Goon Cherl Park

PREDICTION AND MEASUREMENT OF LOCAL TWO-PHASE FLOW PARAMETERS IN A BOILING FLOW CHANNEL

A two-fluid model to predict subcooled boiling flow at low pressure is presented. Although considerable success has been achieved in good axial predictions, this study focuses on the capability of the model to predict local two-phase flow parameters within an annulus channel. Comparison of model predictions is made against local measurements carried out by our Korean collaborators. Although reasonable agreement of local profiles of the void fraction, interfacial area concentration, and bubble frequency were achieved, significant weakness of the model was evidenced in the prediction of the mean Sauter diameter, liquid, and vapor velocities. The formulation of a transport equation to account for the dynamically changing interfacial area concentration is proposed. Further modeling work is in progress to incorporate the bubble coalescence behaviour seen during experiments into the transport equation.

On Population Balance Approach for Subcooled Boiling Flow Prediction

The capability of using the population balance approach combined with a three-dimensional two-fluid model for predicting subcooled boiling flow is investigated. Experi-ments were conducted to study the local flow characteristics of subcooled boiling flow andto provide measured local two-phase flow parameters. Calculations were performed usingthe newly developed population balance boiling model to study the effects of variousfactors on numerical predication of local two-phase flow parameters in the subcooledboiling regime. Comparison of model predictions against local measurements was madefor the radial distribution of the bubble Sauter diameter and void fraction covering arange of different mass and heat fluxes and inlet subcooling temperatures. Additionalcomparison using recent active nucleation site density models and empirical relationshipsto determine the local bubble diameter adopted by other researchers was also investi-gated. Overall, good agreement was achieved between predictions and measurementsusing the newly formulated population balance approach based on the modified MUSIG(multiple-size-group) model for subcooled boiling and two-fluid model.@DOI: 10.1115/1.1857952#

RCCS Experiments and Validation for High-Temperature Gas-Cooled Reactor

Abstract An air-cooled helical coil reactor cavity cooling system (RCCS) unit immersed in the water pool was proposed to overcome the disadvantages of the weak cooling ability of an air-cooled RCCS and the complex structure of a water-cooled RCCS for the high-temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool-type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool, and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

ESTIMATION OF THE POWER PEAKING FACTOR IN A NUCLEAR REACTOR USING SUPPORT VECTOR MACHINES AND UNCERTAINTY ANALYSIS

Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models’ uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.

Experimental study for multidimensional ECC behaviors in downcomer annuli with direct vessel injection mode during the LBLOCA reflood phase

For the assessment of the multidimensional safety analysis codes various investigations have been performed to provide detailed information for the ECC (Emergency Core Coolant) behavior in a downcomer annulus. In the present study, the multidimensional ECC bypass phenomena in the downcomer annuli of the UPTF (Upper Plenum Test Facility) and APR1400 (Advanced Power Reactor 1400) geometries are studied to fill in the lack of knowledge about the phenomena that could occur in the DVI (Direct Vessel Injection) system downcomer. The experiments for the direct ECC bypass have been conducted in the transparent downcomer models of the UPTF and APR1400 using air and water. The flow patterns and the bypass mechanisms of the ECC bypass were identified and the characteristics of it in both downcomers were compared with each other. Based on the visual observations, the cross flow between the downward liquid film and circumferential gas was found to be the most important flow pattern of the bypass phenomena. An analysis for the flow regime was conducted and the Wallis parameters were introduced as the significant non-dimensional parameters of the multidimensional two-phase flow phenomena.

Multiphase flow modeling of molten material-vapor-liquid mixtures in thermal nonequilibrium

This paper presents a numerical model of multiphase flow of the mixtures of molten material-liquid-vapor, particularly in thermal nonequilibrium. It is a two-dimensional, transient, three-fluid model in Eulerian coordinates. The equations are solved numerically using the finite difference method that implicitly couples the rates of phase changes, momentum, and energy exchange to determine the pressure, density, and velocity fields. To examine the model’s ability to predict an experimental data, calculations have been performed for tests of pouring hot particles and molten material into a water pool. The predictions show good agreement with the experimental data. It appears, however, that the interfacial heat transfer and breakup of molten material need improved models that can be applied to such high temperature, high pressure, multiphase flow conditions.

Integral test and engineering analysis of coolant depletion during a large-break loss-of-coolant accident

This study concerns the development of an integrated calculation methodology with which to continually and consistently analyze the progression of an accident from the design-basis accident phase via core uncovery to the severe accident phase. The depletion rate of reactor coolant inventory was experimentally investigated after the safety injection failure during a large-break loss-of-coolant accident utilizing the Seoul National University Integral Test Facility (SNUF), which is scaled down to 1/6.4 in length and 1/178 in area from the APR1400 [Advanced Power Reactor 1400 MW(electric)]. The experimental results showed that the core coolant inventory decreased five times faster before than after the extinction of sweepout in the reactor downcomer, which is induced by the incoming steam from the intact cold legs. The sweepout occurred on top of the spillover from the downcomer region and expedited depletion of the core coolant inventory. The test result was simulated with the MAAP4 severe accident analysis code. The calculation results of the original MAAP4 deviated from the test data in terms of coolant inventory distribution in the test vessel. After the calculation algorithm of coolant level distribution was improved by including the subroutine of pseudo pressure buildup, which accounts for the differential pressure between the core and downcomer in MAAP4, the core melt progression was delayed by hundreds of seconds, and the code prediction was in reasonable agreement with the overall behavior of the SNUF experiment.

Study of the Design of Data Acquisition and Analysis Systems for Multi-purpose Regional Energy Systems

Recently, the smart grid has become a hot issue and interest in related power sources have increased accordingly. The implementation of a smart grid can enable many generation resources to be linked to the power system, including small-scale reactors for the purpose of co-generation. Research on small-scale reactors is being carried out all over the world. Similarly, Korea is also conducting research on multi-purpose regional energy systems using nuclear energy. This paper proposes a real-time data acquisition and analysis system for small-scale reactors, and is known as the REX-10 (Regional Energy rX 10 MVA). This analysis requires real-time simulations for the power system since it needs data communication with a remote REX-10. A RTDS (Real Time Digital Simulator) has been used for the simulation, and a SCADA/HMI system interfaced with the RTDS is proposed for the purpose of monitoring and control of the regional energy system.

Experiments on a water pool-type reactor cavity cooling system in a high-temperature gas-cooled reactor

In this study, a new concept in reactor cavity cooling systems (RCCSs) for high-temperature gas-cooled reactors (HTGRs) is proposed. The proposed RCCS consists of both water pools and active air-cooling systems, in order to overcome the disadvantages of the weak cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. The cooling capability of the RCCS during normal operation and under accident conditions was evaluated on the basis of a series of experiments that were performed in a scaled test facility. The reactor vessel of the test facility was a 1/10 linear scaled model of a 265-MW pebble bed modular reactor (PBMR), and the RCCS of the test facility was designed to limit the volumetric-averaged reactor vessel wall temperature below the maximum permissible wall temperature of the prototype reactor. The experiments were conducted by simulating the heat released from the reactor vessel wall to the RCCS. The power was reduced by 1/100 to preserve the heat flux, and the timescale was reduced by 1/10 to preserve the stored energy per volume. In the normal operation tests, detailed information on the temperature distribution and heat removal fraction of the upper pool and side pool was obtained. In the loss of all forced convection accident test, the passive afterheat removal capability of the RCCS was evaluated. These experimental results will be used to validate the reactor safety analysis codes and to evaluate the feasibility of the water pool-type RCCS.

Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.Copyright