Numerical Simulation Research On Thermal-Hydraulic Characteristics Of Primary Loop In Integrated Small Modular Pressurized Water Reactor

The Integrated Small Modular pressurizer water Reactors(SMRs)have been highly valued by the international nuclear energy community for its enhanced safety,convenient construction and flexible deployment.Accelerating the construction of SMRs is imperative for the further development of China's nuclear industry.As a new type of advanced nuclear energy system,key technologies related to SMRs are still in the stage of exploration and tackling.The current progress of SMRs in China is also being promoted.The analysis of thermal-hydraulic characteristics of SMRs plays an important role in designing and assessing the safety characteristics of SMRs.In this paper,the corresponding thermal-hydraulic analysis models and codes are established and developed for main devices of SMRs' primary loop.The thermal-hydraulic characteristics of the IRIS(International Reactor Innovative and Secure)small module reactor are studied at different levels.Firstly,for thermal-hydraulic characteristics of SMRs' core,a more stable subchannel analysis code for lower power and small flow reactor core is constructed.Based on the traditional subchannel analysis model,the developed model couples the mass and momentum equations to reconstruct a more stable lateral momentum equation by considering the instability of subchannel analysis model,which is caused by the lateral flow in SMRs,and correlation between the governing equations.Based on the constructed subchannel analysis model,the subchannel analysis code SRSC is developed and relevant validations are carried out.Validations indicate that the developed subchannel analysis model and code have higher accuracy and stability,can predict thermal-hydraulic characteristics of SMRs well.The SRSC code is then applied to the safety analysis of the steady-state and typical loss of flow accidents of IRIS reactor.The steady-state results show that the pressure drop of each channel in the core is the same.Although the temperature,enthalpy and quality of the hottest channel are the highest,coolant in hottest channel is still subcooled.The coolant mass flow rate at the outlet of the hottest channel is the least due to the balance of the pressure drop by lateral flow,whereas the velocity is the highest.The spacers in the reactor core enhance the safety of the core by balancing the flow field and temperature field.Analysis of loss of flow accidents indicates that the design flow of the IRIS primary loop system can effectively remove the core decay heat.The minimum departure from nucleate boiling ratio is always higher than the safety limit of 1.5.The highest central temperature of core fuel rod and surface temperature of cladding is always below the material melting temperature which meets the safety criterionSecondly,for thermal-hydraulic characteristics of SMRs' helical coil Once-Through Steam Generator(OTSG),a thermal-hydraulic analysis model,coupling the flow and heat transfer of the primary and secondary side is developed.The two-fluid model with distributed parameters method is used to develop the model by considering the actual structure and operation of OTSG.The thermal-hydraulic analysis code THOSG for OTSG is also developed based on the analysis model.The THOSG code is then applied to the analysis of steady-state of IRIS's steam generator and transients of change in feedwater temperature and flow rate.Results are compared with RELAP5 system code.Comparisons indicate that THOSG code can predict the thermal-hydraulic characteristics of OTSG well.The steady-state results show that the primary side fluid is subcooled water,the fluid temperature increases gradually along the length of the tube,and the heat transfer coefficient remains basically constant.Whereas,the secondary side fluid is gradually converted from the subcooled water into superheated steam.The temperature rises rapidly along the length of the tube,and then saturated temperature is maintained until there is no liquid and the temperature continues to rise.The corresponding heat transfer coefficient first keeps the single-phase liquid heat transfer coefficient unchanged.When the fluid enters the two-phase heat transfer region,the heat transfer coefficient gradually increases until there is no liquid.Then the heat transfer coefficient drastically decreases to the single-phase steam heat transfer coefficient and keeps unchanged.Transient results show that the temperature at the outlet of the primary and secondary sides both increases when the feedwater flow rate is reduced.The more the flow rate is reduced,the greater the temperature increases.However,the steam flow rate decreases proportionally when the feedwater flow rate is reduced.For transients of change in feedwater temperature,the temperature at the outlet of the primary and secondary sides both decreases when the feedwater temperature is lowered.The more the feedwater temperature is lowered,the greater the temperature decreases.However,the steam flow does not change with the change of feedwater temperature.At last,an integral thermal-hydraulic analysis code for the primary loop of SMRs is developed.The code combines the thermal-hydraulic analysis models and codes of the main devices in SMRs' primary loop by using the multi-modules correlation and thermal parameters of the inlet and outlet coupling method.The code is then applied to analyze the thermal-hydraulic characteristics of primary loop in IRIS reactor.Results show that the coupling analysis is more accurate,compared to the thermal analysis of a single device due to considering the interaction between the devices.Studies in this paper reveal the thermal-hydraulic characteristics of the SMRs' primary loop,and provide numerical simulation tools for the thermal design and safety analysis of SMRs,which has important guiding significance for the design and development of SMRs.