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4.8 FIBWR (Flow in Boiling Water Reactors)

DOI: 10.1615/ICHMT.1982.AdvCourHeatTransfNucReactSaf.440
pages 683-698

A. A. F. Amsari
Yankee Atomic Electric Company, Westborough, Massachusetts, USA

B. C. Slifer
Yankee Atomic Electric Company, Westborough, Massachusetts, USA

B. J. Gitnick
Science Applications, Inc., McLean, Virginia, USA

R. S. Borland
Science Applications, Inc., McLean, Virginia, USA

R. R. Gay
Rensselaer Polytechnic Institute, Troy, New York, USA


An accurate prediction of core wide flow and void distribution in currently operating BWRs was of interest to Yankee Atomic Electric Company as an owner of a BWR. Research indicated that the existing codes were far from meeting the objective, therefore, an effort to develop a new code that would meet the objective was undertaken. As a result, a computer code called FIBWR (Flow In Boiling Water Reactors) has been developed to analyze the steady-state thermal hydraulics of boiling water reactors. FIBWR code is unique in its ability to handle the varied geometrical configurations that are present in currently operating BWRs. Up to one hundred parallel flow channel types plus a bypass (leakage flow) region can be modeled. The detailed geometric modeling of each BWR fuel assembly includes the effects due to the inlet orifice, fuel support piece, lower tie plate, unheated fuel regions, grid spacers, water tubes, upper tie plate and chimney. Three bypass flow paths located downstream of the inlet orifice, but upstream of the active fuel region and up to eight bypass flow paths dependent on the core support plate pressure differential can be modeled. This paper describes the analytical methodology used in FIBWR code to solve the conservation equations of continuity, momentum and energy in steady-state, two-phase flow. Comparison of results with the experimental data (ASEA-ATOM FRIGG LOOP DATA) and results from a test case on Quad Cities Unit 1 reactor are also included.

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