AP600 passive residual heat removal heat exchanger test

Conference · Mon Jan 01 00:00:00 EST 1990 · Transactions of the American Nuclear Society; (USA)

OSTI ID:5822274

Corletti, M M; Hochreiter, L E; Squarer, D

The AP600 reactor is a pressurized water reactor being designed to utilize a passive residual heat removal (PRHR) heat exchanger as the safety grade means for residual heat removal. The PRHR heat exchanger is utilized during many design basis events and is especially important in mitigating non-loss-of-coolant accidents such as loss of normal feedwater and feedwater line break. The PRHR system transfers decay heat from the reactor coolant system to the containment by heating and boiling the water in the in-containment refueling water storage tank (IRWST). The steam produced transfers heat to the atmosphere by condensing on the inside of the containment shell. The condensate is collected by gutters on the containment shell and is returned to the IRWST, which provides a heat sink for an indefinite amount of time. The PRHR test facility is a prototypical representation of the PRHR heat exchanger with respect to tube material, diameter, pitch, and tube length, such that the gravity-induced flow characteristics in the pool are representative of the design. The main scaling parameter for the pool is the pool volume per tube, which preserves the buoyancy and pool mixing effects. A generalized PRHR boiling correlation was developed using the approach given by Rohsenow such that pressure effects can be induced.

• OSTI ID: 5822274
• Report Number(s): CONF-901101-; CODEN: TANSA
• Journal Information: Transactions of the American Nuclear Society; (USA), Vol. 62; Conference: American Nuclear Society (ANS) winter meeting, Washington, DC (USA), 11-15 Nov 1990; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PWR TYPE REACTORS
RHR SYSTEMS
NATURAL CONVECTION
AFTER-HEAT REMOVAL
CONTAINMENT SYSTEMS
CRITICAL HEAT FLUX
FLOW RATE
HEAT SINKS
HEAT TRANSFER
HYDRAULICS
MATHEMATICAL MODELS
MITIGATION
MIXING
NUCLEATE BOILING
PASSIVITY
PERFORMANCE
PRIMARY COOLANT CIRCUITS
REACTOR ACCIDENTS
REACTOR SAFETY
STEAM
TEMPERATURE CONTROL
TEST FACILITIES
THERMOCOUPLES
ACCIDENTS
BOILING
CONTAINMENT
CONTROL
CONVECTION
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
HEAT FLUX
MASS TRANSFER
MEASURING INSTRUMENTS
MECHANICS
PHASE TRANSFORMATIONS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
REMOVAL
SAFETY
SINKS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled