As one of the generation IV reactors, pool-type Sodium-cooled Fast Reactors (SFRs) is attracting more and more attention. Loss of flow and heat sink accident is one of the most serious accidents for SFRs. Therefore, the decay heat removal capacity after emergency shutdown is of great importance. This paper has developed a one-dimensional code named Decay heat Removal Analysis Code for Sodium-cooled Fast Reactor (DRAC-SFR) to analyze the decay heat removal capacity. In the code, the decay heat removal system contains the primary loop, the intermediate loop and air circuit. The decay heat is removed out step by step with the above three loops. Many studies have been conducted on code verification. The international benchmark analysis of EBRII reactor is applied in the code verification. The calculation is compared with the experimental data and the results of DRAC-SFR agreed well with the experimental data. The comparison with the steady state of China Experimental Fast Reactor (CEFR) shows a good agreement with the design value. The errors of all the compared parameters are within 2%. What’s more, calculation is performed to analyze the characteristics of the decay heat removal capacity for CEFR. Thus, code verification shows that DRAC-SFR is proper to evaluate the decay heat removal capacity for SFRs and has the ability to provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.
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2018 26th International Conference on Nuclear Engineering
July 22–26, 2018
London, England
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5148-7
PROCEEDINGS PAPER
Development and Basic Verification of Decay Heat Removal Analysis Code of Sodium-Cooled Fast Reactor
Ping Song,
Ping Song
Xi'an Jiaotong University, Xi'an, China
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Dalin Zhang,
Dalin Zhang
Xi'an Jiaotong University, Xi'an, China
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Tangtao Feng,
Tangtao Feng
Xi'an Jiaotong University, Xi'an, China
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Shibao Wang,
Shibao Wang
Xi'an Jiaotong University, Xi'an, China
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Yapei Zhang,
Yapei Zhang
Xi'an Jiaotong University, Xi'an, China
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Mingjun Wang,
Mingjun Wang
Xi'an Jiaotong University, Xi'an, China
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Suizheng Qiu,
Suizheng Qiu
Xi'an Jiaotong University, Xi'an, China
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G. H. Su
G. H. Su
Xi'an Jiaotong University, Xi'an, China
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Ping Song
Xi'an Jiaotong University, Xi'an, China
Dalin Zhang
Xi'an Jiaotong University, Xi'an, China
Tangtao Feng
Xi'an Jiaotong University, Xi'an, China
Shibao Wang
Xi'an Jiaotong University, Xi'an, China
Yapei Zhang
Xi'an Jiaotong University, Xi'an, China
Mingjun Wang
Xi'an Jiaotong University, Xi'an, China
Suizheng Qiu
Xi'an Jiaotong University, Xi'an, China
G. H. Su
Xi'an Jiaotong University, Xi'an, China
Paper No:
ICONE26-81630, V06AT08A061; 7 pages
Published Online:
October 24, 2018
Citation
Song, P, Zhang, D, Feng, T, Wang, S, Zhang, Y, Wang, M, Qiu, S, & Su, GH. "Development and Basic Verification of Decay Heat Removal Analysis Code of Sodium-Cooled Fast Reactor." Proceedings of the 2018 26th International Conference on Nuclear Engineering. Volume 6A: Thermal-Hydraulics and Safety Analyses. London, England. July 22–26, 2018. V06AT08A061. ASME. https://doi.org/10.1115/ICONE26-81630
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