Article Sidebar

Published: 2013-06-30
Abstract Views: 49
DOI: https://doi.org/10.53747/nst.v3i2.255
Issue
Vol. 3 No. 2 (2013)
Section
Articles
How to Cite
Le , D. D., & Le , T. D. (2013). Analysis of steam generator tube rupture accident for Korean reactor APR1400. Nuclear Science and Technology, 3(2), 7-14. https://doi.org/10.53747/nst.v3i2.255

Analysis of steam generator tube rupture accident for Korean reactor APR1400

Le Dai Dien1, Le Tri Dan1
1 Nuclear Safety Center - Institute for Nuclear Science and Technology 179 Hoang Quoc Viet, Cau Giay, Ha Noi

Main Article Content

Abstract

The APR1400 is an advanced light water reactor designed by KEPCO, Korea. Steam generator tube rupture (SGTR) is one of the design basic accidents (DBA) which needs to be analyzed in safety analysis report.


In this paper, a steam generator tube rupture event was simulated by using RELAP/SCDASIM system code and some results are compared with MARS code with modeled MULTID components of the reactor vessel and two steam generators. Further simulations for tube rupture depend on the broken positions and number of ruptured tubes is also estimated. The results are in fairly good agreement with one simulated by MARS code.

Article Details

Keywords

APR1400, steam generator tube rupture, safety analysis report, direct vessel injection, reactor trip, reactor coolant pump trip, nodalization

References

[1] C. M. Allison and R.J. Wagner, RELAP/SCDAPSIM/MOD3.2 (am+) Input Manual, Supplemental, Innovative Systems Software, LLC, Dec. 2001.
[2] Le Van Hong et al. Summary Report on Cooperation project between KAERI and VAEI: Study on Safety Analysis of PWR reactor core in Transient and Accident Conditions (2009-2010).
[3] Le Đai Dien, Hoang Minh Giang, Le Van Hong, Le Thi Thu, Nguyen Thi Tu Oanh, Vo Thi Huong, Phạm Tuan Nam, Nguyen Thanh Thuy, Some preliminary results of LOCA problem for Korean Reactor APR1400. The Ninth National Conference on Nuclear Science and Technology, Ninh Thuan, 18-19 August 2011.
[4] Le Đai Dien et al. Loss of Coolant Accident Analysis of APR1400 Reactor. Nuclear Science and Engineering, No.2 June 2011.
[5] Nguyen Thi Thanh Thuy, Le Dai Dien and Hoang minh Giang. Feed Water Line Break (FWLB) Analysis for APR1400 Reactor Using RELAP5. The Ninth National Conference on Nuclear Science and Technology, Ninh Thuan, 18-19 August 2011.
[6] APR1400 SSAR. Chapter 5: Reactor Coolant System and Connected Systems.
[7] Chung B.D., et al, 2005, “Development and assessment of multi-dimensional flow models in the thermal-hydraulic system analysis code MARS,” KAERI/TR-3011/2005, KAERI.
[8] MARS CODE MANUAL, VOLUME IV: Developmental Assessment Report. KAERI/TR-3042/2005.

[9] Ji Hwan Jeong et al. Best Estimate Analysis of MSTGR event in APR1400 Aiming to Examine the Effect of Affected Steam.
[10] Generator Selection. Journal of KNS, Vol.33, No.4, August 2002.
[11] Ji Hwan Jeong et al. The effect of tube rupture location on the consequences of multiple steam generator tube rupture event. Annals of Nuclear Energy 29 (2002).
[12] Ji Hwan Jeong et al. Effects of tube rupture modeling and the parameters on the analysis of multiple steam generator tube rupture event progression in APR1400. Nuclear Engineering and Design 224 (2003).
[13] A. Auvinen et al. Steam generator tube ruptures (SGTR) scenarios. Nuclear Engineering and Design 235 (2005).