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- [7] Development of Thermal-hydraulic Analysis Code for Spent Fuel Assembly of Sodium Cooled Fast Reactor Yuanzineng Kexue Jishu/Atomic Energy Science and Technology, 2020, 54 (04): : 606 - 614
- [8] Development and validation of a thermal-hydraulic analysis code for dual cooled assemblies in molten salt reactors He Jishu/Nuclear Techniques, 2024, 47 (09):
- [9] DEVELOPMENT OF THERMAL-HYDRAULIC AND SAFETY ANALYSIS CODE FOR A HEAT PIPE COOLED REACTOR PROCEEDINGS OF 2024 31ST INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 4, ICONE31 2024, 2024,