TRACE code modeling of the horizontal steam generator of the PACTEL facility and calculation of a loss-of-feedwater experiment

This paper describes the modeling of horizontal steam generator with the TRACE code and calculation results of a loss-of-feedwater (LOF-10) experiment at the PACTEL facility. Parallel Channel Test Loop (PACTEL) is an integral test facility for a VVER-440 type nuclear reactor. The main objectives were to prepare a simulation model for its horizontal steam generator with the TRACE thermal hydraulic code and assess different modeling options of the code. PACTEL experiment LOF-10 was chosen for this assessment. The calculation results showed that TRACE is capable in simulating horizontal steam generator behavior both in steady state and during loss-of-feedwater transient. The phenomenon of heat transfer from primary to secondary side, steam superheating and flow reversal in the lowest heat exchange tubes were studied in detail. Different nodalization options were introduced. In the simulation of PACTEL loss-of-feedwater experiment LOF-10, the main parameters of the calculations in rather good agreement with the experiment. However, at the final state the calculated secondary side collapsed level had decreased more than in the experiment. The heat transfer from the primary to the secondary side degraded gradually during the uncovery of the heat exchange tubes. The calculations overestimated slightly this heat transfer. In the experiment the steam started to superheat immediately when the uppermost tube layer had uncovered. The steam superheating in the calculations was possible only after the uppermost cell on the secondary side had voided thoroughly. Therefore, more detailed modeling of the pipe layers increased the accuracy of the results. Especially the timing of initiation of the superheating was better estimated. The expected flow reversal in the lowest tube layers during the simulated natural circulation phase was found also in the TRACE code calculations.