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

Abstract

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 conductancemore » 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.« less

Authors:
;  [1];  [2]
  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)
Publication Date:
OSTI Identifier:
23050373
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 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); Journal ID: ISSN 0003-018X
Country of Publication:
United States
Language:
English
Subject:
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

Citation Formats

Toptan, Aysenur, Avramova, Maria N., and Salko, Robert K. Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions. United States: N. p., 2017. Web.
Toptan, Aysenur, Avramova, Maria N., & Salko, Robert K. Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions. United States.
Toptan, Aysenur, Avramova, Maria N., and Salko, Robert K. 2017. "Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions". United States.
@article{osti_23050373,
title = {Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions},
author = {Toptan, Aysenur and Avramova, Maria N. and Salko, Robert K.},
abstractNote = {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.},
doi = {},
url = {https://www.osti.gov/biblio/23050373}, journal = {Transactions of the American Nuclear Society},
issn = {0003-018X},
number = ,
volume = 116,
place = {United States},
year = {Sat Jul 01 00:00:00 EDT 2017},
month = {Sat Jul 01 00:00:00 EDT 2017}
}