Abstract: Analysis of a small break loss of coolant accident（SB-LOCA） for the latest design of ACR-700 PHWR nuclear power plant（NPP） developed by Atomic Energy of Canada Limited has been performed with CATHENA MOD 3.5d,a PHWR system thermal-hydraulic analysis code based on the analysismodel established.The limiting break size has been found by performing the sensitivity study for three different break locations [i.e.reactor inlet head（RIH）,HTS pump suction（PS） pipe and reactor outlet head（ROH）] under the limiting case（i.e.SB-LOCA with subsequent loss of class Ⅵ power with all safety system available）,and main analysis results were also provided.