Best-estimate severe accident and source term analysis for an ISLOCA scenario of a CANDU-6 plant using the MAAP-ISAAC code

Abstract An interfacing system LOCA (ISLOCA) is a bypass accident that results in the release of reactor coolant directly to the outside of a reactor building through a break in part of the low-pressure boundary of the reactor coolant system. Though the core damage frequency resulting from an ISLOCA scenario in the CANDU-6 type Wolsong nuclear power plant is very low, a direct release of radioactive nuclides to the environment could lead to severe radiation exposure among residents near the plant, and cannot be overlooked. For this reason, a quantitative evaluation of the amount of radioactive nuclides to be released to the environment was carried out by applying a best-estimate methodology to the MAAP-ISAAC code 1 , and a relevant accident management strategy for it was investigated in terms of a deterministic analysis. As part of this, an auxiliary building was additionally modeled to simulate deposition effects in the auxiliary building as well as pool scrubbing effects at the submerged break location following an ISLOCA. The decontamination effect by deposition and so on within the piping following an ISLOCA was evaluated. The analysis results showed that the decontamination factor for other nuclide groups excluding inert gas, Sb and Te2 nuclide groups was about 2.5. The analysis with consideration of all these decontamination effects was identified to yield a 78.4% reduction of Cs released to the environment compared with previous conservative analysis, which did not consider the decontamination effect within the ISLOCA piping and in the auxiliary building. In the case of the Wolsong CANDU plant, prior to external cooling water injection to the reactor vault (which is considered to be a Severe Accident Management Guidance (SAMG) accident management strategy), the amount of Cs-137 released to the environment exceeds the domestic regulatory requirement (100 TBq). In contrast, a sensitivity analysis in which an operator closes the test valve installed in a medium-pressure emergency core cooling system piping path and relieves the pressure of a reactor building using a containment filtered venting system (CFVS), met the aforementioned limitation for the amount of Cs-137 when an operator closes the test valve within 1 h following the accident and operates the CFVS properly (i.e., according to its operating procedure). This means that the appropriate accident management strategy for an ISLOCA scenario in the Wolsong CANDU plant can sufficiently prevent severe radiation exposure around the plant. Moreover, it is expected that these best-estimate methodologies developed from this study can be applied to all CANDU-6 plants.