Xiaoying Zhang’s research while affiliated with Sun Yat-Sen University and other places

To examine the thermo-hydraulic characteristics of the primary and secondary sides of a helical tube once-through steam generator (HCOTSG) under steady state conditions, a one-dimensional primary and secondary side heat balance calculation model is established. The accuracy of the numerical simulation is verified by a comparison with the results of a RELAP5 program. The thermo-hydraulic parameters of HCOTSGs are obtained using the calculation program. To study the detailed flow and heat transfer characteristics in helical tubes, a CFD simulation model of HCOTSGs secondary side is established. The velocity and temperature distributions of the fluid in the cross section of the helical tube are nonuniform. The temperature of the fluid inside the helix is higher than that outside the helix. The velocity of the fluid inside the helix is lower than that outside the helix. The fluid inside the helix begins to boil before that outside the helix.

Design improvement of two-phase natural circulation systems with thermoelectric generator (TEG) is proposed in this paper. In the systems designed, the TEG produces electricity with heat energy of the systems. The electricity is further supplied to a pump and a group of agitators. In this way, circulation flow in the systems and external heat transfer of heat exchangers can be enhanced. Feasibility and effectiveness of the designed systems are analyzed through numerical simulation. The results indicate that by regulating the allocation of electricity between different loads, enough inlet subcooling of the pump can be ensured. Thus, the potential cavitation can be avoided. Comparative analysis of the systems performance with that of the natural circulation systems are carried out. Analysis results indicate that circulation flow and heat transfer are markedly enhanced with the systems designed.

To investigate the startup characteristics of the supercritical pressure water-cooled reactor (SCWR) system, a complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the high-performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with a circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, turbine, reheaters, steam extraction and heaters at each stage were obtained. The calculation results showed that the startup sequence and startup thermal parameters could fully meet the expectations: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays superheated; the core inlet temperature can reach 280 °C after the adoption of the high-pressure and low-pressure heaters; and the inlet pressure of the high-pressure turbine can be kept constant. During the startup process, the maximum cladding surface temperature (MCST) remains below the limit of 650 °C, suggesting that the entire startup procedure is safe and reliable.

To analyze the typical characteristics of accidents occurring during the startup of a supercritical water reactor (SCWR), a comprehensive SCWR system model was established with the SCTRAN analysis code, based on models of the high-performance light-water reactor (HPLWR) steam cycle, SCWR circulation startup loop, passive safety systems, and CSR1000 core parameters. According to the sequence of the loss of coolant flow accident (LOFA), without a safety system and trigger signal of the safety system under the rated condition, a new trigger signal during the startup was designed and then used to assess “LOFA,” “uncontrolled CR withdrawal accident,” and “loss of coolant accident (LOCA)” during the startup process. The results show that the new trigger signal can ensure the effective and timely response of the safety system, as well as reactor safety during the startup process. Moreover, the maximum cladding surface temperature (MCST) of the reactor peaks (850 °C, well below the safety criterion of 1260 °C) at the end of the fourth stage of the startup process in the case of LOCA.

In this paper, the driving power conversion rate of a twophase natural circulation loop is analyzed numerically. In operation of the two-phase natural circulation loop, part of thermal energy is converted into driving power, that is, the gravity difference between steam and water in vertical loops. The power drives circulation flow in this loop and it is converted into kinetic energy of the fluid, which is partially dissipated eventually with flow resistance. In steady state, the conversion rate is equal to the dissipation rate. In this study, modeling and simulation of a general two-phase natural circulation loop are performed. The dissipation rate of kinetic energy with flow resistance is calculated, which is equal to the driving power conversion rate in value. Furthermore, the conversion rate from thermal energy to the power is obtained with the conversion rate calculated indirectly and heat transfer rate of the heat source. Simulation results indicate that the conversion rate can be as low as 0.01%. Based on this result, design improvement of the natural circulation loop is proposed. In this design, the thermoelectric material Bi2Te3 is attached outside part of the heat exchanger tubes. With the material, the temperature difference between fluid inside and outside the heat exchanger leads to voltage, and the heat exchanger with the material actually works as a thermoelectric generator (TEG). In this way, part of thermal energy is converted into electricity with an efficiency which is much greater than that of original two-phase natural circulation loop. The electricity converted is further used to enhance circulation flow of water through a pump and natural convection outside the heat exchanger with a group of agitators. With this design, the driving power conversion rate of the two-phase natural circulation loop is increased, the driving force is enhanced, circulation flow and heat transfer can be improved.

Startup system, the design of startup sequences analysis is an important part of SCWR design. A thermal hydraulic system analysis code for supercritical water reactor named SCTRAN is used to model the entire startup system based on the circulation loop for startup and once-through direct cycle. The problem of the heat transfer coefficient (HTC) does not accurately capture deterioration phenomenon, the HTC is calculated as a discontinuity in the mode transfer region, its low prediction accuracy above the quasi-critical region have been solved by the new wall heat transfer model. Especially, the look-up table would not be used to obtain the HTC and achieves high prediction accuracy across the critical region, unless the pressure is higher than 19 MPa. After that, to get a smooth recirculation variable pressure startup process, the system model integrates the control system which can controls the temperature, the steam drum water level, the thermal power, and the coolant flow rate. Based on the CSR1000 core and entire once-through direct cycle and circulation loop for startup, four stages under control systems, from low pressure to full power condition in recirculation startup process, were analyzed with code SCTRAN and wall heat transfer model was modified. The calculation results show that the recirculation system can startup from subcritical state to full power state without issue of CHF. The control system can control the parameters quite well and maximum cladding temperature (MCST) can be limited under 650 °C in the startup process. The modified SCTRAN code in this paper can further expand the computational range and computational accuracy. The full-scale control system can meet the needs of parameters expected response.

A thermal-hydraulic analysis model, coupling the flow and heat transfer of the primary and secondary sides, is developed to describe the thermal hydraulic behavior of OTSG (helically coiled Once-Through Steam Generator). This numerical model is developed based on the two-fluid model with distributed parameters method to predict all flow variables at each position along the tube. Correlations, validated a lot, are adopted to consider the effect of helical structure on the flow and heat transfer of both the primary and secondary sides in OTSG. The computational code THOSG (Thermal-Hydraulic analysis code of Once-through Steam Generator) for OTSG is then developed based on the numerical model with the modified SIMPLE algorithm. To benchmark the developed physic model and computer code, steady-state simulation of OTSG of IRIS reactor is conducted. Results are compared with RELAP5 code with respect to the design parameters. Comparisons indicate that the THOSG code can predict the thermal-hydraulic characteristics of OTSG well. In order to investigate the effect of changes in the input conditions on the output of the OTSG, transient analysis has been performed. Simulations of change in feedwater flow rate and temperature are conducted to assess the performance of OTSG. Both steady-state and transient results indicate that the THOSG code can be utilized for the design and performance analysis of OTSG. However, further analysis, coupling with the thermal-hydraulic behavior of the reactor core, is required to comprehensively evaluate the safety and reliability of OTSG design in nuclear power plant.

Most of the system codes for reactor Thermal hydraulics simulation are based on the two-fluid model. The accurate and efficient simulations of the two-fluid model are therefore important. This paper introduces a numerically convenient set of equations and the associated one-step coupled solution method for the two-phase two-fluid model. These equations are devised from the RELAP5 difference equations, but they are cast into a new form having great numerical effectiveness in forming a coupled matrix equation. A one-step solution of the coupled matrix equation is performed to obtain the whole unknown variables of the flow system. Proper numerical treatments, closure relations and coupling of wall conduction are accounted for by modeling boiling phase change convective heat transfer behavior. Finally, the numerical solutions of the new two-fluid model equations were successfully implemented and validated using the water faucet problem and a subcooled flow boiling experiment, and good agreements were obtained.

Supercritical-pressure light water-cooled reactors (SCWR) are the only water cooled reactor types in Generation IV nuclear energy systems. Startup systems, and their associated startup characteristic analyses, are important components of the SCWR design. To analyze the entire startup system, we propose a wall heat transfer model in a paper, based on the results from a supercritical transient analysis code named SCTRAN developed by Xi'an Jiao tong University. In this work, we propose a new heat transfer mode selection process. Additionally, the most appropriate heat transfer coefficient selection method is chosen from existing state-of-the-art methods. Within the model development section of the work, we solve the problem of discontinuous heat transfer coefficients in the logic transformation step. When the pressure is greater than 19 Mpa, a look-up table method is used to obtain the heat transfer coefficients with the best prediction accuracy across the critical region. Then, we describe a control strategy for the startup process that includes a description of the control objects for coolant flow rate, heat-exchange outlet temperature, system pressure, core thermal power, steam drum water-level and the once-through direct cycle loop inlet temperature. Different control schemes are set-up according to different control objectives of the startup phases. Based on CSR1000 reactor, an analytical model, which includes a circulation loop and once-through direct cycle loop is established, and four startup processes, with control systems, are proposed. The calculation results show that the thermal parameters of the circulation loop and the once-through direct cycle meets all expectations. The maximum cladding surface temperature remains below the limit temperature of 650°C. The feasibility of the startup scheme and the security of the startup process are verified.

This work aims to evaluate the pressure-control capacity of a passive suppression pool system applied in PWR containment. The numerical models construct for wall condensation and direct contact condensation based on Mixture model are firstly examined with the experimental data of POOLEX STB-28. The simulation under Mixture model show good accordance with experimental result. Under a double-end rupture of the main pipe of the first loop, the pressure and temperature evolutions under the effect of suppression pool are analyzed. During the operation of the suppression pool, the existence of non-condensable gas affects the efficiency of wall and direct contact condensation. The absolute pressure in the free area of the containment rises to 3.7 bar in 50 s then becomes stable. During a simulation period of 300 s, the gas temperature rises to the saturation temperature, while the variation of water temperature is less than 5 K.