An improved structural mechanics model for the FRAPCON nuclear fuel performance code
Author(s)Mieloszyk, Alexander James
Massachusetts Institute of Technology. Dept. of Nuclear Science and Engineering.
Mujid S. Kazimi.
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In order to provide improved predictions of Pellet Cladding Mechanical Interaction (PCMI) for the FRAPCON nuclear fuel performance code, a new model, the FRAPCON Radial-Axial Soft Pellet (FRASP) model, was developed. This new model uses 1.5D structural mechanics to represent both the fuel pellet and cladding along with their interaction via interfacial forces. The fuel pellet and cladding are modeled as concentric annular cylinders using similar governing equations with slight differences to allow for cracking of the semi-brittle fuel matrix and plastic behavior in a ductile cladding. By accounting for the structural mechanics of the fuel pellet, FRASP allows for stress-induced deformations which were previously unattainable with the rigid pellet model used by FRAPCON. Because of the significant differences between FRAPCON's previous mechanical model, FRACASI, and FRASP, simply replacing the treatment of PCMI within the code was not a viable option. This led to a complete replacement of FRACAS-I and all associated fuel rod structural calculations. Feedback effects are likely to result from such a major change due to the complexity of nuclear fuel simulation. The potential for these feedback effects dictated a preliminary validation of FRASP against FRACAS-I for typical case. This evaluation was not limited to the investigation of mechanical parameters, but covered a wide variety of predicted parameters by the new and unaltered versions of FRAPCON. The differences which were found in this validation were limited in nature and easily attributable to the differing assumptions of FRASP and FRACAS-I. The newly developed mechanical model was used with the improved fuel behavior models of FRAPCON-EP (Enhanced Performance) to assess the mechanical behavior of fuel rods with a composite silicon carbide (SiC) cladding under Pressurized Water Reactor (PWR) conditions. The fuel rod designs were selected to match previously chosen values for both solid and annular fuel pellets under current and uprated power conditions. Unlike FRACAS-I, which is hindered by the rigid pellet model, FRASP was able to successfully analyze PCMI behavior with the more rigid SiC, even though "hard contact" of the fuel and cladding was encountered. Simulations using the improved models showed that the SiC clad fuel rods may not provide adequate safety margins at the desired burnup, or simply fail to achieve their desired final burnup. Previous analyses which relied on FRAPCON-3.3 may have been overly optimistic in this regard. The new, more conservative predictions are largely due to FRASP's treatment of the inner radius of the annular fuel pellets, which was assumed not to change in previous versions of FRAPCON. These new findings suggest that SiC fuel rod general design and operation require further optimization.
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012.Cataloged from PDF version of thesis.Includes bibliographical references (p. 149-152).
DepartmentMassachusetts Institute of Technology. Dept. of Nuclear Science and Engineering.
Massachusetts Institute of Technology
Nuclear Science and Engineering.