| As the primary coolant boundary to contain the radiation from reactor as well as the heat transfer boundary from primary side to second side fluid, steam generator (SG) is one of the major devices in nuclear power plants. As the primary loop heat sink for nuclear reactor under normal operating conditions, the secure operation and efficient heat transfer of SG is critical for the the safety of nuclear power plant. As the primary pressure boundary for radioactive product containing, the integrity of SG is the key for controling radiation dosage. Therefore, the reliability of SG is of significance for the safety and economy of nuclear power plants.However, according to the analysis of accidents statistics of pressurized water reactor (PWR), the probability of steam generator tube rupture (SGTR) during the operation of a nuclear power plant is high, a quarter of unplanned outage was caused by SGTR. Thus, SG is one of the damageable equipment in nuclear reactor system. In order to improve the reliability of SG, such as structural design, operation scheme, etc., the flow and heat thransfer characteristics of SG need to be simulated and assessed under operation conditions.Because of the giant volum, complex structural, it’s difficult to simulate the flow and heat thransfer characteristics of SG under operation conditions. Furthermore, there are lots of long and narrow channels formed by heat transfer tubes lead to complicated flow and heat transfer field. Moreover, the polytropic two-phase flow and boiling heat transfer had obvious influence on second side flow field. Billions of mesh need to be divided for such complex structure, even super computer can’t satisfied the requirements.The porous-media appoach can be used to simplify the geometric model and reduce the quantity of mesh. Thus this approach provide an efficient method for simulating the flow and heat thransfer characteristics of SG.In this paper, the SG tubes was simplified as porous-media. The volume ratio between fluid and solid was reflected by porosity, and the flow resistance caused by heat transfer tubes were equivalent to distributed resistance. The two-fluid model was adopted to simulate the two-phase flow in the second side of SG, and the auxiliary model was used to describe the two-phase flow pattern, heat transfer and pressure drop. By applied these mathematical model, the thermal-hydraulic computation code THAC-SG for SG was developed. The validity of the subroutines was verified including the porous-media approach, the thermophysical property of water and steam, the numerical calculation solution. The design parameters of steam generator in Dayawan nuclear power station was used to calculate the distribution of void fraction, the variation of the coid fraction, the distribution of temperature of primary side and secondary side coulped heat transfer, the flow field, the pressure distribution and the variation of heat transfer coefficient. The results comparison with design parameter verified the validity of the code.The results shows that, the computational method of porosity based on position and the computational method of permeability based on region shortening could used to calculate the porous-media coefficients effectively. The subroutine PMA has good calculation accuracy. The results of THAC-SG shows that two-fluid model can be used to describe the two-phase flow and heat transfer behavior accurately. The thermal hydraulic characteristics of each phase can be analyze separately, which is helpful for the study of the complex flow and heat transfer in secondary side of SG. |
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