Improvement of Wall Drag Modeling Capabilities of CTF

Journal Article · Fri Jul 01 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:23042910

Toptan, Aysenur; Avramova, Maria ^[1]; Salko, Robert K. ^[2]

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1. Nuclear Engineering Department, North Carolina State University, Raleigh, NC (United States)
2. Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. It solves discretized forms of nine coupled equations for the mass, momentum, and energy equations for the liquid and vapor fields; the mass and momentum equation for the droplet field; and the energy equation for solid structures. This work focuses on improving the wall drag modeling capabilities of CTF by adding new friction models that account for surface roughness. Prior to this work, CTF used simple correlations that assumed that surfaces were smooth. In reality, different materials in the reactor have varying degrees of roughness that act to increase drag and frictional pressure drop. The frictional pressure drop is important, as it partly defines the driving head necessary for achieving certain flow conditions. This is particularly important in passively cooled systems that rely on natural circulation. Therefore, the effects of surface roughness cannot be neglected, as it affects flow rate and distribution, heat transfer and turbulence. (authors)

• OSTI ID: 23042910
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 115; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
97 MATHEMATICS AND COMPUTING
BOILING
COMPUTERIZED SIMULATION
COOLANTS
DROPLETS
FLOW RATE
FRICTION
LIQUIDS
MATERIALS
NATURAL CONVECTION
PRESSURE DROP
ROUGHNESS
THERMAL HYDRAULICS
TURBULENCE
VAPORS