Lower Core Support Plate (LCSP) and Core Barrel (CB) are key components of reactor vessel internals. Especially, since the fuel assemblies are installed on the LCSP, its flatness is critical for the safe operation of fuel assemblies. However, for SM1 and HY1 nuclear power plant (NPP), after heat treatment of the weld between LCSP and CB, the LCSP deforms seriously and its flatness exceeds the limitation, which results in a time-consuming and costly reprocessing. A numerical model of heat treatment process between LCSP and CB was developed first. The general rules of temperature and deformation distribution of LCSP and CB were obtained. Also, an experiment was conducted to validate the model. With the validated model, the deformation mechanism of LCSP due to heat treatment is studied. At last, the heat treatment process between LCSP and CB was optimized to avoid similar issues for the following NPPs.
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2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5780-9
PROCEEDINGS PAPER
Analysis and Optimization of Heat Treatment Process for Lower Core Support Plate and Core Barrel of Reactor Vessel Internals
Dongan Liu,
Dongan Liu
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
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Shaoxuan Lin,
Shaoxuan Lin
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
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Zonghua Ding
Zonghua Ding
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Dongan Liu
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Shaoxuan Lin
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Zonghua Ding
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Paper No:
ICONE25-66678, V002T03A059; 4 pages
Published Online:
October 17, 2017
Citation
Liu, D, Lin, S, & Ding, Z. "Analysis and Optimization of Heat Treatment Process for Lower Core Support Plate and Core Barrel of Reactor Vessel Internals." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 2: Plant Systems, Structures, Components and Materials. Shanghai, China. July 2–6, 2017. V002T03A059. ASME. https://doi.org/10.1115/ICONE25-66678
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