Suizheng Qiu

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.

Computational Fluid Dynamics Analysis of a Fluoride Salt-Cooled Pebble-Bed Test Reactor

Abstract The fluoride salt–cooled high-temperature reactor (FHR), combining high-temperature graphite-matrix coated-particle fuel (TRISO) for high-temperature gas-cooled reactors and liquid salts developed for molten salt reactors with safety systems that originate from sodium fast reactors, is a new concept reactor. The thermal-hydraulic characteristics of the fluoride salt–cooled high-temperature test reactor (FHTR) are of great importance to the development of the FHR technology, which is mainly ongoing in both China and the United States. In this paper, the thermal hydraulics of the FHTR designed by Shanghai Institute of Applied Physics is studied in different power modes. The one-dimensional temperature distributions of the coolant and the fuel pebble are obtained using a steady-state thermal-hydraulic analysis code for FHR. The detailed local flow and heat transfer are investigated by computational fluid dynamics for the locations that may have the maximum pebble temperature based on the results of a single-channel model. Profiles for temperature, velocity, pressure, and Nusselt number of the coolant on the surface of a pebble as well as the temperature distribution of a fuel pebble are obtained and analyzed. Numerical results indicate that the results of the three-dimensional simulation are in reasonable agreement with those of the single-channel model with a maximum deviation of 17.9%. They also illustrate the safety operation of FHTR in different power modes. This study aims to provide useful information for experimental and mechanism research of FHRs.

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.

An investigation of flow characteristics and critical heat flux in vertical upward round tube

Abstract Prediction of critical heat flux (CHF) in annular flow is important for the safety of once-through steam generator and the reactor core under accident conditions. The dryout in annular flow occurs at the point where the film is depleted due to entrainment, deposition, and evaporation. The film thickness, film mass flow rate along axial distribution, and CHF are calculated in vertical upward round tube on the basis of a separated flow model of annular flow. The theoretical CHF values are higher than those derived from experimental data, with error being within 30%.

Thermal-Hydraulic Analyses of Transportable Fluoride-Salt-Cooled High-Temperature Reactor with CFD Modeling

Abstract The technology for the 20-MW(thermal) Transportable Fluoride Salt–Cooled High-Temperature Reactor (TFHR) is proposed by Massachusetts Institute of Technology for off-grid applications such as Antarctic bases and remote mining sites. The preliminary thermal-hydraulic analyses and improvements based on a 1/12th full-core model were performed using three-dimensional computational fluid dynamics (CFD). A benchmark study was conducted by comparing the CFD results against empirical correlations and experimental data obtained by Cooke, Silverman, and Grele. In the 1/12th full-core analysis, three practical considerations that may challenge the TFHR temperature limits are evaluated as bounding analysis. These include (1) helium gap between fuel compact and graphite block, (2) thermal conductivity degradations of graphite matrix due to neutron irradiation, and (3) full-core scale power distribution obtained from neutronic calculations. These design considerations lead to insufficient margin between the normal operating condition and the predefined thermal limits. In this context, additional design features are implemented to improve the thermal-hydraulic safety of the TFHR. First, bypass flow in the interstitial gaps between the active core and the reflector is found capable of reducing the temperature peaks at the core periphery. Second, improvements of the flow distribution from the central downcomer to individual coolant channels enable a higher mass flow rate to the regions with compromised cooling access. Overall, thermal-hydraulic performance was significantly improved with a fuel temperature margin from 10 to 150 K and a coolant temperature margin from 16 to 160 K, as well as the more uniform temperature distribution across the reactor core. Furthermore, thermal-hydraulic safety can be maintained at a 20% overpower operating condition [i.e., 24 MW(thermal)]. Overall, this study provides an engineering basis for the TFHR thermal-hydraulic design to improve its safety margin.

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

The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

Abstract The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MI-TAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.

Experimental Investigation of Air-Water CCFL in the Pressurizer Surge Line of AP1000

Abstract Countercurrent flow limitation (CCFL) may occur under certain flow conditions in the surge line, restricting the draining of water from the pressurizer and thus affecting the coolant inventory and water level in the reactor pressure vessel (RPV). The complexity of the AP1000 pressurizer surge line structure makes predicting CCFL fairly difficult, and there are still not enough CCFL studies on this complex structure. Based on an extensive literature survey, the authors of this paper are particularly aware of the need for improved CCFL models for the pressurizer surge line of AP1000. To investigate the CCFL phenomenon in the surge line assembly fixture of AP1000, a whole-visual test model of the surge line is designed with a scaling ratio of 1:4, and a test loop is established to carry out visualization experiments with air-water countercurrent flow (CCF). The whole-visual test section made of acrylic material is composed of a pressurizer simulator, a surge line tube, a hot leg T-type tube, and an RPV simulator. The air-water CCF experiments are conducted at atmospheric pressure and room temperature with the pressurizer simulator water level varying from 150 to 900 mm. The visual CCF experimental processes and CCFL phenomena are filmed by a high-speed camera and analyzed in detail. The pressure drops at different CCFL locations are measured and evaluated to explore the relationships between the CCFL characteristics and flow patterns in the surge line. The development process of the CCFL is defined as the CCFL region, which can be divided into different regions according of the changes in water mass flow and CCF flow behavior. The CCFL data are analyzed and compared using the air and water superficial velocities to study the effects of hysteresis and water level. Small discrepancies are found between the data of different water levels, reflecting the small but not-negligible influence of the upper tank water level. Empirical models for the CCFL in the surge line assembly fixture are explored preliminarily using Kutateladze-type correlation and Froude-Ohnesorge correlation. Deficiencies still exist in the present semiempirical models, inspiring a more in-depth study on the empirical models for CCFL in the surge line assembly fixture that considers the complex two-phase flow behaviors in the upper tank and near the joint between the upper tank and surge line tube. The present CCFL data are compared broadly and in detail with groups of CCFL data of similar former experiments to demonstrate the applicability of the present air-water CCFL data to the development of a CCFL prediction model for the prototype large-diameter surge line assembly fixture of the AP600/AP1000. We will perform much more experimental and theoretical work to study the detailed mechanism of these special phenomena and to develop a more applicable CCFL model for the geometry and conditions of the prototype large-diameter surge line assembly fixture.