RELAP5 is a best estimate system code suitable for the analysis of all transients and postulated accidents in Light Water Reactor systems. It was usually used to solve plant thermal-hydraulic problems on system scale. RELAP5 was used to study thermal-hydraulic behavior on component level in this paper. The Test Blanket Module (TBM) was a key component in Chinese Helium-Cooled Solid Breeder (CN HCSB) system. One sub-module of TBM was simulated by RELAP5/MOD3.4. The flow paths, Be pebbles neutron multiplier as well as Li4SiO4 pebbles tritium breeder of TBM were modeled by using hydrodynamic components models as well as heat structure models provided by RELAP5. Steady-state condition was studied and the results were compared with CFD results provided by Fluent code. The steady-state results were in consistent with CFD results when the sub-module was well modeled by RELAP5. The results showed that RELAP5 could be used to solve thermal-hydraulic problems on component scale when the component was well modeled. With a detail-modeled TBM, the transient conditions of CN HCSB system could be simulated more precisely by RELAP5.
Skip Nav Destination
2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5582-9
PROCEEDINGS PAPER
Modeling TBM Components by Using RELAP5 Code Available to Purchase
Yaoli Zhang,
Yaoli Zhang
Xiamen University, Xiamen, Fujian, China
Search for other works by this author on:
Tianji Peng
Tianji Peng
Tsinghua University, Beijing, China
Search for other works by this author on:
Yaoli Zhang
Xiamen University, Xiamen, Fujian, China
Tianji Peng
Tsinghua University, Beijing, China
Paper No:
ICONE21-16304, V005T14A016; 5 pages
Published Online:
February 7, 2014
Citation
Zhang, Y, & Peng, T. "Modeling TBM Components by Using RELAP5 Code." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 5: Fuel Cycle, Radioactive Waste Management and Decommissioning; Reactor Physics and Transport Theory; Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls; Fusion Engineering. Chengdu, China. July 29–August 2, 2013. V005T14A016. ASME. https://doi.org/10.1115/ICONE21-16304
Download citation file:
9
Views
Related Proceedings Papers
Related Articles
Simulation of the Super Critical Water Loop Using ATHLET Code During an Abnormal Scenario
ASME J of Nuclear Rad Sci (April,2021)
Thermal Hydraulic Safety Assessment of LLCB Test Blanket System in ITER Using Modified relap/scdapsim/mod4.0 Code
ASME J of Nuclear Rad Sci (April,2018)
Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor
J. Eng. Gas Turbines Power (October,2010)
Related Chapters
New Generation Reactors
Energy and Power Generation Handbook: Established and Emerging Technologies
Studies Performed
Closed-Cycle Gas Turbines: Operating Experience and Future Potential
Fissioning, Heat Generation and Transfer, and Burnup
Fundamentals of Nuclear Fuel