An Investigation into Critical Heat Flux Correlations within CTF

Open Access
Bieniawski, Austin James
Area of Honors:
Mechanical Engineering
Bachelor of Science
Document Type:
Thesis Supervisors:
  • Nicholas Robert Brown, Thesis Supervisor
  • Dr. Jacqueline Antonia O'Connor, Honors Advisor
  • Alexander S Rattner, Faculty Reader
  • CTF
  • CHF
  • PWR
  • Mechanical Engineering
  • Nuclear Engineering
  • W-3
  • Groeneveld
The safety margins of a nuclear reactor is something that is under constant scrutiny but is difficult to make progress towards due to the need for valid and accurate simulations since safety concerns severely limit the amount of actual tests that can be done. In order to work towards safety margin improvement, a validation of a model of a Westinghouse 17x17 Pressurized Water Reactor in CTF at a steady-state and 50% overpower condition and a comparison of CHF correlations including the W-3 correlation and the Groeneveld lookup tables was needed. The model was validated in CTF versus a provided model in COBRA-EN using factors such as temperature, density, void fraction, and pressure drop. The various CHF models were compared looking at mainly DNBR at various axial locations on the fuel rods. This model was determined to be valid and, in some areas, significantly more accurate than previous models. Both the CHF models proved accurate, and can be reasonably used to investigate alternative cladding materials in the future in order to potentially increase reactor safety margins.