CASL COBRA-TF improvements for PWR DNB analysis
- Department of Fuel Engineering and Safety Analysis, Westinghouse Electric Company, LLC, Cranberry Township, PA (United States)
- Department of Structural and Thermal Analysis, Sandia National Laboratories, Albuquerque, NM (United States)
- Oak Ridge National Laboratory, Oak Ridge, TN (United States)
- Global Technology Development, Westinghouse Electric Company, LLC, Cranberry Township, PA (United States)
COBRA-TF (CTF) is a thermal hydraulic (T/H) subchannel code using either three-dimensional (3D) Cartesian or subchannel coordinate formulations for two-phase fluid flow and heat transfer solutions. CTF has been improved under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program for Pressurized Water Reactor (PWR) applications, including software optimization, new closure models, pre- and post-processing and parallelization for modeling full reactor core T/H responses under normal operating and accident conditions. As a result of collaboration among CASL partners including the Westinghouse Electric Company, the Oak Ridge National Laboratory (ORNL), and the Sandia National Laboratories, additional modeling improvements were made to CTF specifically for PWR Departure from Nucleate Boiling (DNB) analysis, including a code option to evaluate fuel thermal margin in terms of DNB Ratio (DNBR) and an axial shape factor to account for effect of non-uniform axial power distribution on DNB. Multiple DNB correlations are now linked with CTF for different applications, including the Westinghouse proprietary WRB-1 correlation for fuel designs containing mixing vane grid spacers. The improved CTF code with the WRB-1 correlation (CTF/WRB-1) was validated using the DNB data from the PWR Subchannel Bundle Tests (PSBT). In addition to the comparison with the test data, the CTF/WRB-1 DNBR results and the associated local fluid conditions were compared to the results of the Westinghouse T/H design code, VIPRE-W, which is an enhanced version of the VIPRE-01 code originally developed by the Electric Power Research Institute (EPRI). The comparisons showed that CTF/WRB-1 DNBR predictions are in good agreement with the VIPRE-W results within the applicable range of the DNB correlation. A model sensitivity study was performed to confirm that the CTF void drift model had an insignificant effect on DNBR under the steam line break (SLB) low pressure condition. (authors)
- Research Organization:
- American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22765226
- Resource Relation:
- Conference: TOP FUEL 2016: LWR fuels fuels with enhanced safety and performance, Boise, ID (United States), 11-15 Sep 2016; Other Information: Country of input: France; 15 refs.; Related Information: In: TOP FUEL 2016 Proceedings| 1670 p.
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
42 ENGINEERING
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
C CODES
COMPARATIVE EVALUATIONS
COMPUTERIZED SIMULATION
CORRELATIONS
DEPARTURE NUCLEATE BOILING
HEAT TRANSFER
NUCLEAR FUELS
NUCLEAR INDUSTRY
ORNL
PWR TYPE REACTORS
REACTOR CORES
SANDIA NATIONAL LABORATORIES
SENSITIVITY ANALYSIS
STEAM LINES
THERMAL HYDRAULICS
THREE-DIMENSIONAL CALCULATIONS
TWO-PHASE FLOW