As to passive nuclear power plant (NPP) transient/accident analysis, conservative codes rather than best estimated code such as RELAP5 are mainly used to do the non-LOCA analysis work. While considering the severer accident consequence, RELAP5 is selected in some cases as a tool for LOCA analysis so that safety margin can be gained and meanwhile, the safety NPP condition is guaranteed. In order to extend the RELAP5 application rang to passive NPP, a simulation capability research of RELAP5 to typical non-LOCA consisting of condition II, III and IV transient analysis of CAP1000 was conducted, and the results were compared with that analyzed by safety analysis code. Meanwhile, as to simulation skill, an innovative RELAP5 model was developed. In despite of different conservative degree of the key parameters, the response of the key equipment together with the variation tendency of the key parameters were consistent with that predicted by safety analysis code. Besides, the results met the acceptance criteria. It was showed that RELAP5 has the capability to simulate the CAP1000 non-LOCA.
Skip Nav Destination
2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5784-7
PROCEEDINGS PAPER
An Extended Application Research of RELAP5 to Typical Non-LOCA of Passive NPP
Wang Haitao,
Wang Haitao
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Liu Zhan
Liu Zhan
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Search for other works by this author on:
Wang Haitao
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Liu Zhan
Shanghai Nuclear Engineering Research and Design Institute, Shanghai, China
Paper No:
ICONE25-66487, V006T08A034; 7 pages
Published Online:
October 17, 2017
Citation
Haitao, W, & Zhan, L. "An Extended Application Research of RELAP5 to Typical Non-LOCA of Passive NPP." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 6: Thermal-Hydraulics. Shanghai, China. July 2–6, 2017. V006T08A034. ASME. https://doi.org/10.1115/ICONE25-66487
Download citation file:
19
Views
Related Proceedings Papers
Related Articles
The Fabulous Nuclear Odyssey of Belgium
J. Pressure Vessel Technol (June,2009)
Low-Power and Shut-Down Condition Medium-Break Loss-of-Coolant Accident Success Criterion Analysis for a Typical Three-Loop Nuclear Power Plant
ASME J of Nuclear Rad Sci (October,2016)
Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients
ASME J of Nuclear Rad Sci (January,2019)
Related Chapters
QRAS Approach to Phased Mission Analysis (PSAM-0444)
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)
A PSA Update to Reflect Procedural Changes (PSAM-0217)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)