Chang-Hwan Shin

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

Operating Vibration Measurements of Test Fuel Assembly in Reactor Thermo-hydraulic Test Condition

The design verification of a newly-developed nuclear fuel assembly requires a long-term endurance test under thermo-hydraulic test condition simulating power reactor core. For this verification test, vibration of the test fuel assembly inside the simulated test core should be measured under high system temperature (over 200 °C), high system pressure (over 2.5 MPa) and fast moving coolant flow (over 5 m/s). To measure the vibration, we use specially fabricated accelerometers, various sealing techniques and conduit channel design for signal cable protection. The effects of the flow rate, coolant temperature, pre-sized support clearance on the test fuel vibration response and orbit motion were discussed. The measured data is used for fuel compatibility evaluation and a basis for endurance verification, as well as the validation tool for theoretical response prediction model.

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