Abstract
Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.
Highlights
In the earthquake and tsunami-induced station blackout (SBO) accident at the Fukushima Daiichi boiling water reactor, water was alternatively injected into the reactor as a flexible applied action utilizing fire engines due to loss of the core cooling functions [1]
An optional accident management (AM) measure was initiated by fully opening the PZR relief valve at 9400 s to enhance the primary depressurization, which caused that the primary pressure decreased to a certain low pressure of below 1 MPa (Figure 15) and an increase occurred in the PZR liquid level (Figure 21)
Three rig of safety assessment (ROSA)/large-scale test facility (LSTF) tests were carried out simulating AM measures during pressurized water reactor (PWR) SBO transient with loss of primary coolant under assumptions of nitrogen gas ingress and totally failed high-pressure injection system
Summary
In the earthquake and tsunami-induced station blackout (SBO) accident at the Fukushima Daiichi boiling water reactor, water was alternatively injected into the reactor as a flexible applied action utilizing fire engines due to loss of the core cooling functions [1]. Yoshihara [12] has put forward that accumulator (ACC) system of ECCS should be isolated by the closure of motoroperated valves located at the outlet of ACC tanks utilizing emergency power generators when the primary pressure decreases to a certain low pressure during PWR SBO transient with leakage from primary coolant pump seal. We carried out three tests with the LSTF simulating AM measures during PWR SBO transient with loss of primary coolant under assumptions of nitrogen gas inflow and totally failed high-pressure injection system concerning the severe multiple system failures in 2015–2017. The LSTF initial core power of 10 MW corresponds to 14% of the volumetric-scaled (1/48) PWR nominal core power because of a limitation in the capacity of power supply
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