Boiling Water Reactor Simulations, Models, and Benchmarking Using the Thermal Hydraulics Sub-channel Code Ctf

Open Access
Author:
Gosdin, Christopher A
Graduate Program:
Nuclear Engineering
Degree:
Master of Science
Document Type:
Master Thesis
Date of Defense:
None
Committee Members:
  • Maria Avravmova, Thesis Advisor
Keywords:
  • CTF
  • COBRA-TF
  • BWR
  • Thermal-Hydraulics
Abstract:
CTF, the version of the thermal-hydraulic sub-channel code COBRA-TF being jointly developed and maintained by Pennsylvania State University (PSU) and Oak Ridge National Laboratory (ORNL) for applications in the U.S. Department of Energy (DOE) supported Consortium for Advanced Simulation of Light Water Reactors (CASL) project, uses a two-fluid, three-field representation of two-phase flow, which makes the code capable of modeling two-phase flow in Boiling Water Reactors (BWR) during nominal operating conditions. The sub-channel code CTF is used for Pressurized Water Reactors (PWR) for best-estimate evaluations of the nuclear reactor safety margins; however, due to its capabilities, CTF is powerful and valuable computational tool for modeling BWRs. CTF has been subjected to a strict verification procedure, by addressing the mathematical accuracy of the numerical solutions on multiple stages. The code was then validated using numerous of experimental databases, including the U.S. Nuclear Regulatory Commission (NRC) / Nuclear Energy Agency of the Organization for Economic Co-operation and Development (NEA-OECD) Boiling Water Reactor Full Bundle Tests (BFBT) Benchmark. The BFBT benchmark contains a large amount of test cases representative of BWRs steady-state and off-nominal operating conditions, which makes it one of the most widely used benchmark for validating BWR modeling tools. Two of the main experimental tests involve critical power tests and void distribution tests. Specific experimental cases were chosen and simulated using CTF. Statistical studies were carried out on the void distribution cases to evaluate the code modeling uncertainties. This thesis also focuses on application of CTF to mini- and whole-core BWR calculations on a pin-cell resolved level; as well as on demonstrating that CTF can properly model bypass flow in BWR cores. To increase the confidence in the CTF’s BWR modeling capabilities, extensive simulations have been performed using the international NEA-OECD / US NNRC Oskarshamn-2 benchmark, including modeling of a single and 2x2 assemblies on a pin-by-pin level, and a full core model on an assembly level. Each model is varied, with an increasing amount of detail. The results demonstrate that CTF is capable of modeling basic and complex BWR core configurations and operating conditions. Using the three Oskarshamn-2 simulations, CTF’s capabilities of modeling BWRs was further verified.