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 Principle and Preliminary Validation of a New Subchannel Code
- Views Icon Views
- Share Icon Share
- Search Site
Bai, N, Zhu, Y, Ren, Z, He, H, Lu, H, Chen, J, Li, W, & Li, J. "The Principle and Preliminary Validation of a New Subchannel Code." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 4: Thermal Hydraulics. Chengdu, China. July 29–August 2, 2013. V004T09A038. ASME. https://doi.org/10.1115/ICONE21-15464
Download citation file: