Modelling the role of pellet crack motion in the ( r -θ) plane upon pellet-clad interaction in advanced gas reactor fuel

Abstract

A finite element model of pellet fragment relocation in the r-θ plane of advanced gas-cooled reactor (AGR) fuel is presented under conditions of both ‘hard’ and ‘soft’ pellet-clad interaction. The model was able to predict the additional radial displacement of fuel fragments towards the cladding as well as the stress concentration on the inner surface resulting from the azimuthal motion of pellet fragments. The model was subjected to a severe ramp in power from both full power and after a period of reduced power operation; in the former, the maximum hoop stress in the cladding was found to be increased by a factor of 1.6 as a result of modelling the pellet fragment motion. The pellet-clad interaction was found to be relatively insensitive to the number of radial pellet crack. However, it was very sensitive to both the coefficient of friction used between the clad and pellet fragments and power ramp duration

The impact of azimuthally asymmetric carbon deposition upon pellet-clad mechanical interaction in advanced gas reactor fuel

“…Secondly, AGR fuel pellets contain a central bore. The opening and closure of pellet cracks at the bore and displacement of the bore itself form a significant part of the PCMI response of AGR fuel; this is demonstrated in Figure 5(b) and [25].…”

“…Heat transfer across the radial pellet cracks is based upon the URGAP model [30], [31]. It is described in more detail in [25]. As per our previous work [25], a coefficient of friction of 0.8 was applied between all surfaces.…”

“…It is described in more detail in [25]. As per our previous work [25], a coefficient of friction of 0.8 was applied between all surfaces. In this paper, we report the results obtained for fuel element four with no clad-fuel bonding.…”

Preliminary modelling of crack nucleation and propagation in SiC/SiC accident-tolerant fuel during routine operational transients using peridynamics

“…Clearly modelling a material using a Poisson's ratio greater than that of the experimental value will have some impact upon the stress state predicted. Given that the largest stresses in nuclear fuel cladding are in the hoop direction [39], using an overly large Poisson's ratio would result in the incorrect prediction of stresses in the radial direction. However, given that the inner and outer surfaces are effectively free surfaces and pellet-clad contact is not being considered, the impact of this is likely to be minor.…”

“…Prior to contact between the cladding and fuel pellets, it could be argued that the cladding will be in a state closer to that of generalised plane strain. Following contact, which is likely to be later in fuel life for SiC/SiC cladding than traditional zirconium-based alloys due to a lower creep rate and greater swelling, the stress state in the cladding is likely to become tri-axial and concentrated to regions ahead of pellet cracks [39].…”