Test results of Run-1 and Run-2 in Steam Generator Safety Test Facility (SWAT-3)

Kurihara, Akikazu ; Yatabe, Toshio; Hiroi, Hiroshi; Tanabe, Hiromi

Large leak sodium-water reaction tests were carried out using SWAT-1 rig and SWAT-3 facility in Power Reactor and Nuclear Fuel Development Corporation (PNC) O-arai Engineering Center to obtain the data on the design of the prototype LMFBR Monju steam generator against a large leak accident.This report provides the results of SWAT-3 Runs 1 and 2.In Runs 1 and 2, the heat transfer tube bundle of the evaporator, fabricated by TOSHIBA/IHI, were used, and the pressure relief line was located at the top of evaporator.The water injection rates in the evaporator were 6.7kg/s and 14.2 (initial) - 9.7kg/s in Runs 1 and 2 respectively, which corresponded to 3.3 tubes and 7.1 (initial) - 4.8 tubes failure in actual size system according to iso-velocity modeling.Approximately two hundreds of measurement points were provided to collect data such as pressure,Temperature, strain,sodium level, void, thrust load, acceleration, displacement, flow rate, and so on in each run.Initial spike pressures were 1.13MPa and 2.62MPa nearest to injection point in Runs 1 and 2 respectively, and the maximum quasi-steady pressures in evaporator were 0.49MPa and 0.67MPa in Runs 1 and 2. No secondary tube failure was observed. The rupture disc of evaporator (RD601) burst at 1.1s in Run-1 and at 0.7s in Run-2 after water injected, and the pressure relief system was well-functioned though a few items for improvement were found.