Browse > Article
http://dx.doi.org/10.3795/KSME-A.2014.38.4.355

Evaluation of Pressure History due to Steam Explosion  

Kim, Seung Hyun (Dept. of Nuclear Engineering, Kyung Hee Univ.)
Chang, Yoon-Suk (Dept. of Nuclear Engineering, Kyung Hee Univ.)
Song, Sungchu (Korea Institute of Nuclear Safety)
Hwang, Taesuk (Korea Institute of Nuclear Safety)
Publication Information
Transactions of the Korean Society of Mechanical Engineers A / v.38, no.4, 2014 , pp. 355-361 More about this Journal
Abstract
Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.
Keywords
External Reactor Vessel Cooling; In-vessel Retention; Severe Accident; Steam Explosion;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
연도 인용수 순위
1 Kim, H. D., Kim, D. H., Kim, J. T., Kim, S. B., Song, J. H. and Hong, S. W., 2009, "Investigation on the Resolution of Severe Accident Issues for Korean Nuclear Power Plants," Nuclear Engineering and Technology, Vol. 41, pp. 617-648.   DOI   ScienceOn
2 Flectcher, D. F., 1995, "Validation of the CHYMES Mixing Model," Nuclear Engineering and Design, Vol. 155, pp.85-96.   DOI   ScienceOn
3 Liu, J. and Koshizuka, S., 2002, "Propagation Investigations Using the CULDESAC Model," Nuclear Engineering and Design, Vol. 216, pp. 121-137.   DOI   ScienceOn
4 Fletcher, D. F., 1992, "A Comparison of Mixing Predictions Obtained from the CHYMES and PMALPHA Models," Nuclear Engineering and Design, Vol. 135, pp. 419-425.   DOI   ScienceOn
5 Sandia National Lab., 1999, "IFCI 7.0 Models and Correlations," SAND99-1000.
6 Corradini, M. L., Murphy, J. and Nilsuwankosit, S., 2002, "User's Manual for TEXAS-V one Dimensional Transient Fluid Model," University of Wisconsin.
7 Kim, B. J., 1986, "Overview of Steam Explosion," Trans. Korean Soc. Mech. Eng., Vol. 28, pp. 270-280.
8 JAERI, 1997, "Proceedings of the OECD/CSNI Specialist Meeting on Fuel-Coolant Interaction," NEA/CSNI/R.
9 OECD/NEA, 2007, "OECD Research Programme on Fuel-Coolant Interaction: SERENA Final Report," NEA/CSNI /R.
10 Park, I. K., Kim, J. H. and Min, B. T., 2009, "An Evaluation of the Ex-Vessel Steam Explosion Load Against TROI Experimental Result," Nuclear Engineering and Design, Vol. 33, pp. 622-628.
11 Cizelj, L., Koncar, B. and Leskovar, M., 2006, "Vulnerability of a Partially Flooded PWR Reactor Cavity to a Steam Explosion," Nuclear Engineering and Design, Vol. 236, pp. 1617-1627.   DOI   ScienceOn
12 ANASYS CFX, 2012, "Introduction of CFX Ver. 14.0," ANASYS Inc.
13 Huhtiniemi, I., Magallon, D. and Hohmannn, H., "Results of Recent KROTOS FCI Test: Alumina Versus Corium Melts," 1999, Nuclear Engineering and Design, Vol. 189, pp. 379-389.   DOI   ScienceOn
14 Sehgal, B. R., Theerthan, A., Giri, A., Karbojian, A., Willschutz, H. G., Kymalainen, O., Vandroux, S., Bonnet, J. M., Seiler, J. M., Ikkonen, K., Sairanen, R., Bhandari, S., Burger, M., Buck, M., Widmann, M., Dienstbier, J., Techy, Z., Kostka, P., Taubner, R., Theofanous, T. and Dinh, T. N., 2005, "Assessment of Reactor Integrity," Nuclear Engineering and Design, Vol. 235, pp. 213-232.   DOI   ScienceOn
15 OECD/CSNI/NEA, 2002, "OECD Lower Head Failure Project Final Report ," NEA/CSNI/R.