In view of control rod ejection accident of the traditional pressurized water reactor, the safety thought of the design phase is to validate reliability and availability of DCS I&C in the severe accidents. Now the most important and effective means is simulation calculation and analysis. It is applied for the imaginary accident of the nuclear power plant by using computer software. The new safety analysis steps based on the analysis of cause-and-effect logic failure: firstly, the composition and working principle of control rod drive mechanism is analyzed; secondly, a list of factors-the dynamics and structure, environmental reasons, the function of the control rod drive mechanism and status analysis-are all taken into account, the initial cause of failure modes with causal logic analysis is carried out; thirdly, based on cause-and-effect logic failure, the prevention and improvement measures of accidents, the new criterion of design are put forward. The advantages of cause-and-effect logic failure safety analysis: 1.be based on causal logic. 2. the system aspects is added compared with the past method that is only based on simulation calculation and analysis of the hypothetical accident, the accident the transient process of the key security parameters as the acceptance criteria. 3. The verification and audit of the lack of safety design criteria, completeness of design content, sufficiency problem are performed before the simulated calculation and analysis. 4. The coverage of safety analysis is expanded. Some good advices are provided for the design, operation and maintenance of nuclear power plant.
Skip Nav Destination
2018 26th International Conference on Nuclear Engineering
July 22–26, 2018
London, England
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5146-3
PROCEEDINGS PAPER
A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship
Shiyu Yan,
Shiyu Yan
University of South China, Heng Yang City, China
Search for other works by this author on:
Hua Liu,
Hua Liu
University of South China, Heng Yang City, China
Search for other works by this author on:
Zhaohui Liu,
Zhaohui Liu
University of South China, Heng Yang City, China
Search for other works by this author on:
Xiaohua Yang,
Xiaohua Yang
University of South China, Heng Yang City, China
Search for other works by this author on:
Meng Li,
Meng Li
University of South China, Heng Yang City, China
Search for other works by this author on:
Zhi Chen
Zhi Chen
Nuclear Power Institute of China, Cheng Du City, China
Search for other works by this author on:
Shiyu Yan
University of South China, Heng Yang City, China
Hua Liu
University of South China, Heng Yang City, China
Zhaohui Liu
University of South China, Heng Yang City, China
Xiaohua Yang
University of South China, Heng Yang City, China
Meng Li
University of South China, Heng Yang City, China
Zhi Chen
Nuclear Power Institute of China, Cheng Du City, China
Paper No:
ICONE26-81879, V004T06A027; 7 pages
Published Online:
October 24, 2018
Citation
Yan, S, Liu, H, Liu, Z, Yang, X, Li, M, & Chen, Z. "A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship." Proceedings of the 2018 26th International Conference on Nuclear Engineering. Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation. London, England. July 22–26, 2018. V004T06A027. ASME. https://doi.org/10.1115/ICONE26-81879
Download citation file:
20
Views
Related Proceedings Papers
Related Articles
Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor
ASME J of Nuclear Rad Sci (October,2016)
Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel
J. Pressure Vessel Technol (June,2016)
Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria
ASME J of Nuclear Rad Sci (April,2017)
Related Chapters
A PSA Update to Reflect Procedural Changes (PSAM-0217)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
On the Exact Analysis of Non-Coherent Fault Trees: The ASTRA Package (PSAM-0285)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)