THE MECHANICAL RESPONSE OF HIGH-BURNUP 17 × 17 PWR FUEL RODS UNDER BENDING

As a part of the US Department of Energy's Office of Nuclear Energy High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory has performed four-point bend tests on a series of 17 × 17 pressurized water reactor (PWR) fuel rods that were discharged with rod average burnups between 50 and 60 GWd/MTU. The rods include four different kinds of fuel cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Six rods of each type (combining Zirc-4 and LT Zirc-4 as one type) were tested to failure at room temperature (RT) and 200 °C. Three of each type experienced a full-length heat treatment (FHT) cycle from RT to 400 °C to RT with an 8 h soak at 400 °C to simulate the expected limiting vacuum drying condition, which potentially induces hydride reorientation and potentially causes some annealing of irradiation damage in the cladding. Overall, the fuel rods show capacity to deflect in bending and exhibit nonlinear deformation before failing. The calculated failure strains range from 1.5 to 3.8%. The simulated vacuum drying heat treatment had the largest effect on fuel rods with ZIRLO and M5 cladding. In both cases, the heat treatment increased the plastic strain at failure, suggesting that some annealing of irradiation damage is occurring. The finite element modeling of the four-point bend test assumes that the pellet is a collection of pellet wedge pieces. The results for the Zirc-4 rods suggest that the pellets are bonded to the cladding and that friction controls the relative motion of the pellet wedge pieces.