enhancement in a fully heated rod bundle with mixing-vane spacers. The predicted Nusselt numbers downstream of the split-vane spacer are compared with the available experimental measurements and with correlation. The CFD calculations at Re = 28000 and 42000 showed a lower heat transfer enhancement close to the space grid but a good agreement of the decay rate with the fully heated experimental data at ~6Dh downstream of the grid. The CFD simulations also showed a maximum enhancement of the heat transfer at 6 to 7Dh downstream of the split-vane spacer due to the multiple vortices predicted near the spacer. In addition, the present paper compares the thermal-hydraulic performance of two different mixing vane spacers, i.e., a split-vane spacer and a hybrid-vane spacer, based on CFD simulations at a pressurized water reactor’s operating conditions. The split vane is predicted to have a higher overall heat transfer enhancement but a lower local heat transfer far downstream of the spacer where the minimum departure from nucleate boiling ratio is anticipated.

Pressure Loss Coefficient of Spacer Grid for a Tight-Lattice Rod Bundle

Experimental and CFD (Computational Fluid Dynamics) analyses were conducted to determine the pressure loss coefficient of a spacer grid for a tight-lattice rod bundle simulating a dual-cooled annular fuel (DCAF) assembly. The DCAF is designed to have both internally and externally cooled channels by adopting annular pellet and dual claddings. The Korea Atomic Energy Research Institute proposed the DCAF assembly for the Korean optimum power reactor (OPR1000) which is a 12×12 square rod bundle with a rod pitch-to-diameter ratio (P/D) of 1.08. The pressure loss across the spacer grid specially designed for the 12×12 rod bundle is required in order to determine the inner to outer cooling channel flow split. Hence, the pressure drop of the spacer grid was measured using the full-size rod bundle for a Reynolds number ranging from 2e+04 to 2e+05. A CFD analysis was also performed to predict the pressure drop of the spacer grid used in the full-size bundle experiment. Only a single grid span of the test rod bundle was modeled in this CFD simulation. The grid loss coefficients by both the experiment and CFD analysis was shown to decrease as the Reynolds number increases. The measured loss coefficient was estimated from 1.30 to 1.50 for the normal operating condition of the reactor core with the DCAF, i.e., Re=3.5e+05. The CFD simulation predicted the grid loss coefficient being approximately 5% higher than the measured value.Copyright

CFD Simulation of PWR Subchannel Void Distribution Benchmark

A CFD study was performed to simulate the steady-state void distribution benchmark based on the NUPEC PWR Subchannel and Bundle Tests (PSBT). The void distribution benchmark provides measured void fraction data over a wide range of geometrical and operating conditions in a single subchannel and fuel bundle. This CFD study simulated the boiling flow in a single subchannel. A CFD code was used to predict the void distribution inside the single subchannel. The multiphase flow model used in this CFD analysis was a two-fluid model in which liquid (water) and vapor (steam) were considered as continuous and dispersed fluids, respectively. A wall boiling model was also employed to simulate bubble generation on a heated wall surface. The CFD prediction with a small diameter of vapor bubble shows a higher void fraction near the heated wall and a migration of void in the subchannel gap region. A measured CT image of void distribution indicated a locally higher void fraction near the heated wall for the test conditions of a subchannel averaged void fraction of less than about 20%. The CFD simulation predicted a subchannel averaged void fraction and fluid density which agree well with the measured ones for a low void condition.Copyright

Developed a Spacer Grid for the Future PWR Fuel Assembly by Considering the Thermal/Hydraulic and Mechanical/Structural Performance

The spacer grid is one of the structural components for the fuel assembly. In order to increase or extend the fuel life cycle, a spacer grid which has a much higher performance from the thermal/hydraulic and mechanical/structural point of view will be needed. From the thermal/hydraulic viewpoint, the CHF margin is very important in order to extend its life. Particularly, the mixing flow or cross flow between the subchannels have to be reinforced for this purpose. From the mechanical/structural viewpoint, the critical strength and the fuel rod support behaviour of a spacer grid are the same as the TH performance improvement for the next generation fuel. A computational fluid dynamics (CFD) analysis was performed to investigate the coolant mixing in a nuclear fuel bundle that is promoted by the mixing vane on the grid spacer. Single and multiple subchannels of one grid span of the fuel bundle were modeled to simulate a 5by5 rod array experiment with the mixing vane. The three-dimensional CFD models were generated by a structured multi-block method. The standard k-e turbulence model was used in the current CFD simulation since it is practically useful and converges well for the complex turbulent flow in a nuclear fuel bundle. The CFD predictions of the axial and lateral mean flow velocities showed a somewhat larger difference from the experimental results near the spacer but represented the overall characteristics of the coolant mixing well in a nuclear fuel bundle with the mixing vane. Comparison of the single and multiple subchannel predictions shows a good agreement for the flow characteristics in the central subchannel of the rod array. The simulation of the multiple subchannels shows a slightly off-centered swirl in the peripheral subchannels due to the external wall of the rod array. It also shows no significant swirl and crossflow in the wall subchannels and the corner subchannels. In addition to this, the impact and the stress analysis of a spacer grid are accomplished by the FE method. The FE model was created using I-DEAS [4], and the ABAQUS/explicit version 6.3 commercial code was used for the solver. The FE analysis procedure was established, the FE analyses results were verified by the experimental method. The developed spacer grid will be evaluated from the thermal/hydraulic and mechanical/structural design criteria.Copyright