Loss-of-Fluid Test Facility
E99197
The Loss-of-Fluid Test Facility is a specialized experimental installation used to study and simulate loss-of-coolant and related accident scenarios in nuclear reactors.
Statements (45)
| Predicate | Object |
|---|---|
| instanceOf |
experimental facility
→
nuclear safety research facility → thermal-hydraulic test facility → |
| aimsTo |
improve predictive capability of thermal-hydraulic models
→
reduce uncertainties in reactor safety analysis → |
| contributesTo |
development of safety standards for nuclear power plants
→
understanding of nuclear reactor accident progression → validation of best-estimate thermal-hydraulic codes → |
| dataUsedBy |
nuclear regulators
→
nuclear safety analysts → reactor designers → |
| field |
nuclear engineering
→
reactor safety → thermal-hydraulics → |
| hasCharacteristic |
capable of high-pressure and high-temperature operation
→
instrumented for detailed thermal-hydraulic measurements → scaled representation of reactor coolant system components → |
| measures |
mass flow rates
→
pressure distributions → temperature distributions → void fractions in two-phase flow → |
| purpose |
provide experimental data for code validation
→
simulate reactor accident transients → study loss-of-coolant accident scenarios in nuclear reactors → support development of emergency core cooling system designs → |
| relatedTo |
emergency core cooling system performance
→
loss-of-coolant accident analysis → nuclear power plant safety assessment → |
| researchFocus |
blowdown and reflood phenomena
→
loss-of-coolant accidents → reactor system thermal-hydraulic behavior → safety system performance under accident conditions → two-phase flow in reactor systems → |
| simulates |
breaks in reactor coolant system piping
→
depressurization of reactor coolant systems → emergency core cooling system injection conditions → loss-of-coolant transients → |
| supports |
development of accident management strategies
→
improvement of reactor safety systems → |
| typeOfTests |
integral system tests
→
transient thermal-hydraulic tests → two-phase flow experiments → |
| usedFor |
benchmarking of system thermal-hydraulic computer codes
→
evaluation of nuclear reactor safety margins → supporting regulatory safety analyses → |
Referenced by (1)
| Subject (surface form when different) | Predicate |
|---|---|
|
Idaho National Engineering Laboratory
→
|
hasFacility |