Article Sidebar

Published: 2011-06-30
Abstract Views: 35
DOI: https://doi.org/10.53747/nst.v1i2.300
Issue
Vol. 1 No. 2 (2011)
Section
Articles
How to Cite
LE , D. D., HOANG , M. G., LE , V. H., LE , T. T., VO , T. H., NGUYEN , T. T. T., NGUYEN , T. T. O., & PHAM , T. N. (2011). LOSS OF COOLANT ACCIDENT ANALYSIS OF APR1400 REACTOR. Nuclear Science and Technology, 1(2), 27-36. https://doi.org/10.53747/nst.v1i2.300

LOSS OF COOLANT ACCIDENT ANALYSIS OF APR1400 REACTOR

LE DAI DIEN1, HOANG MINH GIANG1, LE VAN HONG2, LE THI THU1, VO THI HUONG1, NGUYEN THI THANH THUY1, NGUYEN THI TU OANH1, PHAM TUAN NAM1
1 Institute for Nuclear Science and Technology, 179 - Hoang Quoc Viet, Hanoi
2 VietNam Atomic Energy Institute (VAEI), 59- Ly Thuong Kiet, Hanoi

Main Article Content

Abstract

Several features of NPP with APR1400 nuclear reactor during a loss of coolant accidents (LOCAs) are investigated in this study. The paper describes some main design characteristics of the engineering safety features of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Two LOCA categories are simulated with maximum break sizes for each. The Large Break LOCA accident has been analyzed for the case of double end guilotine break and safety parameters are evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with the broken area of 0,0465m2 are derived in comparison with reference case [4]. The scenarios for the LOCA are performed for high pressure injection system (HPIS) failed to operate, 2 and 4 HPIS pumps are activated respectively. The presented results in the paper are taken from the research project between INST’s researchers and KAERI experts in using RELAP5 tool for safety analysis of NPPs under supported by Ministry of Science and Technology (MOST).

Article Details

Keywords

APR1400, Direct Vessel Injection, Safety Injection Tank, Peak Clading Temperature, Loss of Coolant Accident

References

[1] “APR1400 – Design Description. Center for Advanced Reactors Development”, KHNP, 3-2002.
[2] 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light water cooled nuclear power reactors”, January 1974, Appendix K to 10 CFR part 50 - ECCS Evaluation Models, US Code of Federal Register.
[3] Chul-Hwa Song, “Proven Technology for the APR1400’s Development, Licensing and Deployment: Focusing on Thermal-Hydraulic Safety Aspect”. IAEA INPRO Dialogue Forum Workshop, IAEA Vienna, 2- 2010.
[4] INEEL/EXT-04-02561 – “In-Vessel Retention Strategy for High Power Reactors”, Final Report, 1- 2005.
[5] KAERI/TR-2811/2004 – “MARS Code Manual- Vol II: Input Requirements”. THSR Div. KAERI, 7 - 2006.
[6] “RELAP5/MOD3 Code Manual – User’s Guide and Input Requirements”, Vol. II. INEEL, 6-1995.