Void Drift Validation and Sensitivity Analysis in CTF

Open Access
Gergar, Marcus Stephen
Graduate Program:
Nuclear Engineering
Master of Science
Document Type:
Master Thesis
Date of Defense:
Committee Members:
  • Maria Nikolova Avramova, Thesis Advisor
  • CTF
  • void drift
  • turbulent mixing
  • validation
CTF, the version of the thermal-hydraulic subchannel code COBRA-TF being jointly developed by The Pennsylvania State University (PSU) and Oak Ridge National Laboratory (ORNL) for applications in the U.S. Department of Energy (DOE) Consortium for Advanced Simulation of Light Water Reactors (CASL) project, employs a simplified void drift model proposed by Levy, which states that the equilibrium void distribution is directly proportional to the equilibrium mass flux distribution. This phenomenon equates to void migrating from smaller cross-sectional area subchannels to larger cross-sectional area subchannels. In an ongoing effort to validate CTF’s governing equations, various experiments are selected to ensure the validity of the code’s models describing the void drift phenomenon. To further this goal, the air/water experiment described in the NUREG/CR-3373 report, issued by the U.S. Nuclear Regulatory Commission (NRC); the void distribution data in the U.S. NRC/NEA-OECD (Nuclear Energy Agency of the Organization for Economic Co-operation and Development) Pressurized Water Reactor Subchannel and Bundle Tests (PSBT) Benchmark Series 5 through 7 experiments, and the void distribution data in the General Electric (GE) 3x3 bundle experiments were selected. The NUREG/CR-3373 air/water experiment was conducted at the Rensselaer Polytechnic Institute (RPI) Jonsson Engineering Center and included a 2x2 rod bundle facility. Air was injected at the base of the rod bundle to simulate Boiling Water Reactor (BWR) flow conditions. Because CTF does not directly solve the non-condensable gas equations, these experiments were modeled with saturated steam with phase change disabled instead of air. Coupled with CTF’s convergence issues at the low experimental pressures, comparable flow conditions (i.e., system pressure, water density, steam density) were not reached. However, this simulation was used to validate the qualitative void drift trends as described in Levy’s model. The PSBT experiments were conducted to support the advancement of subchannel analysis of fluid flow in rod arrays. The PSBT exercise used in this validation provided average center subchannel void content, which was then compared with the CTF simulation results. This data was used to provide validation for the void content of center subchannel types, which should provide higher than bundle average void as implied by Levy’s model. These analyses provided successful validation of the void drift mechanism followed by a sensitivity study on center subchannel void with respect to various system boundary conditions and mixing options. The GE 3x3 experiments were designed to contribute to the improvement of thermal/hydraulic (T/H) characteristics of nuclear reactor cores and were used in this validation to provide reliable void distribution data. The experimental results included equilibrium quality and mass flux. This data was then compared to CTF simulation results with various mixing options for void drift and turbulent mixing. This comparison demonstrated acceptable agreement with experimental results for each subchannel type and displayed expected trends when mixing options were altered. Additionally, sensitivity studies on individual subchannel void content were performed with respect to various mixing options. Overall, these validation and sensitivity exercises provide a sound addition to CTF’s expansive validation suite. The lateral void drift mechanism as proposed by Levy implemented in CTF provides the desired and expected results. These validation and sensitivity exercises were performed within a collaboration with CASL.