CTF is a modernized version of the thermal-hydraulics subchannel code COBRA-TF. CTF uses a two-ﬂuid, two-phase ﬂow modeling approach for modeling ﬂuid ﬂow and heat transfer and provides both subchannel and 3D solution options. Flow can be modeled as three independent ﬁelds: continuous liquid, vapor, and liquid droplets. Two-phase flow and heat transfer models consider both pre- and post-critical heat flux conditions. CTF also provides a set of solid thermal conduction modeling capabilities that can be used for tubes, cylinders, and nuclear fuel rod geometries. A nuclear fuel rod solver, CTFFuel, is also provided for modeling of burnup-dependent fuel temperatures.
Applications and Integrations
CTF is used primarily for modeling in-core light water reactor ﬂuid ﬂow and heat transfer in normal and oﬀ-normal operating conditions. It is used for performing steady-state and transient simulations of PWR and BWR conditions including depletion conditions, prediction of DNB, and RIA analysis
Coupling to the neutron transport code, MPACT, the crud-chemistry code, MAMBA, and the fuel performance code, BISON. CTF has also been integrated into the NEAMS Workbench and has been coupled to the NRC fuel performance code, FAST, as well as the system code, TRACE.
Planned capabilities for the future include:
Current development activities are focusing on improvements for BWR modeling and simulation, including development and implementation of more accurate closure models, addition of support for BWR-specific geometry, and addition of critical power modeling capabilities. Work is also being done to expand on the systems code coupling capability for modeling of LWR transients such as main steam-line break and loss of flow.
- CTF User Manual
- CTF Theory Manual
- R. Salko, V. Kumar, B. Hizoum, and W. Gurecky, “Improvements to CTF Closure Models for Modeling of Two-Phase Flow”, Oak Ridge National Laboratory, ORNL/TM-2020/1605, 2020.
- R. Salko, B. Hizoum, B. Collins, and M. Asgari, “Improvements to CTF for Modeling of Boiling Water Reactor Geometry and Operating Conditions”, Oak Ridge National Laboratory, ORNL/TM-2020/1746, 2020.
- N. Porter, R. Salko and M. Pilch, “Development and Implementation of a CTF Code Verification Suite”, Nuclear Engineering and Design, 370, pg. 110879, 2020.
- A. Wysocki, K. Borowiec, and R. Salko, “L3:PHI.TRN.P19.01 Coupling Interface to Systems Code for Transient Analysis”, Consortium for Advanced Simulation of Light Water Reactors, CASL-U-2019-1909-000, 2019.
- A. Toptan, R. Salko, M. Avramova, K. Clarno, and D. Kropaczek, ”A new fuel modeling capability, CTFFuel, with a case study on the fuel thermal conductivity degradation”. United States: N. p., 2018. Web. doi:10.1016/j.nucengdes.2018.11.010.
- R. Salko, et al., “Implementation of a Spacer Grid Rod Thermal-Hydraulic Reconstruction (ROTHCON) Capability into the Thermal-Hydraulic Subchannel Code CTF”, Nuclear Technology, https://doi.org/10.1080/00295450.2019.1585734, 2019.
- R. Salko, et al., “Development of COBRA-TF for modeling full-core, reactor operating cycles”, Advances in Nuclear Fuel Management V (ANFM 2015), 2015.
CTF prediction of liquid enthalpy distribution at most limiting point of main steamline break transient with stuck control assembly
Example of modeling resolution of CTF for a 4-loop PWR when used as part of VERA