Water Reactor Safety Research Division: Quarterly Progress Report, July 1-September 30, 1980

Technical Report · Sat Nov 01 04:00:00 EST 1980

DOI:https://doi.org/10.2172/5450095· OSTI ID:5450095

Abuaf, N. ^[1]; Cerbone, R. J. ^[1]; Levine, M. M. ^[1]; Saha, P. ^[1]; van Rooyen, D. ^[1]

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1. Brookhaven National Laboratory (BNL), Upton, NY (United States)

A summary of all the experimental results for the Thermal/Hydraulic Development Program, i.e., pressure distributions, transverse distributions of the chordal averaged void fraction at given axial locations, and the area averaged void fraction distributions are presented. The effects of the flow parameters studied, i.e., inlet temperature, inlet mass flux and exit pressure on the recorded data are highlighted. RAMONA-III is being tested and modified to calculate small pipe break LOCAs in a BWR. The MARK II once-through steam generator model has been implemented in the IRT code and is currently being tested for an over cooling transient for a typical B and W plant. The version of the RETRAN code at BNL has been modified to improve the plotting capability on the BNL computer. Work on the independent assessment of TRAC-P1A has been terminated during the reporting quarter since an up-to-date version of TRAC i.e., TRAC-PD2, was received at BNL in July 1980. However, for the sake of completeness, a few simulations of the Super-CANON, RPI phase-separation, and FRIGG rod bundle tests have been performed with TRAC-P1A. The TRAC-PD2, Version 25.3, has been made operational on the BNL computer. Seven of the LASL-supplied sample problems have been run successfully.

• Research Organization: Brookhaven National Laboratory (BNL), Upton, NY (United States)
• Sponsoring Organization: USDOE; USNRC
• DOE Contract Number: AC02-76CH00016
• OSTI ID: 5450095
• Report Number(s): NUREG/CR--1800; BNL-NUREG--51298; ON: TI85015884
• Country of Publication: United States
• Language: English

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