U-tube steam generator predictions: New tube bundle heat transfer correlations

Conference · Wed Nov 01 00:00:00 EST 1989 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5376125

Kalyanasundaram, M; Hassan, Y A ^[1]

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1. Texas A M Univ., College Station (USA)

Steam generators play a very important role in the safe and reliable operation of pressurized water reactor (PWR) power plants. They determine the thermal-hydraulic responses of the primary coolant system during operational and accident transients. The Westinghouse Model Boiler No. 2 (MB-2) is an experimental test facility to provide comprehensive data on the steady-state and transient responses of a U-tube steam generator. In a previous study, an optimized nodalization scheme for MB-2 was developed using the one-dimensional RELAP5/MOD2 system analysis code. Analysis of the system steady-state and transient responses predicted by RELAP5 indicated an underprediction of the secondary convective heat transfer from the primary to the secondary side of the steam generator. The objective of the present study is to evaluate the effect of modifications in the heat transfer correlations and critical heat flux correlations in RELAP5 on predicting better secondary heat transfer during steady state and loss-of-feedwater (LOF) transient in a U-tube steam generator. The modified code has mitigated the underprediction of the secondary convective heat transfer during both steady state and LOF transient.

• OSTI ID: 5376125
• Report Number(s): CONF-891103-; CODEN: TANSA; TRN: 91-024628
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 60; Conference: Winter meeting of the American Nuclear Society (ANS) and nuclear power and technology exhibit, San Francisco, CA (United States), 26-30 Nov 1989; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PWR TYPE REACTORS
STEAM GENERATORS
HEAT TRANSFER
HYDRAULICS
COMPUTERIZED SIMULATION
FEEDWATER
HEAT FLUX
LOSS OF COOLANT
MITIGATION
MODIFICATIONS
ONE-DIMENSIONAL CALCULATIONS
PRIMARY COOLANT CIRCUITS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
RELIABILITY
STEADY-STATE CONDITIONS
TEST FACILITIES
TRANSIENTS
TUBES
ACCIDENTS
BOILERS
COMPUTER CODES
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
HYDROGEN COMPOUNDS
MECHANICS
OXYGEN COMPOUNDS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SAFETY
SIMULATION
VAPOR GENERATORS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200* - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled