https://orcid.org/0000-0003-3424-4153
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Abstract
As one of the future directions of nuclear energy development, small modular reactor (SMR) designs meet the demands of safety, sustainability, and efficiency by eliminating circulating pumps and using natural circulation–driven flows to transfer fission energy to power. However, natural circulation–driven flows could be affected by two-phase-flow instability that may occur during accidental scenarios of pressurized water reactor (PWR)-type SMRs due to relatively small driving force. In view of the influence of two-phase-flow instability during accident transients for a PWR-type SMR, experiments are performed in a well-scaled test facility to investigate potential thermal-hydraulic flow instabilities during blowdown events. The test facility has a height of 3.44 m, and the operating pressure limit is 1.0 MPa. The scaling analyses ensure that the scaled phenomena, i.e., depressurization of the reactor pressure vessel (RPV) and emergency core cooling system valve actuation, could be accurately simulated in the test facility. Important thermal-hydraulic parameters including RPV pressure, containment pressure, local void fraction and temperature, pressure drop, and natural circulation flow rate are measured and analyzed during the blowdown events. Test results show that throughout the experiment the liquid level is always maintained above the heated core and the RPV pressure decreases. Oscillations of the natural circulation flow rate, water level, and pressure drop are observed during blowdown transients. Specific reasons for and mechanisms of the observed instability phenomena are discussed.
Nomenclature
| A = | = | cross-sectional area (m2) |
| cp = | = | heat capacity [J/(kg·°C)] |
| g = | = | gravitational acceleration (m/s2) |
| ifg = | = | latent heat (J/kg) |
| isub = | = | subcooling in terms of enthalpy (J/kg) |
| l = | = | axial length scale (m) |
| m = | = | mass flow rate (kg/s) |
| Nd = | = | drift number |
| NFr = | = | Froude number |
| Nsub = | = | subcooling number |
| Nth = | = | thermal inertial ratio number |
| Npch = | = | Zuber (phase change) number |
| Q = | = | total heat transfer rate (W) |
| u = | = | velocity (m/s) |
| Vgj = | = | drift velocity (m/s) |
Greek
| α = | = | void fraction |
| Δ = | = | difference |
| ρ = | = | mass density (kg/m3) |
| ψp = | = | parameter of prototype |
| ψm = | = | parameter of model |
Subscript
| f = | = | liquid |
| g = | = | gas |
| o = | = | reference point/component (heated section) |
| R = | = | ratio of model over prototype |
| s = | = | solid |
Acknowledgments
This material is based upon work supported under a U.S. Department of Energy Nuclear Energy University Program.