Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Journal Article · Sat Jul 01 00:00:00 EDT 2017 · Transactions of the American Nuclear Society

OSTI ID:23050373

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

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1. Nuclear Engineering Department, North Carolina State University, Raleigh, NC 27695 (United States)
2. Oak Ridge National Laboratory, CASL Division, Oak Ridge, TN 37831 (United States)

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

• OSTI ID: 23050373
• Journal Information: Transactions of the American Nuclear Society, Vol. 116; Conference: 2017 Annual Meeting of the American Nuclear Society, San Francisco, CA (United States), 11-15 Jun 2017; Other Information: Country of input: France; 7 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
BOILING
CLADDING
COMPUTERIZED SIMULATION
CORROSION
FINITE ELEMENT METHOD
FISSION PRODUCT RELEASE
FUEL PELLETS
FUEL RODS
HYDRIDATION
HYDROGEN
IRRADIATION
NUCLEAR FUELS
OXIDATION
PERFORMANCE
STEADY-STATE CONDITIONS
THERMAL HYDRAULICS
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS