TY - JOUR
T1 - Modelling the role of pellet crack motion in the (r-θ) plane upon pellet-clad interaction in advanced gas reactor fuel
AU - Haynes, T. A.
AU - Ball, J. A.
AU - Wenman, M. R.
PY - 2017/4/1
Y1 - 2017/4/1
N2 - 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.
AB - 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.
UR - http://www.scopus.com/inward/record.url?eid=2-s2.0-85013333306&partnerID=MN8TOARS
U2 - 10.1016/j.nucengdes.2017.02.007
DO - 10.1016/j.nucengdes.2017.02.007
M3 - Article
VL - 314
SP - 271
EP - 284
JO - Nuclear Engineering and Design
JF - Nuclear Engineering and Design
SN - 0029-5493
ER -