Abstract
In a hypothetical large break loss of coolant accident (LOCA), a break occurs on one of the cold legs and the emergency core cooling system (ECCS) must provide sufficient coolant to the core to remove decay heat and prevent the cladding from exceeding 1477.6 K, the temperature at which the thermal integrity of the cladding material is compromised. During reflood, flow to the core is gravity driven, resulting in an oscillatory delivery of coolant to the core. These oscillations are attributed to vapor generation in the core and the dynamic response of the downcomer water level. Most reflood experiments have been conducted with constant forced reflood rates, and have not considered the effect of oscillations on rod bundle thermal-hydraulics. The few studies conducted for oscillating flow indicate enhanced entrainment of liquid at the quench front. While higher entrainment can provide precursory cooling ahead of the quench front, it can also expel more coolant out of the system for oscillatory reflood. The amount of liquid entrained can be significant because in an accident scenario, the quench rate will be slowed and it can take longer to fully recover the core. At the NRC-PSU Rod Bundle Heat Transfer (RBHT) Test Facility, an electrically heated 7×7, 3.66 m rod bundle array has the capabilities to perform both constant and oscillatory forced flooding rate experiments. The heavily instrumented facility is equipped with seven spacer grids to analyze the droplet and heat transfer phenomena. In this study, reflood experiments have been performed in the RBHT test facility to investigate the separate effects of the oscillation frequency (4 to 20 seconds), magnitude (2.5 to 10.2 cm/s), and nominal flooding rates (2.5 to 5.1 cm/s) on the entrained droplet dynamics and heat transfer under constant and oscillatory flow conditions. The results have been compared and analyzed for the observed phenomena.
Original language | English (US) |
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State | Published - Jan 1 2017 |
Event | 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 - Xi'an, Shaanxi, China Duration: Sep 3 2017 → Sep 8 2017 |
Other
Other | 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 |
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Country | China |
City | Xi'an, Shaanxi |
Period | 9/3/17 → 9/8/17 |
All Science Journal Classification (ASJC) codes
- Nuclear Energy and Engineering
- Instrumentation