A feasibility study on feed and bleed for pressurized water reactors
Journal Article
·
· Am. Soc. Mech. Eng., (Pap.); (United States)
OSTI ID:6703407
By injecting coolant with a high pressure emergency core cooling system, and removing the heated/ vaporized fluid by way of the pressurizer power operated relief valve, primary feed and bleed cooling denotes an operation whereby reactor core cooling is maintained. This paper presents the results from an experimental and analytical study that includes a simplified analysis of mass and energy balances associated with the feed and bleed, examination of test data from the Semiscale system, RELAP5 code analyses of both Semiscale and a four-loop Westinghouse plant, and the primary coolant system behavior for a transient that leads to the need for feed and bleed. Examination of the parameters that govern a stable feed and bleed operation identifies four key parameters such as: core decay heat, cooling water injection capacity, power operated relief valve (PORV) energy removal rate, and PORV mass removal rate. A simplified analytical approach to determining if stable feed and bleed is feasible, has been developed and corroborated by experimental data and computer code calculations. The Semiscale tests have not only provided test data for code assessment, but also have identified the factors influencing the PORV discharge, which is the most variable of the boundary conditions influencing feed and bleed. The RELAP5 computer code has demonstrated the capability of predicting the Semiscale experiments, and when applied to a four-loop Westinghouse plant has indicated that primary feed and bleed is a viable cooling mechanism. This has also been shown by using the simplified analytical method.
- Research Organization:
- US Nuclear Regulatory Commission
- OSTI ID:
- 6703407
- Journal Information:
- Am. Soc. Mech. Eng., (Pap.); (United States), Journal Name: Am. Soc. Mech. Eng., (Pap.); (United States) Vol. 83-HT-16; ISSN ASMSA
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
COMPUTER CALCULATIONS
COMPUTER CODES
COOLANTS
COOLING SYSTEMS
DATA
ECCS
ENERGY BALANCE
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL DATA
FEEDWATER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
HYDROGEN COMPOUNDS
INFORMATION
MASS TRANSFER
MECHANICS
NUMERICAL DATA
OXYGEN COMPOUNDS
PERFORMANCE TESTING
PRESSURE GRADIENTS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
TESTING
TRANSIENTS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
COMPUTER CALCULATIONS
COMPUTER CODES
COOLANTS
COOLING SYSTEMS
DATA
ECCS
ENERGY BALANCE
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL DATA
FEEDWATER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
HYDROGEN COMPOUNDS
INFORMATION
MASS TRANSFER
MECHANICS
NUMERICAL DATA
OXYGEN COMPOUNDS
PERFORMANCE TESTING
PRESSURE GRADIENTS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTORS
TESTING
TRANSIENTS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS