基于两步法的堆芯物理-热工耦合系统的开发与验证

With the increasing requirements for the accuracy of numerical calculations in reactors, 0more and more attention has been paid to multi-physics coupling calculation in nuclear reactor analysis. [Purpose]: As one of the main calculation methods in industry, the study of the neutronics and thermal-hydraulics coupling calculation method which is suitable for the deterministic two-step method has a clear value for industry. [Methods]: Based on the two-step codes DRAGON/DONJON and the subchannel code COBRA-EN, a neutronics and thermal-hydraulics coupling code system based on a unified framework has been developed, and the coupling code system has been validated around the Virtual Environment for Reactor Applications (VERA) series of benchmark problems 6 and 7, which were proposed by the Consortium for Advanced Simulation of LWRs (CASL), in the United States. [Results]: The results show that the error of keff for Problem 6 is within 10010-5, and the radial fission rate of the component is within 1%, the distribution trends of fuel temperature and coolant temperature agree well with reference values; While the computational error of the critical boron concentration for Problem 7 is within 2010-6, and the root-mean-square error of the power distribution is 0.86%, the coolant temperature at the core outlet differs from the reference value within 278.15K .[Conclusions]: The computational capability of the coupling code system is verified.