Calculations for AP1000 nuclear power plant SGTR accident based on RELAP5/mod 3.3 program
AP1000 nuclear power plant model is founded via RELAP5/mod 3.3 program in detail. Refer to SGTR accident basic assumptions in the safety analysis report, calculations for AP1000 nuclear power plant SGTR accident are carried out. And through comparison between the result got and the result from safety analysis report calculated by LOFTTR2 program, the two results are very consistent. It's found that even if the operators do not interfere, AP1000 nuclear power plant protection systems and passive design measures will trigger an automatic response measures, it can terminate the SG heat transfer tube leaking and make RCS be stable in a safe condition and to prevent SG overflowing and ADS actions occurring.
Frequency and consequences associated with a steam generator tube rupture event
The results of a study into the frequency of steam generator tube rupture (SGTR) events are presented, including estimates on the upper and lower bound frequencies for U-tube and once-through steam generator plants and single- and multiple-tube ruptures. In addition, commercial pressurized water reactor operational data have been researched and iodine spiking data used to develop data bases of maximum resultant iodine concentrations and release rates. The frequencies and iodine spiking magnitudes are compared with other studies, and conclusions are drawn regarding current guidelines for analysis of this design-basis transient.
Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors
A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113/sup 0/F) for TMI-1 and approx.44 K (approx.80/sup 0/F) for Zion-1.