Evaluation of missing pellet surface geometry on cladding stress distribution and magnitude
Abstract Missing pellet surface (MPS) defects are local geometric defects in nuclear fuel pellets that result from pellet mishandling or manufacturing. The presence of MPS defects can cause significant clad stress concentrations that can lead to through-wall cladding failure for certain combinations of fuel burnup, and reactor power level or power change. Consequently, the impact of MPS defects has limited the rate of power increase, or ramp rate, in both pressurized and boiling water reactors (PWRs and BWRs, respectively). Improved three-dimensional (3-D) fuel performance models of MPS defect geometry can provide better understanding of the probability for pellet clad mechanical interaction (PCMI), and correspondingly the available margin against cladding failure by stress corrosion cracking (SCC). The Consortium of Advanced Simulations of Light Water Reactors (CASL) has been developing the Bison-CASL fuel performance code to consider the inherently multi-physics and multi-dimensional mechanisms that control fuel behavior, including cladding stress concentrations resulting from MPS defects. This paper evaluates the cladding hoop stress distributions as a function of MPS defect geometry with discrete pellet radial cracks for a set of typical operating conditions in a PWR fuel rod. The results provide a first step toward a probabilistic approach to assess cladding failure during power maneuvers. This analysis provides insight into how varying pellet defect geometries affect the distribution of the cladding stress, as well as the temperature distributions within the fuel and clad; and are used to develop stress concentration factors for comparing 2-D and 3-D models.
Finite element simulation of Pellet-Cladding Interaction (PCI) in nuclear fuel rods
Abstract This paper deals with the use of the finite element method to simulate locally the phenomenon of Pellet-Cladding Interaction (PCI) in nuclear fuel rods. The main physical processes involved in PCI are first presented. Then, the problem of PCI modeling is explained, putting the emphasis on finite element code TOUTATIS used for present analyses. Several results of two-dimensional FE analyses carried out by AREVA NP are shown. It appears that fuel pellet cracking during power ramp test is due to the friction between fuel pellet and cladding. The influence of fuel fracture properties is also investigated, focusing on the consequences on clad loading. These results are used to improve the modeling of PCI in AREVA NP’s future global fuel rod performance code COPERNIC3.
The thermal expansion of solid UO2 and (U, Pu) mixed oxides — a review and recommendations
Abstract Published experimental data relating to the thermal expansion, between 291 K and the melting temperature, of stoichiometric UO 2 , UO 2 + x and mixed (U, Pu) oxides, with compositions of relevance to fast-reactor fuel, is reviewed. For stoichiometric UO 2 it is recommended that L = L 273 (9.9734 × 10 −1 + 9.802 × 10 −6 T − 2.705 × 10 −10 T 2 + 4.391 × 10 −13 T 3 ) for 273 K ≤ T ≤ 923 K and L = L 273 (9.9672 × 10 −1 + 1.179 × 10 −5 T − 2.429 × 10 −9 T 2 + 1.219 × 10 −12 T 3 ) for 923 K ≤ T ≤ 3120 K , where L , L 273 are lengths at temperatures of T (K) and 273 K. This recommendation may also be employed to describe the thermal expansion in the solid phase of UO 2+x , with x values in the ranges 0–0.13 and 0.23–0.25, and also stoichiometric mixed (U, Pu) oxides with Pu contents up to ~ 30%. In the case of hypostoichiometric mixed oxides, the thermal expansion should be taken as the corresponding UO 2.00 value, multiplied by a factor [1 + 3.9 x ], where x is the deviation from stoichiometry. Errors and uncertainties in these recommendations are discussed.