Byoung-Uhn Bae

Computational Analysis of a Subcooled Boiling Flow with a One-group Interfacial Area Transport Equation

For a multidimensional analysis of a two-phase flow, a computational fluid dynamics (CFD) code was developed with the implementation of an interfacial area transport equation that is beneficial for dynamically estimating the interfacial area concentration (IAC). The code structure was based on the two-fluid model and the Simplified Marker and Cell (SMAC) algorithm. The SMAC algorithm was extended to a two-phase flow simulation with a phase change. Various well-known constitutive models regarding boiling, condensation, and nondrag forces have been implemented into the code. To verify the robustness of the code to predict wall boiling and void propagation phenomena, a subcooled boiling test in a vertical annulus channel was analyzed as a benchmark problem. As the analysis results, a model for bubble departure diameter on the heated wall was identified as the principal factor for subcooled boiling phenomena, and the limitation of the current departure diameter models under a low-pressure condition resulted in a deviation of the void fraction and IAC when compared with the results of the experiment. It is necessary that the research on the interfacial area transport equation focuses on modeling reliable source terms for the boiling mechanism as a future work.

ASSESSMENT OF CONDENSATION HEAT TRANSFER MODEL TO EVALUATE PERFORMANCE OF THE PASSIVE AUXILIARY FEEDWATER SYSTEM

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

Development of a Multi-Dimensional Fluid Dynamics Code and Its Benchmarking for the Subcooled Boiling Flow

In a two-phase flow analysis, the interfacial area concentration (IAC) is a dominant factor governing the interfacial transfer of the momentum or energy. For a dynamic analysis with the implementation of IAC transport equation, a multi-dimensional computational fluid dynamics code was developed. The code is based on the two-fluid model and the simplified marker and cell algorithm by using the finite volume method, where the conventional approach for a single-phase flow has been modified in order to consider the term for a phase change. As benchmark problems of a single-phase flow and two-phase flow, a natural convection in a rectangular cavity and a subcooled boiling in an annulus channel were selected, respectively. In the calculation for the single-phase flow, the developed code predicted a reasonable behavior for a buoyancy-driven flow depending on the Rayleigh number. In the analysis of the subcooled boiling, the calculation results showed the robustness of code for the analysis of the boiling phenomena and void propagation, where they represented limitations of the one-dimensional IAC model. To conduct a multi-dimensional analysis for the two-phase flow, it is confirmed that the implementation of an IAC transport equation into the code is essential.

Experimental Investigation into the Effect of the Passive Condensation Cooling Tank Water Level in the Thermal Performance of the Passive Auxiliary Feedwater System

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+) and is designed to completely replace a conventional, active auxiliary feedwater system. With the aim of validating the cooling and operational performance of the PAFS, a separate effect test facility, the PAFS Condensing heat removal Assessment Loop (PASCAL), was constructed by simulating a single passive condensation heat exchanger (PCHX) tube submerged in the passive condensation cooling tank (PCCT) according to the volumetric scaling methodology. During heat removal of the PAFS, the pool water in the PCCT plays a role in the ultimate heat sink of a decay heat. In this study, the effect of the PCCT water level on the cooling performance of the PAFS was experimentally investigated with the PASCAL facility. Quasi-steady-state and PCCT level decrease test cases were sequentially performed by varying the steam generator heater power from 300 to 750 kW to investigate the thermal-hydraulic behavior during the decrease of the PCCT water level. From the experimental results, it was found that the decrease of the PCCT water level enhanced evaporative heat transfer at the outer wall of the PCHX tube by reducing the degree of subcooling around the PCHX. That induced an increase of the heat removal rate by the PCHX during the transient. Thus, it can be concluded that the current design of the PCHX in the PAFS has sufficient capacity to cool down the decay heat during the whole transient of the PCCT water level decrease.

Integral Effect Test on Operational Performance of the PAFS (Passive Auxiliary Feedwater System) for a SGTR (Steam Generator Tube Rupture) Accident

The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace a conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the PCCT (Passive Condensation Cooling Tank). With an aim of verifying the operational performance of the PAFS, the experimental program of an integral effect test is in progress at KAERI (Korea Atomic Energy Research Institute). The test facility, ATLAS-PAFS was constructed to experimentally investigate the thermal hydraulic behavior in the primary and secondary systems of the APR+ during a transient when the PAFS is actuated.Since the ATLAS-PAFS facility simulates a single train of the PAFS, the anticipated accident scenarios in the experiment include FLB (Feedwater Line Break), MSLB (Main Steam Line Break), and SGTR (Steam Generator Tube Rupture). Among them, SGTR was considered as one of the design basis accidents having a significant impact on safety in a viewpoint of radiological release. Therefore, the SGTR test was determined to be the integral effect test item in the frame of the ATLAS-PAFS experimental program.In this study, the PAFS-SGTR-HL-02 test was performed to simulate a double-ended rupture of a single U-tube in the hot side of the steam generator of the APR+. The three-level scaling methodology was taken into account to determine the test conditions of the steady-state and the transient. The pressures and temperatures of the system and the data related to the PAFS operation were collected with the measurement of the break flow.The initial steady-state conditions and the sequence of event of SGTR scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. And it was shown that the pressure and the temperature of the primary system were continuously decreased during the heat removal by the PAFS operation. The water pool in the PCCT was heated up to the saturation condition and the evaporation of the water made a decrease of the PCCT water level.It could be concluded from the present experimental result that the APR+ has the capability of coping with the hypothetical SGTR scenario with adopting the PAFS and the proper set-points of its operation.Copyright

Analysis of LSC phenomena of ATLAS cold leg SBLOCA tests using MARS-KS code

ABSTRACT Loop seal clearing (LSC) is an important phenomenon for the safety of a pressurized water reactor (PWR) during a small-break loss-of-coolant accident (SBLOCA). The investigation on an LSC phenomenon of 4″, 6″, and 8.5″ break cold leg SBLOCAs simulated by Advanced thermal–hydraulic Test Loop for Accident Simulation (ATLAS) is performed using a Multi-dimensional Analysis of Reactor Safety-KINS Standard (MARS-KS) code. The LSC triggers earlier for larger break sizes during tests and calculations. LSCs occur during the simultaneous sudden decrease of steam condensation rate and the sudden increase of the break volumetric flow rate while the core volumetric flow rate increases slowly in calculation. The five phases of an SBLOCA transient are blowdown, pressure plateau, LSC, boil-off, and core-recovery phase, which can be identified by observing the volumetric flow rate and the time-dependent pressure variation. Loop seal refilling (LSR) occurs due to the trivial steam flow rate to the crossover leg inlet in calculation. The sensitivity analysis shows that the combination of countercurrent flow limitation (CCFL) model option for hot leg and steam generator (SG) inlet (Kutateladze, c = 1.36, m = 1), crossover legs (Kutateladze, c = 1, m = 1), and SG U-tubes (Wallis, c = 1, m = 1) provide good prediction of the LSC phenomenon and thermal-hydraulics behaviors in an SBLOCA transient by MARS-KS code calculation.

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

ABSTRACT KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As an integral effect test database for major design basis accidents has been accumulated, a domestic standard problem (DSP) exercise using ATLAS was proposed in order to transfer the database to domestic nuclear industries and to contribute to improving the safety analysis technology for pressurized water reactors (PWRs). As the third DSP exercise, a double-ended guillotine break of the main steam-line at an 8% power without loss of off-site power was decided as a target scenario. Seventeen domestic organizations joined this DSP exercise. They include universities, government, and nuclear industries. The participants of DSP-03 were classified into three groups and each group has focused on the specific subject related to the enhancement of the code assessment; (1) scaling capability of the ATLAS test data by comparing with the code analysis for a prototype, (2) multi-dimensional thermal–hydraulic phenomena anticipated during the steam-line break transient, (3) effect of various models in the one-dimensional safety analysis codes.

Development of sweepout model in APR1400 during an LBLOCA

Abstract As a result of experiments with the Upper Plenum Test Facility and the 1400-MW(electric) Advanced Power Reactor (APR1400), sweepout in the downcomer has been identified as playing an important role in the depletion of the core coolant inventory during a large-break loss-of-coolant accident. In order to identify the sweepout mechanism and estimate the amount of coolant discharged during sweepout, separate-effects tests were performed in a rectangular-type test apparatus 1/5 the scale of the APR1400 downcomer. The experimental results showed that the sweepout was dominantly influenced by the hydraulic behaviors of coolant and steam near the intact cold leg. A sweepout model was developed by correlating the experimental results to analytically derived nondimensional parameters. The developed model showed applicability to the prototype, as the experimental results of the counterpart tests were in good agreement, within <25.0% of the uncertainty band.