Validation of LAPUR 6.0 Code (NUREG/CR-7124)
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Publication Information
Manuscript Completed: March 2012
Date Published: March 2012
Prepared by:
Oak Ridge National Laboratory
P. O Box 2008
Oak Ridge, TN 37831-6010
Authors:
Jose Melara, Carmen Lora (Iberdrola Ingenieria y Construccion)
Alberto Escrivá, Jose Luis Muñoz-Cobo (Universidad Politécnica de Valencia)
Manuel Albendea (Iberdrola Generación)
Jose March-Leuba (ORNL)
Dr. Tai L. Huang, NRC Project Manager
NRC Job Code J4430
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington DC 20555-0001
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Abstract
The capabilities of the computer code LAPUR have been upgraded. A new version, LAPUR6 r.0, has been implemented and validated for computing friction and local losses and capabilities for modelling bundles with variable cross areas.
The previous code (LAPUR5) did not consider channels with variable areas and did not model specifically the local pressure drops due to spacers in a bundle. The only way to take into account local pressure losses and gains due to spacers and area changes in LAPUR5 was to input a friction multiplier.
This report documents a twofold validation of the new thermal-hydraulic model implemented in LAPUR6. First, a comparison of each component of the pressure drop for a single-channel model using LAPUR6 and SIMULATE-3 was performed, the result of which showed very good agreement. Slight discrepancies in void fractions between the two codes were found, but the effect on total pressure drop was negligible. In addition, the LAPUR6 void fraction model was tested against FRIGG-2 void fraction data. Void fractions predicted by LAPUR6 showed deviations similar to those of other thermal-hydraulic codes when benchmarked against FRIGG-2 experimental data.
An extensive validation comparing measured against calculated core-wide (CW) decay ratios was also conducted. A set of Average Power Range Monitor (APRM) signals were recorded at steady state during the final coastdowns for Cycles 16b and 17 and start-up for Cycles 17 and 18 in Cofrentes Nuclear Power Plant (NPP). A detailed simulation of these activities was conducted with SIMULATE-3 using cycle-specific CASMO-4 cross sections and the recorded operating data. Selected quasi-steady-state points were analyzed using noise techniques, and decay ratio values were compared to LAPUR6 results. Finally, Cycle 6 Cofrentes out-of-phase (OOP) instability was reproduced using LAPUR6, and the resulting decay ratios showed excellent agreement with the measured data.
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