RELAP5 modeling of the Westinghouse model D4 steam generator
Journal Article
·
· Nuclear Technology; (United States)
OSTI ID:6741108
- Univ. of Ljubljana (Slovenia)
The steam generator is one of the most important components of a pressurized water reactor (PWR) nuclear power plant. Thus, the ability to model and predict the steam generator steady-state and transient thermal-hydraulic behavior is a prerequisite for performing safety analyses of PWR systems. A RELAP5 model of the Westinghouse D4 steam generator with a 70/30 split feedwater system has been developed, and it is tested by simulating five secondary-side-initiated transients. This study of primary-to-secondary heat transfer and the secondary coolant vaporization process has enabled the primary coolant cooldown to be maximized, as required for performing a conservative steamline break analysis. These tests were realized using the RELAP5/MOD2.36.05 and RELAP5/MOD3.5M5 computer codes.
- OSTI ID:
- 6741108
- Journal Information:
- Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 101:2; ISSN 0029-5450; ISSN NUTYBB
- Country of Publication:
- United States
- Language:
- English
Similar Records
Comparison study of the Westinghouse model E steam generator using RELAP5/MOD2 and RETRAN-02
FLASH predictions of the MB-2 steamline break tests
FLASH predictions of the MB-2 steam line break tests
Conference
·
Wed Dec 31 23:00:00 EST 1986
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6942619
FLASH predictions of the MB-2 steamline break tests
Conference
·
Tue Dec 31 23:00:00 EST 1991
· Transactions of the American Nuclear Society; (United States)
·
OSTI ID:6672514
FLASH predictions of the MB-2 steam line break tests
Conference
·
Wed Dec 30 23:00:00 EST 1992
·
OSTI ID:10106485
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
BOILERS
COMPUTER CODES
COOLING SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EVAPORATION
HEAT TRANSFER
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
PHASE TRANSFORMATIONS
POWER PLANTS
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SECONDARY COOLANT CIRCUITS
STEADY-STATE CONDITIONS
STEAM GENERATORS
THERMAL POWER PLANTS
THERMAL REACTORS
TRANSIENTS
VAPOR GENERATORS
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
BOILERS
COMPUTER CODES
COOLING SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EVAPORATION
HEAT TRANSFER
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
PHASE TRANSFORMATIONS
POWER PLANTS
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SECONDARY COOLANT CIRCUITS
STEADY-STATE CONDITIONS
STEAM GENERATORS
THERMAL POWER PLANTS
THERMAL REACTORS
TRANSIENTS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS