Multi-scale simulation of advanced reactor fuel and cladding concepts
Open Access
- Author:
- Rezwan, Aashique Alam
- Graduate Program:
- Mechanical Engineering
- Degree:
- Doctor of Philosophy
- Document Type:
- Dissertation
- Date of Defense:
- December 06, 2019
- Committee Members:
- Reuben H Kraft, Dissertation Advisor/Co-Advisor
Reuben H Kraft, Committee Chair/Co-Chair
Md Amanul Haque, Committee Member
Jing Du, Committee Member
Arthur Thompson Motta, Outside Member
Michael R Tonks, Special Member
Daniel Connell Haworth, Program Head/Chair - Keywords:
- multi-scale simulation
MOOSE
BISON
phase-field
fracture
composite cladding
light water reactor
nuclear fuel performance
anisotropic thermal expansion
alpha-uranium
mesoscale modeling
irradiation growth - Abstract:
- The development of the advanced reactor fuel and cladding concepts, with better safety features during normal operating conditions and in accident scenarios, requires understanding of the material behavior under reactor conditions. Though experimental techniques for assessing the mechanical and thermal response of new fuel and cladding concepts have improved significantly, the time and cost associated with the evaluation of these data are very high. Multi-scale modeling and simulation can play a key role in generating data to help understand and modify new concepts in a fast and inexpensive way, reducing the number of necessary experiments. In this work, multi-scale simulations are used to investigate the behavior of a new cladding concept for light water reactor (LWR) fuel and the behavior of metallic fuel for sodium fast reactors (SFRs). A near-term solution for augmenting the safety of LWRs is presented by assessing the performance of a multi-metallic layered composite (MMLC) cladding. This cladding is designed to sustain a longer period of time during accident scenarios for LWR fuel. The proposed MMLC cladding concept joining Zircaloy to steel via interaction barrier layers will increase the corrosion resistance but result in a neutron absorption penalty. This cladding could also result in negative impacts due to mismatches between the mechanical behavior of the various layers. In this study, the mechanical performance of the MMLC has been evaluated using two types of simulations: small scale thermo-mechanical simulations and full-length fuel rod simulations. Both types of simulations were carried out using the BISON fuel performance code. The small scale simulations predicted a reduction in residual stress and elastic deformation compared with the standard zircaloy cladding. The fuel rod simulations showed a decrease in creep strain, larger pellet-clad gap and possible plastic deformation under long time use when compared to zircaloy and stainless-steel cladding alone. The performance of various gap widths and layer thicknesses were compared to assist in the design of MMLC cladding. It is recommended to design an MMLC with a thickness ratio of at least 1.75 and an initial gap width of 40 $\mu$m. Metallic fuels, especially the U-Pu-Zr system, has already proven an excellent choice for SFRs due in part to passive safety features and excellent thermal conductivity. While experimental research has advanced the performance of a specific fuel design, many inherent properties of metallic fuel are not fully understood. One constituent of the U-Pu-Zr fuel system is alpha-uranium (alpha-U), which has very anisotropic mechanical behavior. Thus, another major topic of this dissertation is to investigate the fracture of alpha-U due to thermal stresses caused by its anisotropic thermal expansion using phase-field fracture simulations. In the simulations, the deformation of 3D alpha-U polycrystals with random crystallographic texture is modeled during temperature transients and a phase-field fracture model is used to predict crack nucleation and growth resulting from the anisotropic deformation in each of the grains. When the material is heated from 298 K to 650 K and linear elastic material behavior is assumed, intergranular cracks initiate at approximately 500 K. When the material is cooled from 673 K to 300 K, cracks initiate at an even smaller change in temperature. Including isotropic yield surface plasticity slows the crack propagation because it decreases the stress. 2D simulations in which one crystallographic coefficient of thermal expansion is set to zero indicates that the expansion behavior in the [010] direction is the primary cause of the fracture. Subsequently, a preliminary analysis of the impact of irradiation growth and thermal expansion anisotropy on the development of internal stress and fracture in metallic reactor fuel is also carried out for a simple alpha-U fuel. In this case, the simulations combine single-crystal eigenstrains from irradiation growth and thermal expansion with a phase field fracture model in a 3D domain of polycrystalline alpha-U during early reactor operation, including an initial power ramp before steady operation. It is found that initially thermal expansion strains are counteracted by irradiation growth strains, resulting in a lower stress state than would be observed by either mechanism separately. However, as burnup continues, irradiation growth strains come to dominate the polycrystal deformation, resulting in fracture. The amount of fracture occurred near the center of the fuel slug is found to be lower, even though the temperature was higher.