Abstract
Blankets are an important component of fusion reactors and the key object of thermal design and safety analysis. The helium-cooled solid breeder (HCSB) blanket is one of the three candidate tritium breeding blanket concepts for the Chinese Fusion Engineering Test Reactor (CFETR). Fusion reactors’ thermal design and safety analysis are mostly based on fission reactor procedures worldwide. In this paper, based on the system safety analysis module cosSyst in the COSINE code package of China’s independently developed system safety analysis code for pressurized water reactors, functional development was carried out for the helium-cooled blanket of fusion reactor, and helium physical property envelope and corresponding heat transfer model were added. The modified code was used to analyze three typical accidents with the blanket and the combined helium cooling system (HCS): In-Vessel LOCA, In-Box LOCA, and Ex-Vessel LOCA. The research findings indicate that under the three accident scenarios, the first wall and the cooling loop system maintained their structural integrity, with no melting or exceeding pressure limits.
Funding source: The National Magnetic Confinement Fusion Program of China
Award Identifier / Grant number: Grant Nos.2019YFE03110000 and 2019YFE03110001.
Acknowledgments
This work is supported by the National Magnetic Confinement Fusion Program of China under Grant Nos.2019YFE03110000 and 2019YFE03110001.
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Research ethics: Not applicable.
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Informed consent: Not applicable.
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Author contributions: The authors have accepted responsibility for the entire content of this manuscript and approved its submission.
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Use of Large Language Models, AI and Machine Learning Tools: None declared.
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Conflict of interest: The authors state no conflict of interest.
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Research funding: The National Magnetic Confinement Fusion Program of China. Grant Nos.2019YFE03110000 and 2019YFE03110001.
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Data availability: The raw data can be obtained on request from the corresponding author.
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- Calendar of events
Articles in the same Issue
- Frontmatter
- Investigations into the development of a new type of internal pipe cutting device for difficult to access pipelines
- Study on flow field characteristics of regulator in uranium enrichment centrifugal cascade
- Innovative materials for enhancing safety, efficiency, and sustainability in nuclear waste management
- Double solitary waves reactor
- Advancing the safety design of heat pipe cooled reactors: a case study of lead liquid bath and monolith stainless steel core
- Evaluation of various calculational models of FA containing burnable absorber rod in the VVER-1000
- Accurate departure from nucleate boiling ratio (DNBR) prediction using SIMCA’s partial least squares regression and clustering
- Effect of glass cooling method on thermal shock behavior of nuclear waste container
- Study on the impact of containment mesh refinement and PAR installation on hydrogen distribution during severe accidents
- Preliminary analysis of typical accidents of CFETR helium-cooled solid breeder blanket system based on COSINE
- Calendar of events