Experimental Study of Heat Transfer to Supercritical Pressure Water Flowing in a 2×2 Rod Bundle

An experiment has recently been performed at Xi’an Jiaotong University to study the wall temperature and pressure drop at supercritical pressures with upward flow of water inside a 2×2 rod bundle. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. Experimental parameters covered the pressure of 23–28 MPa, mass flux of 350–1000 kg/m2s and heat flux on the rod surface of 200–1000 kW/m2. According to the experimental data, it was found that the circumferential wall temperature distribution of a heated rod is not uniform. The temperature difference between the maximum and the minimum varies with heat flux and/or mass flux. Heat transfer characteristics of supercritical water in bundle were discussed with respect to various heat fluxes. The effect of heat flux on heat transfer in rod bundles is similar with that in tubes or annuli. In addition, flow resistance reflected in the form of pressure loss has also been studied. Experimental results showed that the total pressure drop increases with bulk enthalpy and mass flux. Four heat transfer correlations developed for supercritical pressures water were compared with the present test data. Predictions of Jackson correlation agrees closely with the experimental data.

Influence of Inlet Velocity Condition on Unsteady Flow Characteristics in Piping With a Short-Elbow at High Reynolds Number Condition

In design of the Japan Sodium-cooled Fast Reactor (JSFR), mean velocity of the coolant is approximately 9 m/s in the primary hot leg (H/L) piping which diameter is 1.27 m. The Reynolds number in the H/L piping reaches 4.2×107. Moreover, a short-elbow which has Rc/D = 1.0 (Rc: Curvature radius, D: Pipe diameter) is used in the hot leg piping in order to achieve compact plant layout and reduce plant construction cost. In the H/L piping, flow-induced vibration (FIV) is concerned due to excitation force which is caused by pressure fluctuation on the wall closely related with the velocity fluctuation in the short-elbow. In the previous study, relation between the flow separation and the pressure fluctuations in the short-elbow was revealed under the specific inlet condition with flat distribution of time-averaged axial velocity and relatively weak velocity fluctuation intensity in the pipe. However, the inlet velocity condition of the H/L in a reactor may have ununiformed profile with highly turbulent due to the complex geometry in reactor vessel (R/V). In this study, the influence of the inlet velocity condition on unsteady characteristics of velocity in the short-elbow was studied. Although the flow around the inlet of the H/L in R/V could not simulate completely, inlet velocity conditions were controlled by installing the perforated plate with plugging the flow-holes appropriately. Then expected flow patterns were made at 2D upstream position from the elbow inlet in the experiments. It was revealed that the inlet velocity profiles affected circumferential secondary flow and the secondary flows affected an area of flow separation at the elbow, by local velocity measurement by the PIV (particle image velocimetry). And it was found that the low frequent turbulence in the upstream piping remained downstream of the elbow though their intensity was attenuated.

An Improved Model to Study the Density Wave Oscillation in Tubes by Using Frequency Domain Method

Density Wave Oscillation (DWO) in tubes was usually studied by using the frequency domain method. However, in the conventional model, the heat storage of wall metal was usually neglected to simplify the complex solving process of transfer functions, which might cause unreasonable results when the tube wall had a thick wall or complex geometry structures. Hence, in the present paper, an improved mathematical model was proposed based on the frequency domain theory to theoretically study the DWO in tubes. The present model was an improvement of the conventional model. The most notable improvement in the present model was that the heat storage of the tube wall metal, the internal wall heat flux and the external wall heat flux were all considered as dynamic parameters. Based on the improvement, the prediction of the DWO in tubes by using the present model might be more accurate and reasonable than that by using the conventional model, and this was proved by the comparison of the results obtained with the two models to the experimental results gained from literature. In the present study, it was shown that both the present model and the conventional model could predict the DWO in tubes well when the tube wall was thin, and it was also found that the present model was more appropriate than the conventional model when the tube wall was thick. Both the thickness of the tube wall and the specific heat of tube wall metal play negative roles in the system stability.

Application of Film Dryout Model in Liquid Metal CHF Prediction

Liquid metals have been used as coolants of several kinds of nuclear reactors, and the prediction of critical heat flux (CHF) is rather important for the design, safety and economy of these reactors. A film dryout model considering the deposition and entrainment of droplets was established to obtain the CHF of liquid metal in annular flow flowing in tubes. The correlations of deposition rate, entrainment rate and so on for conventional fluids were used, and the initial entrainment fraction was determined according to experimental data. Results showed that the correlations for conventional fluids could be used for liquid metals approximately, but relatively large error might occur for large heat flux. The accuracy of this model for sodium and potassium was similar for small heat flux, but had some differences for large heat flux. Special correlations of deposition rate, entrainment rate and so on should be developed to predict the CHF of liquid metals more accurately.

Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

Thermal Modeling of the HTR-10 Using the RELAP5-3D Code

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.

Numerical Simulations on Supercritical Water Flow Instability

3-D simulation of supercritical water flow instability in parallel channels and a natural circulation loop are presented. Results are obtained for various heating powers. The results show that, in the natural circulation loop the steady state mass flow will firstly increase with the heating power and then decrease. And mass flow grows with the growing of the inlet temperature, decreases with the growing of system pressure. Under a large heat flux, the parallel channels will experience the flow instability of out phase mass flow oscillation. And the oscillation amplitude will grow with the growing of heating power. At last, the numerical simulations are validated by B.T. Swapnalee’s experience formula.

Evaluation of the Scaling Analysis Model of the EU-APR1400 Core Catcher Cooling System Test Facility

This paper describes the basic design features of the EU-APR1400 reactor core catcher cooling system and its test facility, and the associated scaling analysis model. An assessment of the validity of the scaling analysis using the preliminary performance test result of the test facility is described. This includes comparison of the predicted mass flow rate of the test loop as a function of the heat load to the facility, inlet flow subcooling and system pressure to the experimental results.

Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (13) Rising Bubble Motion Under Horizontal Vibration

An earthquake is one of the most serious phenomena for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically, to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In the case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and a shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake. Therefore, in this research project, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate the effects of vibration on bubbly flow in the components and construct an experimental database for validation, we performed visualization experiments of vertical bubbly flow in a rectangular water tank on which a sine wave vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of bubbles, shapes of bubbles and liquid velocity distributions around bubbles, were shown. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble tilt angle was almost same as the cycle of oscillation table velocity.

Experimental Study on Resistance Characteristics in a 3 × 3 Rod Bundle

Experimental study on resistance of air-water two-phase flow in a vertical 3 × 3 rod bundle was carried out under normal temperature and pressure. The rod diameter and pitch were 8 mm and 11 mm, respectively. The ranges of gas and liquid superficial velocity were 0.013∼3.763 m/s and 0.076∼1.792 m/s, respectively. The result indicated that the existing correlations for calculating frictional coefficient in the rod bundle and local resistance coefficient could not give favorable predictions on the single-phase experimental data. For the case of two-phase flow, eight correlations for calculating two-phase equivalent viscosity poorly predicted the frictional pressure drop, with the mean absolute errors around 60%. Meanwhile, the eight classical two-phase viscosity formulae were evaluated against the local pressure drop at spacer grid. It is shown that Dukler model predicted the experimental data well in the range of Rel<9000 while McAdams correlation was the best for Rel⩾9000. For all the experimental data, Dukler model provided the best prediction with MRE of 29.03%. Furthermore, approaches to calculate two-phase frictional pressure drop and local resistance were proposed by considering mass quality, two-phase Reynolds number and densities in homogenous flow model, resulting in a good agreement with the experimental data.