Following China’s road map of nuclear technology development, the development of self-reliant nuclear design codes becomes one of the most significant steps in the plan. Among the nuclear design codes, thermal-hydraulic analysis code is indispensable because it is the foundation of reactor safety analysis and reactor design. Recently, China Guangdong Nuclear Power Group has launched a series of R&D projects of reactor design code development. The sub-channel analysis code-LINDEN becomes one of the key subprojects. Since the sub-channel code is developed for thermal-hydraulic design and safety analysis of pressurized water reactors (PWRs), the basic requirements for the LINDEN code are reliability and stability. Therefore, the mathematical model and numerical method developed in the code are based on the matured approaches that have been used in various industrial applications. These models and methods includes: four-equation drift framework model of two-phase flow; the classical heat transfer model and fuel rod model (Poisson equation) as well as the constitutive relations; explicit formulation and stepping algorithms for equation solutions. The solver module of the code is programmed using object-oriented C/C++ language with the structural design.. With all these features, the code was developed to be stable, scalable and compatible. The code’s applicability has been further improved after model improvement and design optimization according to characteristics of China’s proprietary type of reactor. COBRA-IV and LINDEN have been used to conduct the thermal-hydraulics analysis for the Daya bay unit 1 and 2 nuclear plants at the steady-state conditions. The results demonstrate that the two codes agree well with each other. The preliminary tests show that the LINDEN code should be suitable for thermal-hydraulics analysis of large PWRs.

The CHF in PWR fuel assemblies is usually predicted by the local flow correlation approach based on subchannel analysis while the effects of spacer grids, cold walls, non-uniform heat flux, etc are investigated. By using the subchannel code ATHAS to calculate each set of bundle CHF data, the local thermal-hydraulic parameters at DNB occurrence point were obtained. In present study, the minimum DNBR point method was applied to develop a new CHF correlation for PWR fuel assemblies. The so-called “three-step method” and “magnitude analysis method” were used to determine the shape and the expression of each item, respectively and the least square method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, the CHF correlation named ACC correlation has been developed to calculate the risk of DNB. The analysis and assessment results indicate that the ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Coupled with subchannel code ATHAS, this correlation can simulate the thermal-hydraulics performances of PWR fuel assemblies exactly.