Application of the W̱COBRA/TRAC best-estimate methodology to the AP600 large-break LOCA analysis

• Published in: Nuclear engineering and design Vol. 186; no. 1; pp. 279 - 301
• Main Authors: Zhang, Jinzhao, Bajorek, S.M., Kemper, R.M., Nissley, M.E., Petkov, N., Hochreiter, L.E.
• Format: Journal Article
• Language: English
• Published: Elsevier B.V 01.11.1998
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/S0029-5493(98)00279-9

The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently United States Nuclear Regulatory Commission (USNRC)-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the W̱COBRA/TRAC code to model the AP600 unique features was validated against cylindrical core test facility (CCTF) and upper plenum test facility (UPTF) downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA conditions and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models, as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95th percentile peak cladding temperature (PCT 95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA.