Guanghui Su

Theoretical calculation of annular upward flow in a narrow annuli with bilateral heating

Based on separated flow, a theoretical three-fluids model predicting for annular upward flow in a vertical narrow annuli with bilateral heating has been developed in present paper. The theoretical model is based on fundamental conservation principles: the mass, momentum, and energy conservation equations of liquid films and the momentum conservation equation of vapor core. Through numerically solving the equations, liquid film thickness, radial velocity, and temperature distribution in liquid films, heat transfer coefficient of inner and outer tubes and axial pressure gradient are obtained. The predicted results are compared with the experimental data and good agreements between them are found. With same mass flow rate and heat flux, the thickness of liquid film in the annular narrow channel will decrease with decreasing the annular gap. The two-phase heat transfer coefficient will increase with the increase of heat flux and the decrease of the annular gap. That is, the heat transfer will be enhanced with small annular gap. The effects of outer wall heat flux on velocity and temperature in the outer liquid layer, thickness of outer liquid film and outer wall heat transfer coefficient are clear and obvious. The effects of outer wall heat flux on velocity and temperature in the inner liquid layer, thickness of inner liquid film and the inner wall heat transfer coefficient are very small; the similar effects of the inner wall heat flux are found. As the applications of the present model, the critical heat flux and critical quality are calculated.

Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

Comparison of Hydrogen Generation Rate between CORA-13 Test and MELCOR Simulation: Clad Solid-Phase Oxidation Models Using Self-Developed Code MYCOAC

Abstract To better understand the MELCOR oxidation and degradation models, past work compared the MELCOR model to a CORA experiment (CORA Test 13). These MELCOR analyses specifically focused on fuel bundle heatup and clad oxidation when compared to CORA test data. The comparison allowed the authors to investigate differences between hydrogen generation data and simulation results. Several potential reasons were considered for hydrogen generation rate differences, including MELCOR input power, heat transfer modeling, the clad solid-phase oxidation model, and the gaseous steam diffusion model. This work focuses on the possible uncertainty in the clad oxidation models used in MELCOR. First, the MELCOR nodalization approach for the CORA test was reviewed. Then, the temperature history and spatial variation were examined. One main focus was to consider other clad solid-phase oxidation models to compare the MELCOR models. This was accomplished by developing a separate model, MYCOAC, using MELCOR temperature predictions as input. Finally, the mass transfer resistance of steam diffusion to the clad surface was examined and found to be a small effect. While the Baker-Just solid-phase oxidation model showed better agreement with CORA data at low temperatures, the conclusion in this paper is that the oxidation models are not the major source of uncertainty in hydrogen generation rate differences. Future work will focus on heat transfer modeling of the CORA test.

ANALYSIS OF THE LIMITING SAFETY SYSTEM SETTINGS OF A FLUORIDE SALT-COOLED HIGH-TEMPERATURE TEST REACTOR

Abstract The fluoride salt–cooled high-temperature reactor (FHR) is an advanced reactor concept, which uses high-temperature TRISO fuel with a low-pressure liquid salt coolant. The design of a fluoride salt–cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress in both China and the United States. An FHTR based on a pebble bed core design with coolant temperature 600·C to 700·C is being planned for construction by the Chinese Academy of Sciences” Thorium Molten Salt Reactor Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal-hydraulic licensing analysis of an FHTR using SINAP”s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting safety system settings (LSSSs) considering uncertainties for forced convection and natural convection are obtained. The LSSS power and coolant outlet temperature, respectively, are 24.83 MW and 720·C for forced convection and 1.19 MW and 720·C for natural convection. The maximum temperature for the structural materials of 730·C is the most limiting constraint of the FHTR design.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a FHTR

The Fluoride-salt-cooled High-temperature Reactor (FHR) is an advanced reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Chinese Academy of Sciences (CAS)’s Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal hydraulic transient analyses of an FHTR using SINAP’s pebble core design as a reference case. A point kinetic model is calculated by developing a microcomputer code coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating basic transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that the SINAP’s pebble core design is an inherently safe reactor design.© 2014 ASME

Experimental Study on Heat Transfer of Single-Phase Flow and Boiling Two-Phase Flow in Vertical Narrow Annuli

Water single-phase and nucleate boiling heat transfer were experimentally investigated in vertical annuli with narrow gaps. The experimental data about water single-phase flow and boiling two-phase flow heat transfer in narrow annular channel were accumulated by two test sections with the narrow gaps of 1.0mm and 1.5mm. Empirical correlations to predict the heat transfer of the single-phase flow and boiling two-phase flow in the narrow annular channel were obtained, which were arranged in the forms of the Dittus-Boelter for heat transfer coefficients in a single-phase flow and the Jens-Lottes formula for a boiling two-phase flow in normal tubes, respectively. The mechanism of the difference between the normal channel and narrow annular channel were also explored. From experimental results, it was found that the turbulent heat transfer coefficients in narrow gaps are nearly the same to the normal channel in the experimental range, and the transition Reynolds number from a laminar flow to a turbulent flow in narrow annuli was much lower than that in normal channel, whereas the boiling heat transfer in narrow annular gap was greatly enhanced compared with the normal channel.Copyright

3D Flow Field Analysis for a Traditional II+ PWR Containment Under Normal Condition

Flow field analysis is a foundation to many thermal-dynamic phenomena in the nuclear containment. There are several ventilation systems under normal condition to assure the proper environment for staff and facilities, and the two main ventilation systems are the Reactor Pit Ventilation system (EVC) and the Containment Continuous Ventilation system (EVR). The fans of the ventilation systems recycle the air in the containment to various rooms, and cooling coils are cooled by the Nuclear Island Chilled Water system (DEG). In this paper, we simulated the 3D flow field in a Chinese traditional generation II+ PWR reactor containment under normal conditions with a commercial CFD software ANSYS FLUENT.According the actual geometry data of the reactor plant, we built the geometry model, including the EVC and EVR system, and they were checked by plant experts to assure authenticity. Proper maximum mesh sizes were set for different parts at the ICEM CFD, and the grid number was about ten millions. We used the fan model in the ANSYS FLUENT to simulate fans in the ventilations. The calculated values of flow rate in ventilation systems were in good agreement with the design values.Copyright

The Research on Core Melting Process: Oxidation

Beyond-design basis accidents in the AP1000 may result in reactor core melting and are therefore termed core melt accidents. The aim of this work is to develop a code to calculate and analyze the oxidation of a single fuel rod with total failures of engineered safeguard systems under a certain beyond-design basis accident such as a gigantic earthquake which can result in station blackout and then total loss of coolant flow. Using the code, the responses of the most dangerous fuel rod in the AP1000 were calculated under the accident. A discussion involving fuel pellets melting, cladding rupture and oxidation, and hydrogen production then was carried out, focused on DNBR during coolant pump coastdown, the cladding intactness under different flow rates in natural circulation, and the delay effect on cladding rupture due to cladding oxidation. By the analysis of calculated results, several suggestions on guaranteeing the security of fuel rods were provided.Copyright

Numerical Solution on Spherical Vacuum Bubble Collapse Using MPS Method

Single vacuum bubble collapse in subcooled water has been simulated using the moving particle semi-implicit (MPS) method in the present study. The liquid is described using moving particles, and the bubble-liquid interface was set to be the vacuum pressure boundary without interfacial heat mass transfer. The topological shape of the vacuum bubble is determined according to the location of interfacial particles. The time dependent bubble diameter, interfacial velocity, and bubble collapse time were obtained within a wide parametric range. Comparison with Rayleigh’s prediction indicates a good consistency, which validates the applicability and accuracy of the MPS method. The potential void-induced water hammer pressure pulse was also evaluated, which is instructive for the cavitation erosion study. The present paper discovers fundamental characteristics of vacuum bubble hydrodynamics, and it is also instructive for further applications of the MPS method to complicated bubble dynamics.

Study on the Onset of Nucleate Boiling in Narrow Annular Channel by Using Wavelet and Wavelet Neural Network

In this study, the local modulus maxima of cubic B-spline wavelet transform are introduced to determine the location of onset of nucleate boiling (ONB). Wavelet transformation has the ability of representing a function and revealing the properties of the function in the joint local regions of the time frequency space. Based on wavelet and artificial neural network, a Wavelet Neural Network (WNN) model predicting ONB for upward flow in vertical narrow annuli with bilateral heating has been developed. The WNN mode combining the properties of the wavelet transform and the advantages of Artificial Neural Networks (ANN) has some advantages of solving non-linear problem. The methods of establishing the model and training of wavelet neural network are discussed particularly in the article. The ONB prediction is investigated by WNN with distilled water flowing upward through narrow annular channels with 0.95 mm, 1.5 mm and 2.0mm gaps, respectively. The WNN prediction results have a good agreement with experimental data. At last, the main parametric trends of the ONB are analyzed by applying WNN. The influences of system pressure, mass flow velocity and wall superheat on ONB are obtained. Simulation and analysis results show that the network model can effectually predict ONB.Copyright