Abstract
The flooding caused by counter-current gas-fluid flow is usually an undesirable phenomenon in industrial systems. For example, steam-water flooding might occur during a Loss of Coolant Accident (LOCA) in a nuclear reactor when the downward flow emergency core cooling systems (ECCS), are activated against current steam flow through the coolant boiling in the fuel channel. If the steam velocity is high enough, the water entering from the ECCS will drift away so that the fuel cladding temperature may increase up to the melting temperature due to the residual heat in the fuel and the limited heat transfer. In this study, a transparent experimental setup was designed and produced in order to investigate the flooding phenomenon in vertical annular corrugated channels. At the first stage, an air-water counter-current flow was tested. The experimental system included an air blower, which injects the air upward inside the channel (simulating the steam flow at LOCA). The water is injected to the test channel downwards through a nozzle. The sensitivity of the following parameters was checked: water flow rate, air flow rate and the test element corrugation geometry. The values of the air and water flow rate and the pressure drop on the tested element were measured and collected by a Pico data logger. It was found that the flooding conditions might be detected either by visual observation or by differential pressure sequence measurements. In addition, significant hysteresis between the flooding and deflooding air velocity was found for constant water flow rate. For both smooth and corrugated test element the deflooding air velocity decreased gradually due to the water velocity increase. The deflooding velocity of the corrugated test element is lower than for the smooth test element. A possible reason for that behavior is that for corrugated element the turbulence intensity is more significant than for the smooth element. In addition, some horizontal air velocity component might exist.
Original language | English |
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Pages | 5218-5229 |
Number of pages | 12 |
State | Published - 1 Jan 2019 |
Event | 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, United States Duration: 18 Aug 2019 → 23 Aug 2019 |
Conference
Conference | 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 |
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Country/Territory | United States |
City | Portland |
Period | 18/08/19 → 23/08/19 |
Keywords
- Corrugated element
- Counter-current flow
- Deflooding
- Flooding
- Gas-fluid two-phase flow
ASJC Scopus subject areas
- Nuclear Energy and Engineering
- Instrumentation