RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

Conference · Tue Oct 31 23:00:00 EST 1989 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5420692

Blanchat, T; Hassan, Y ^[1]

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

The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data.

• OSTI ID: 5420692
• Report Number(s): CONF-891103--
• Journal Information: Transactions of the American Nuclear Society; (United States), Journal Name: Transactions of the American Nuclear Society; (United States) Vol. 60; ISSN TANSA; ISSN 0003-018X
• Country of Publication: United States
• Language: English

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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42 ENGINEERING
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ACCIDENTS
BOILERS
BOILING
BUBBLES
COMPUTER CODES
COOLING SYSTEMS
CRITICAL HEAT FLUX
ENERGY SYSTEMS
ENERGY TRANSFER
FLOW MODELS
FLUID MECHANICS
GEOMETRY
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
IMPLEMENTATION
MATHEMATICAL MODELS
MATHEMATICS
MECHANICS
MODIFICATIONS
NUCLEATE BOILING
OPERATION
PHASE TRANSFORMATIONS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR OPERATION
REACTORS
SECONDARY COOLANT CIRCUITS
STEADY-STATE CONDITIONS
STEAM GENERATORS
TRANSIENTS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS