International Science Index


Effect of Modeling of Hydraulic Form Loss Coefficient to Break on Emergency Core Coolant Bypass


Emergency Core Coolant Bypass (ECC Bypass) has been regarded as an important phenomenon to peak cladding temperature of large-break loss-of-coolant-accidents (LBLOCA) in nuclear power plants (NPP). A modeling scheme to address the ECC Bypass phenomena and the calculation of LBLOCA using that scheme are discussed in the present paper. A hydraulic form loss coefficient (HFLC) from the reactor vessel downcomer to the broken cold leg is predicted by the computational fluid dynamics (CFD) code with a variation of the void fraction incoming from the downcomer. The maximum, mean, and minimum values of FLC are derived from the CFD results and are incorporated into the LBLOCA calculation using a system thermal-hydraulic code, MARS-KS. As a relevant parameter addressing the ECC Bypass phenomena, the FLC to the break and its range are proposed.

[1] United States Nuclear Regulatory Commission, “Compendium of ECCS Research for Realistic LOCA Analysis,” Final Report, NUREG-1230, 1988, pp 6.3-1-33.
[2] P. S. Damerell, J. W. Simons, “Reactor Safety Issues Resolved by the 2D/3D Program,” International Agreement Report, NUREG/IA-0127, 1993, pp 4.1-4.33.
[3] Korea Hydro and Nuclear Co. and Korea Electric Power Co.-Nuclear Fuel, “LBLOCA Best-Estimate Evaluation Model for APR1400 Type Nuclear Power Plants Using the SPACE Code,” Topical Report TR-KHNP-0030, August 2017, pp. 6-7.
[4] Korea Institute of Nuclear Safety, “MARS Code Manaual - Volume II: Input Requirements,” KINS/RR-1282, Rev.2, May 2016.
[5] Ansys Inc., ANSYS CFX-18, 2017.
[6] Wadle, M. A., “A new formula for the pressure recovery in an abrupt diffuser,” International Journal of Multiphase Flow, Vol. 15, No. 2, 1989, pp. 241~256.
[7] Schmidt, J. and Friedel, L., “Two-phase pressure drop across sudden contractions in duct areas,” International Journal of Multiphase Flow, Vol. 23, No. 2, 1997, pp. 283~299.
[8] Roul, M. K. and Dash, S. K., “Two-phase pressure drop caused by sudden flow area contraction/expansion in small circular pipes,” Int. J. Numer. Mech. Fluids, Vol. 66, 2011, pp. 1420~1446
[9] D. H. Yoon, Y. S. Bang, “CFD Analysis for Two-phase Flow Pressure Drop through Broken Cold Leg in Reactor Downcomer,” Trans. KSME 2017 Autumn Meeting, 2017, pp. 2501-2505.
[10] Information Systems Laboratories, Inc. RELAP5/MOD3.3 Code Manual Volume I: Code Structure, System Models, and Solution Methods, NUREG/CR-5535/Rev P5-Vol I, 2016.
[11] Korea Hydro and Nuclear Co. “Shinkori Units 3 and 4, Final Safety Analysis Report,” 2015, pp. 6.3-1~89.
[12] R. E. Henry and H. K. Fauske. “The Two-Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes.” Transactions of ASME, Journal of Heat Transfer. 93. 1971, pp. 179-187.