g-Jung Youn

Single-phase convective heat transfer enhancement by spacer grids in a rod bundle

The spacer grids within a fuel assembly of a nuclear reactor core disrupt and re-establish the momentum and thermal boundary layers so that they enhance the local heat transfer within and downstream of the spacer grids. An experimental study in a 6×6 rod bundle has been performed to investigate the effects of spacer grids on the single-phase convective heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. The conventional correlations showed poor predictions of the heat transfer enhancement by spacer grids at low Reynolds numbers; in particular, the maximum heat transfer rate at the top end of the spacer grids was significantly overestimated. Furthermore, the conventional correlations did not properly account for the effects of the Reynolds numbers on the heat transfer enhancement. Therefore, more systematic experiments should be performed using various spacer grids with large blockage ratios at low Reynolds numbers, considering an early phase of the reflood conditions.

Advancement of Reactor Coolant Pump (RCP) Performance Verification Test in KAERI

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). The RCP Test Facility (RCPTF) has the capability to test a RCP under the operation condition of an Advanced Power Reactor 1400 MW (APR1400). The design values of the facility are 17.2 MPa, 343 °C, 11.7 m3/s, and 13 MW in maximum pressure, temperature, flow rate, and electrical power, respectively. In the facility, it is possible to perform a type test for a newly-developed RCP as well as a production test for a RCP before its installation in a nuclear power plant. For the production test, H-Q curves under the cold and hot conditions are acquired. For the type test, various transient tests are additionally performed including four types of seal transient tests, a thrust bearing transient test, a cost down test, and so on.To acquire H-Q curves of a RCP, the flow rate should be controlled by varying the flow resistance in the test loop. The RCPTF uses a Variable Restriction Orifice (VRO) whose flow area can be controlled by moving the two orifice plates installed in-parallel. The need for flow control valves and bypass lines was eliminated using the VRO such that the flow disturbance was minimized. The flow rate in the main loop of the RCPTF is measured by a standard venture flow meter. The flow rate in the RCPTF is very high and thus the venture flow meter could not be calibrated in the entire range of Reynolds number corresponding to the operating condition in the APR1400. The calibration was conducted at the Colorado Experiment Engineering Station Inc. (CEESI) in the USA where natural gas is used for a working fluid. If a discharge coefficient calibrated with the gas is applied in the test results performed using the water as a working fluid, a discrepancy can occur due to the static hole error. Therefore, the static hole error was compensated in the test results and the result shows the improvement.The effect of the temperature on the pressure pulsation amplitude was also evaluated. During a cold performance test and heat-up phase to the condition of a hot performance test, an abnormal increase in the pressure pulsation amplitude was observed near the specific temperature range. This is acoustic resonance phenomena that occur when a blade passing frequency of the RCP is proportional to the harmonic resonance frequency of the RCPTF.© 2014 ASME

Effect of Flow Blockage on the Coolability during Reflood in a 2 2 Rod Bundle

During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

Characteristics of Downcomer Boiling Phenomena During the Reflood Phase of a Postulated Large Break LOCA for the APR1400

Downcomer boiling phenomena in a conventional pressurized water reactor have an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling is being progressed in the reflood phase of a postulated LBLOCA. For this, the test facility was designed as a one side heated rectangular test section which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II) measurement of the local two-phase flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and parts of Phase-II were introduced.Copyright