[en] In the Korea Next Generation Reactor (KNGR) design, the Safety Depressurization System (SDS) plays a great role in reducing the core damage frequency and improving the severe accident performance of the KNGR. The actuation of POSRVs results in a time-varying high-energy flow of air, steam, two-phase, and liquid from the pressurizer into the Incontainment Refueling Water Storage Tank (IRWST). The successive discharge of water, air, and steam induces thermal hydraulic phenomena such as a water jet, air clearing and steam condensation, and these phenomena impose the relevant hydrodynamic forces on the IRWST structures. In the KNGR design, totally twelve 6inch I-type spargers will be installed in the IRWST. Therefore, an understanding of the related phenomena such as the characteristics of dynamic pressure loads is a prerequisite for the design of the sparger and IRWST structure to withstand the pressure loads. In the present study, four 2inch I-type spargers test were performed to characterize the pressure loads at the quench tank wall

[en] An experimental study has been carried out to investigate the performance of a I-type sparger in view of pressure oscillation and thermal mixing in a pool. Its pitch-to-hole diameter, P/D, varies from 2 to 5. The test conditions are restricted to the condensation oscillation regime. In the present study, two different hole patterns, staggered and parallel types, are employed under the various test conditions. The amplitude of the pressure pulse shows a peak at the pool temperature range of around 45 - 85 .deg. C, which depends on P/D and the steam mass flux. The effect of the hole pattern on the pressure load is relatively smaller than that of P/D. The results show that the dominant frequency increases with the subcooling temperature of pool water and P/D. A correlation of the dominant frequency is proposed in terms of the pitch-to-hole diameter ratio and other dimensionless thermal hydraulic parameters

[en] The number density transport equations for various bubble groups are used to predict the void fraction and the interfacial area concentration. As the closure relations for number density transport equation, the coalescence due to random collisions and the breakup due to the impact of turbulent eddies is modified based on the previous studies and the bubble expansion term due to the pressure reduction is considered. Also, the coalescence due to a wake entrainment is modeled newly to apply to the number density transport equation. In order to predict the local experimental data, the code is developed that the two-fluid model is coupled systematically with the number density transport equation for each bubble group. As for the results of the numerical analysis, the void fraction and interfacial area concentration are predicted well by the developed code and models although some deviations exist in the values between the prediction and experiment, especially, for the high void fraction conditions

[en] An experimental study on critical heat flux (CHF) has been performed for water flow in a uniformly heated vertical 3 by 3 rod bundle under low flow and a wide range of pressure conditions. Especially, this study has an objective to investigate the effects of unheated rods and system pressure under low flow conditions. The experiments have been conducted in the reactor coolant system thermal hydraulic loop facility (RCS loop facility) of KAERI. The test section consists of six heated rods and three unheated rods which have prototypical heated length of 3670 mm, diameter of 9.52 mm, and pitch of 12.6 mm of PWR type nuclear power plants. Total 163 CHF data have been obtained in the system pressure from 0.49 to 15.07 MPa, mass flux from 44 to 652 kg/m2s, inlet subcooling from 45 to 354 kJ/kg, and exit quality from 0.25 to 1.27. The general trends of the CHF are coincident with previous understandings. CHF increases as the mass flux or inlet subcooling increases. The effects of mass flux and inlet subcooling become large at low pressure and high mass flux, respectively. At low flow and system pressure above 3 MPa, some critical qualities are larger than 1 due to counter-current flow in the test section. Since there is a supply of water to the heated section from unheated section, the maximum CHFs at system pressure between 2 and 4 MPa are not shown

[en] In a nuclear reactor, spacer grids support the fuel rods and maintain proper geometrical configuration of fuel rods within a rod bundle assembly. The spacer grids alternate the thermal hydraulic behaviors near the spacer grids. They reduce the flow area by contracting the flow area and then expanding it downstream of the spacer grid. Thus, the flow and thermal boundary layers are disrupted and re established by the spacer grid. This enhances the local heat transfer within and downstream of the spacer grid. When single phase steam flow occurs in the early phase of the reflood, the cladding temperature may increase abruptly and reaches usually a maximum value due to the low heat transfer from the fuel to the steam. Hence, it is of importance to investigate the effect of the spacer grid on the heat transfer enhancement in single phase steam flow. Yao et al. reported that the heat transfer between wall and steam shows the maximum value at the top end of the spacer grid and that the Nusselt number decays exponentially downstream of the spacer grid. They developed an empirical correlation that only takes into account the flow blockage ratio and is applicable for simple egg crate types of grids

[en] Interfacial area concentration is one of the important parameters in the two-phase flow models. Five-sensor probe method is a useful measurement technique to measure the interfacial area concentration. It is essentially based on the four-sensor probe method but improves it by adapting one more sensor. The passing types of the interfaces through the sensors are categorized into four and independent methods are applied to the interfaces belonging to each category. To verify the applicability of the five-sensor probe method, benchmarking tests are performed for the rectangular visual channel by using the photographic method. The bubble velocity, void fraction, and Sauter mean diameter measured by the probe are also benchmarked. In this study, the design of the five-sensor conductivity probe, the signal processing procedure of the probe signal and the data analysis method by photography are also described

[en] An experimental study of the Critical Heat Flux (CHF) has been performed for a water flow in a non-uniformly heated vertical 3x3 rod bundle under low flow and a wide range of pressure conditions. Since most of experimental studies on the low flow CHF have been performed under low pressure conditions, present study has investigated the effects of various parameters on the CHF under low flow and a wide range of pressure conditions. Especially, these experiments are focused on the CHF under Return-To-Power (RTP) conditions that are expected to occur in a main steam line break accident of Pressurized Water Reactors (PWRs). Using present CHF data, the applicability of conventional CHF correlations are investigated in a return-to-power condition. The CHF data have been collected for system pressures ranging from 0.47 to 15.06 MPa, mass flux from 49.66 to 654.44 kg/m²s, inlet subcooling from 67.90 to 722.70 kJ/kg and exit quality from 0.36 to 1.29. In this study, the return-to-power conditions are defined as conditions with low mass flux less than 250 kg/m²s, intermediated pressure between 6.0 MPa and 12.0 MPa, and high inlet subcooling greater than 200 kJ/kg. Total 299 CHF data including 93 CHF data in return-to-power conditions are obtained. The effects of various parameters on the CHF are consistent with previous understandings on the round tube CHF. Conventional CHF correlations predict the present return-to-power CHF data with reasonable accuracies. However, the prediction capabilities become worse in a very low mass flux below than about 100 kg/m²s

[en] The fluidic device controls the discharge flow rate during LBLOCA (Large Break-Loss of Cooling Accident). For the assessment of the device performance, a prototypical full scale test facility, called VAPER (Valve Performance Evaluation test Rig), had been constructed, and the performance of the fluidic device had been validated without considering nitrogen solubility. Present paper shows the recent VAPER results including the solubility effect. Full scale experiments for the assessment of a fluidic device in SIT were performed. The influences of the N₂ gas solubility in ECC water on the K factor are analyzed. Finally, tests were done for various pressure, temperature and solubility conditions. As a result, a significant change in solubility did affect a little portion of the K factor in the low flow rate case

[en] Up to now, KAERI has performed critical heat flux experiments in water under zero-flow and low-flow conditions using a RCS CHF loop facility with uniformly and non-uniformly heated vertical annulus. Since the existing CHF experiments were mainly performed under low-pressure conditions, we performed the CHF experiment to investigate the pressure effect on the CHF under zero-flow and low-flow conditions for a wide range of system pressures. Also, two vertical annuli with the same geometry have been used to investigate the axial heat flux distributions on the CHF. This report summarizes the experimental results and provides the CHF data that can be used for the development for CHF correlation and a thermal hydraulic analysis code. The CHF data have been collected for system pressures ranging from 0.57 to 15.15 MPa, mass flux 0 and from 200 to 650 kg/m2s, inlet subcooling from 75 to 360 kJ/kg and exit quality from 0.07 to 0.57. At low-flow conditions, the total number of data are 242 and 290 with uniformly heated- and non-uniformly heated test sections, respectively. 41 and 94 CHF data are generated with uniformly heated- and non-uniformly heated test sections, respectively, in zero-flow CHF experiments that are performed by blocking test section bottoms. The CHF experiment result shows that the effects of system pressure, mass flux and inlet subcooling are consistent with conventional understandings and similar to those for round tubes. The behavior of the CHF is relatively complex at low pressures. Also, the effects of axial heat flux profile are large at low-pressure conditions

[en] A turbulent jet induced by a steam jet condensation in a water pool was investigated experimentally. An experimental apparatus equipped with a steam boiler, a single-hole steam sparger, and a water pool, etc. was used. For the measurements, a pitot tube and thermocouples were used for the turbulent flow velocity and temperatures, respectively. Overall flow shapes of the turbulent jet by the steam jet condensation are similar to those of axially symmetric turbulent jet flows. The angular coefficients of the turbulent rays are quantitatively comparable between the traditional turbulent jet flows and the turbulent jet flows induced by the steam jet condensation in this work. Although the turbulent flows were induced by a steam jet condensation, the general theory for turbulent jets was found to be applicable to the turbulent flows of this work