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.

No abstract

A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.