To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.
Skip Nav Destination
2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5783-0
PROCEEDINGS PAPER
A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV Available to Purchase
Eltayeb Yousif,
Eltayeb Yousif
Harbin Engineering University, Harbin, China
Search for other works by this author on:
Zhang Zhijian,
Zhang Zhijian
Harbin Engineering University, Harbin, China
Search for other works by this author on:
Tian Zhao-fei,
Tian Zhao-fei
Harbin Engineering University, Harbin, China
Search for other works by this author on:
A. M. Mustafa
A. M. Mustafa
Harbin Engineering University, Harbin, China
Search for other works by this author on:
Eltayeb Yousif
Harbin Engineering University, Harbin, China
Zhang Zhijian
Harbin Engineering University, Harbin, China
Tian Zhao-fei
Harbin Engineering University, Harbin, China
A. M. Mustafa
Harbin Engineering University, Harbin, China
Paper No:
ICONE25-67469, V005T05A043; 12 pages
Published Online:
October 17, 2017
Citation
Yousif, E, Zhijian, Z, Zhao-fei, T, & Mustafa, AM. "A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 5: Advanced and Next Generation Reactors, Fusion Technology; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues. Shanghai, China. July 2–6, 2017. V005T05A043. ASME. https://doi.org/10.1115/ICONE25-67469
Download citation file:
31
Views
Related Proceedings Papers
Related Articles
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)
Methodology for Calculating Minor Radioactive Releases From VVER 1000 Using TRACE Code
ASME J of Nuclear Rad Sci (April,2021)
Assessment of the Predictive Capability of VERA—CS for CASL Challenge Problems
J. Verif. Valid. Uncert (June,2021)
Related Chapters
Design of Indian Pressurized Heavy Water Reactors
Global Applications of the ASME Boiler & Pressure Vessel Code
LOCA Frequencies Estimated from Operating Experience (PSAM-0282)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
QRAS Approach to Phased Mission Analysis (PSAM-0444)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)