Experimental study on the measurement of effective themal conductivity for VHTR fuel block geometry

Papers presented to the 11th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, South Africa, 20-23 July 2015.Effective thermal conductivity models which can be used to analyze the heat transfer phenomena of a prismatic fuel block were evaluated by the experiments. In the accident condition of VHTR when forced convection is lost, the heat flows in radial direction through the hexagonal fuel blocks that contain the large number of coolant holes and fuel compacts. Due to the complex geometry of fuel block and radiation heat transfer, the detail computation of heat transfer on the fuel block needs excessive computation resources. Therefore, the detail computation isn't appropriate for the lumped parameter code and a system code such as GAMMA+ adopts effective thermal conductivity model. Despite the complexity in heat transfer modes, the accurate analysis on the heat transfer in fuel block is necessary since it is directly relevant to the integrity of nuclear fuel embedded in fuel block. To satisfy the accurate analysis of complex heat transfer modes with limited computing sources, the credible effective thermal conductivity (ETC) models in which the effects of all of heat transfer modes are lumped is necessary. In this study, various ETC models were evaluated with the experiment result. Experiments for measuring the ETC values of the VHTR fuel block geometry were conducted with IG-11 graphite block. And four probable models compared to the experiment result showed good agreement with them, and thus they could be a candidate ETC model for VHTR fuel block.This research was supported by the National Nuclear R&D Program through the National Research Foundation of Korea(NRF) funded by MSIP; Ministry of Science ICT & Future Planning (No. NRF-2014M2A8A2074314)am201

Assesment of safety analysis code on integral effect test with SNUF

Paper presented at the 9th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Malta, 16-18 July, 2012.A Safety and Performance Analysis CodE (SPACE) is a thermal-hydraulics computer code under development of South Korea for analysis of complicated phenomena in the nuclear power plants including anticipated transients and postulated accidents. This study assessed the SPACE code capability for prediction of the direct vessel injection (DVI) system, which is adopted as a key safety system in the APR1400 reactor. This study mainly focused on the DVI line break accident, one of the postulated SBLOCA accidents that might result in loss of main coolant water and a quarter of safety injection water simultaneously, for assessment of the code. In order to evaluate the code prediction capability on the phenomena associated with the downcomer seal clearing at the DVI line break, this study selected the SNUF DVI break experiment as the experimental benchmark. This experiment is a reduced-height and reduced-pressure (RHRP) integral test facility designed for simulation of the primary loop of APR1400 designed in Korea. As a result, the SPACE showed reliable agreement with the experimental data on seal clearing phenomena well predicting both the start point of downcomer seal clearance and loop seal clearance. In the DVI system, downcomer seal clearing appears to be more important than loop seal clearing because the vapor generated from core flows through downcomer to broken DVI line. Therefore, the core collapsed level increases as the vapor pressure decreases in the coredc201

Assessment of Effect of Bubble Departure Frequency in Forced Convective Subcooled Boiling

Bubble departure frequency is one of the important parameters for the prediction of subcooled flow boiling. This present work aims at an assessment of bubble departure frequency by investigating the physical mechanisms of three-dimensional two-fluid model coupled with the population balance equation. The CFX MUltiple-SIze-Group (MUSIG) model is used to predict bubbly flows with the presence of heat and mass transfer processes, particularly in subcooled boiling flows at low pressures. The assessment is carried out for these three models/correlations. The test shows that Podowski et al.'s model, with reasonable physical characteristics, is more realistic than the other two models when compared with the experimental data. The numerical results indicate that the higher the departure frequency, the lower the wall temperature and so the nucleation site density. In addition it is found that for both the axial and radial cases the curves of the void fraction tend to decrease with increase in departure frequency. The benchmark of the current numerical simulation with experimental data in both axial and radial profiles achieves successful agreement

TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper

Conceptual Design of Regional Energy Reactor, REX-10

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SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10

Improvement of CUPID code for simulating filmwise steam condensation in the presence of noncondensable gases

In a nuclear reactor containment, wall condensation forms with noncondensable gases and their accumulation near the condensate film leads to a significant reduction in heat transfer. In the framework of nuclear reactor safety, the film condensation in the presence of noncondensable gases is of high relevance with regards to safety concerns as it is closely associated with peak pressure predictions for containment integrity and the performance of components installed for containment cooling in accident conditions. In the present study, CUPID code, which has been developed by KAERI for the analysis of transient two-phase flows in nuclear reactor components, is improved for simulating film condensation in the presence of noncondensable gases. In order to evaluate the condensate heat transfer accurately in a large system using the two-fluid model, a mass diffusion model, a liquid film model, and a wall film condensation model were implemented into CUPID. For the condensation simulation, a wall function approach with a heat/mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model, and then introduces the simulation result using the improved CUPID for a conceptual condensation problem in a large system

Flow visualization of bubble condensation in forced convective subcooled boiling flow

Subcooled boiling bubble condensation experiments were conducted in a vertical-upward annular channel by using water as the testing fluid at atmosphere pressure. The test runs comprised of bulk liquid temperatures, velocities and wall heat fluxes ranging from 75.0 degrees Celsius to 98.0 degrees Celsius, 0.25 m/s to 1.0 m/s and 150 kW/m2 to 200 kW/m2 respectively. A particle/droplet image analysis system was employed to capture the flow channel at four locations downstream of heated section for a total of 13 test conditions. The bubble Sauter-mean diameter was obtained in the range of 0.1 mm to 0.9 mm. It is also found that bubble sizes increase with the increase of liquid temperature or the decrease of liquid velocity. The condensation Nusselt number was calculated to be in the range of 10-4 to 10-1, which is much smaller than the typical range of 100 to102. This might due to the existence of non-condensable gas in the bubble

Flow Structure of Subcooled Boiling Water Flow in a Subchannel of 3 x 3 Rod Bundles

In this paper, the interfacial flow structure of subcooled water boiling flow in a subchannel of 3 x 3 rod bundles is presented. The 9 rods are positioned in a quadrangular assembly with a rod diameter of 8.2mm and a pitch distance of 16.6 mm. Local void fraction, interfacial area concentration, interfacial velocity, Sauter mean diameter, and liquid velocity have been measured using a conductivity probe and a Pitot tube in 20 locations inside one of the subchannels. A total of 53 flow conditions have been considered in the experimental dataset at atmospheric pressure conditions with a mass flow rate, heat flux, inlet temperature, and subcooled temperature ranges of 250–522 kg/m2 s, 25–185 kW/m2, 96.6–104.9ºC, and 2–11 K, respectively. The dataset has been used to analyze the effect of the heat flux and mass flow rate on the local flow parameters. In addition, the area-averaged data integrated over the whole subchannel have been used to validate some of the distribution parameter and drift velocity constitutive equations and interfacial area concentration correlations most used in the literatur