Communication independence is one of the key criteria of digital safety I&C system design. This paper mainly analyzes the requirements for communication independence in safety regulations and standards, and then introduces the architecture and design features, including communication failure processing measures, of communication networks of ACPR1000 nuclear power plant safety digital protection system based on FirmSys platform developed by CTEC. The communication design meets the regulations requirements and effectively improves the safety and reliability of the system, and it is successfully applied in reactor protection system (RPS) of Yang Jiang nuclear power plant unit 5&6. In addition this design can provide reference for communication designs of other NPPs and industries.
Skip Nav Destination
2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5779-3
PROCEEDINGS PAPER
Design of the Communication Independence for ACPR1000 Nuclear Power Plant Digital Safety I&C System
Shi Gui-lian,
Shi Gui-lian
China Techenergy Co., Ltd, BeiJing, China
Search for other works by this author on:
Xie Yi-qin,
Xie Yi-qin
China Techenergy Co., Ltd, BeiJing, China
Search for other works by this author on:
Li Gang,
Li Gang
China Techenergy Co., Ltd, BeiJing, China
Search for other works by this author on:
Jiang Guo-jin
Jiang Guo-jin
China Techenergy Co., Ltd, BeiJing, China
Search for other works by this author on:
Sun Na
China Techenergy Co., Ltd, BeiJing, China
Shi Gui-lian
China Techenergy Co., Ltd, BeiJing, China
Xie Yi-qin
China Techenergy Co., Ltd, BeiJing, China
Li Gang
China Techenergy Co., Ltd, BeiJing, China
Jiang Guo-jin
China Techenergy Co., Ltd, BeiJing, China
Paper No:
ICONE25-67880, V001T04A050; 5 pages
Published Online:
October 17, 2017
Citation
Na, S, Gui-lian, S, Yi-qin, X, Gang, L, & Guo-jin, J. "Design of the Communication Independence for ACPR1000 Nuclear Power Plant Digital Safety I&C System." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Shanghai, China. July 2–6, 2017. V001T04A050. ASME. https://doi.org/10.1115/ICONE25-67880
Download citation file:
17
Views
Related Proceedings Papers
Related Articles
The Fabulous Nuclear Odyssey of Belgium
J. Pressure Vessel Technol (June,2009)
Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel
J. Pressure Vessel Technol (June,2016)
Performance-Based Reliability of ASME Piping Design Equations
J. Pressure Vessel Technol (June,2017)
Related Chapters
A PSA Update to Reflect Procedural Changes (PSAM-0217)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Application of Probabilistic Methods for the Evaluation of Deterministic Deviations from Technical Specifications (PSAM-0277)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Seismic Protection for Pressure Piping Systems
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards