Study on Thermal-hydraulic Steady-state Comprehensive Performance of Fast Reactor Sodium-water Steam Generator

Abstract

The sodium-heated once-through steam generator (OTSG) plays an important role in separating the second and third circuits of sodium cooled fast reactor (SFR). Once the heat transfer pipe breaks, it will cause serious sodium water reaction, which will seriously affect the availability, economy and reliability of nuclear power plant operation. In order to analyze the flow heat transfer characteristics of the steam generator, based on the China Demonstration Fast Reactor (CFR600) once-through steam generator designed by the China Institute of Atomic Energy, Xi'an Jiaotong University built the PUSA test platform for the comprehensive performance of the fast reactor sodium-water steam generator according to the equal height, equal pressure and equal heat exchange tube diameter, and carried out a series of steady-state and transient comprehensive performance experiments. At the same time, the one-dimensional two-phase thermal hydraulic design code was developed independently. By comparing the experimental data with the calculation results of the steam generator design and the verification analysis code DeCOSS, this paper aimed to verify the rationality of the steam generator design and the accuracy of the steady-state calculation of the design analysis code. Firstly, the self-developed fast reactor sodium-water steam generator two-phase flow thermal fluid design and verification analysis code DeCOSS was used. Five typical working conditions of low power (heating power 7.18% and 12.41% rated power), medium power (heating power 27.87% and 29.74% rated power) and full power (heating power 100% rated power) were selected from the steady-state experiment to carry out steady-state calculation of DeCOSS code. Then, the above thermal hydraulic parameters such as temperature and flow rate obtained under the five stable working conditions of low power, medium power and full power were divided by the temperature and flow rate under full power respectively. After normalization, the DeCOSS code was compared and verified with the axial temperature distribution on the sodium side and the temperature at the exit of the sodium side and the water side of the steam generator. It is found that under different power conditions, the calculated results of the DeCOSS code are in good agreement with the experimental data. The error between the exit temperature of the sodium side and the water side calculated by the DeCOSS code and the experimental data is within 5%, which verifies the correctness of steam generator autonomous DeCOSS code. In addition, the obtained experimental data can also be used to modify the calculation method of the DeCOSS code, which will lay a foundation for the research of the sodium-heated OTSG in the future.

Keywords

• sodium-water steam generator test facility
• decoss code
• thermal-hydraulic analysis