HIGH-FIDELITY MULTI-PHYSICS COUPLING FOR PREDICTION OF ANISOTROPIC POWER AND TEMPERATURE DISTRIBUTION IN FUEL ROD: IMPACT ON HYDRIDE DISTRIBUTION
Open Access
- Author:
- Davis, Ian James
- Graduate Program:
- Nuclear Engineering
- Degree:
- Master of Science
- Document Type:
- Master Thesis
- Date of Defense:
- None
- Committee Members:
- Maria Nikolova Avramova, Thesis Advisor/Co-Advisor
- Keywords:
- hydrogen distribution
hydride
BISON
DeCART
CTF
COBRA
MOOSE
clad temperature
coupling
high-fidelity
multi-physics - Abstract:
- Hydrogen absorbed into the nuclear fuel cladding during reactor exposure is distributed in highly inhomogeneous fashion as a result of temperature and stress gradients. To correctly describe and predict this hydrogen distribution there exists a need for multi-physics coupling to provide accurate heterogeneous temperature distributions in the cladding. The Department of Energy (DOE) recognized the need for better hydrogen modeling, and generously sponsored a project at the Pennsylvania State University (PSU) to investigate this need under the Nuclear Energy University Programs (NEUP). Furthermore, this study was conducted with the idea of a two-fold approach: 1. Combine accurate high-fidelity thermal-hydraulic models for heat transfer, reactor physics models for neutron flux, and thermal-mechanics models for fuel performance calculations to acquire detailed temperature and stress distributions in the fuel rod; 2. Analytically model and experimentally test the temperature and/or stress dependent hydrogen pick-up, diffusion, and precipitation in nuclear fuel cladding. The work and results specific to this thesis fall under the first part of the two-fold approach. Further detailing the first part, combination of the computational models is achieved by coupling a subchannel code to a neutronics code and a fuel performance code. The thermal-hydraulics code chosen is COBRA-TF (CTF), which is a subchannel code that was modernized and further developed at PSU. CTF’s capabilities include modeling two-phase flow in transient or quasi-steady-state conditions for Light Water Reactor (LWR) design and safety analysis. The neutronics code used in this research is DeCART (supported by the University of Michigan (UM)), a code which uses the method of characteristics approach to calculate the three dimensional (3-D) neutron flux. The power distribution with respect to (x,y,z) coordinates calculated with DeCART was used to calculate the radial power fraction (ratio of power produced in one fuel pin to the power averaged over all the fuel pins in the array), and the axial and radial pin power distributions for input in CTF. This data was, in turn, used to generate the temperature and stress distributions in the cladding. To do this, CTF was coupled to the fuel performance code BISON. BISON is a finite element code developed by Idaho National Laboratory (INL) to study all aspects of thermal and mechanical behavior of fuel rods in the reactor core. Outer clad surface temperature distributions from CTF and power distributions from DeCART are used to accurately model single fuel pins in BISON, which calculates temperature gradients in the cladding material.