Semiscale steam line break transient test prediction with the RELAP5/MOD2 code

Although ruptures of steam generator main steam lines are not expected to occur often in pressurized water reactor (PWR) plants, the potential consequences of these events necessitate their examination. Steam line break transients can lead to overcooling and possibly repressurization of the primary coolant system. This phenomenon, termed pressurized thermal shock, poses a threat to the integrity of the PWR pressure vessel. This paper presents the analysis of test data and comparisons with the pretest calculation results for the first Semiscale steam line break test (S-FS-2) performed in the Semiscale Mod-2C facility. Most of the primary and secondary responses including the overcooling and depressurization of the primary system were reasonably well predicted by RELAP5/MOD2 although a difference was noted in the primary coolant temperature.