Critical flow in small nozzles for saturated and subcooled water at high pressure. [PWR]
Critical flow rate measurements of 4 mm and 16 mm nozzles have been performed with saturated and subcooled water at high pressure. The steady state and transient critical flow tests were conducted by discharging the fluid from a pressurized vessel through a blowdown leg. The fluid stagnation conditions upstream of the nozzle were measured by a gamma densitometer, thermocouple, and pressure transducer. The pressure and temperature of the tests range from 4.5 MPa to 15.0 MPa and from 530 K to 560 K, respectively. The results show that the flow upstream of the nozzle is stratified. The discharge mass flux obtained by this experiment is in good agreement with General Electric (GE) critical flow test data and Henry-Fauske and Burnell critical flow model predictions using a multiplier of 1.0 +- 0.3.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5080651
- Report Number(s):
- CONF-801102-10; TRN: 80-016749
- Resource Relation:
- Conference: ASME winter annual meeting, Chicago, IL, USA, 16 Nov 1980
- Country of Publication:
- United States
- Language:
- English
Similar Records
Summary on the depressurization from supercritical pressure conditions
Critical flow of saturated and subcooled water at high pressure. [PWR; BWR]
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HYDRAULICS
NOZZLES
CRITICAL FLOW
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
ECCS
REACTOR SAFETY
TEMPERATURE GRADIENTS
TEST FACILITIES
ACCIDENTS
COOLING SYSTEMS
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled