Home / Store / Journals / Electronic Articles / Nuclear Technology / Volume 94 / Number 3 / Pages 394-406
Yassin A. Hassan, Mathangi Kalyanasundaram
Nuclear Technology / Volume 94 / Number 3 / Pages 394-406
Format:electronic copy (download)
A RELAP5/MOD2 computer code model for a Model Boiler-2 U-tube steam generator (UTSG) is developed to predict the thermal-hydraulic response of a UTSG during steady-state operation and for a loss-of-feedwater (LOF) transient. Steady-state conditions calculated by RELAP5 are compared with the measured data. The calculated heat transfer from the primary to the secondary side of the steam generator is found to be underpredicted by 30%. The heat transfer correlations used in existing thermal-hydraulic codes are developed for flow inside individual tubes and not for flow around tube bundles. Consequently, the secondary convective heat transfer is not accurately predicted by the codes. A revised version of the RELAP5 code with modified heat transfer correlations reasonably predicts the primary to the secondary heat transfer in bundle environments. Improved heat fluxes and heat transfer coefficients are obtained during steady-state and LOF accident transients. Steady-state behavior of the Semiscale MOD-2C steam generator is also computed with both the original and the revised versions of the code. Good agreement is achieved between the predictions and the test data when the modified heat transfer correlations are utilized.
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