TY - JOUR
T1 - A novel approach to determine the local burnup in irradiated fuels using Atom Probe Tomography (APT)
AU - Bachhav, Mukesh
AU - Gan, Jian
AU - Keiser, Dennis D.
AU - Giglio, Jeffrey J.
AU - Jadernas, Daniel
AU - Leenaers, Ann
AU - Van den Berghe, Sven
N1 - Score=10
PY - 2020/1/1
Y1 - 2020/1/1
N2 - A novel approach is presented to determine the local burnup in irradiated fuels using isotopic quantification
obtained by Atom Probe Tomography (APT). Considering the volume of sample used (<100 mm3)
for APT experiments using the lift-out process in a scanning electron microscope equipped with a
Focused Ion Beam (FIB), the presented method determines the local burnup from a nuclear fuel, where a
minimal amount of waste is produced. In this work, three samples were analyzed with different burnup
conditions: as received low enriched 19.8% U-235, intermediate burnup (~52% U-235 fissioned) and high
burnup (~69% U-235 fissioned) UeMo fuel. APT is used to quantify the isotopes of 235U, 236U, 238U, 239Pu
and 237Np for burnup calculation in the irradiated metallic Ue7Mo dispersion fuel. The equation used to
estimate the burnup of fuels is derived by considering that the initial counts of U is equal to the sum of
remaining atoms of U isotopes and all the U reactions undergone during irradiation. This method provides
U enrichment and local burnup with an unprecedented high spatial resolution based on quantification
of isotopic ratios of U.
AB - A novel approach is presented to determine the local burnup in irradiated fuels using isotopic quantification
obtained by Atom Probe Tomography (APT). Considering the volume of sample used (<100 mm3)
for APT experiments using the lift-out process in a scanning electron microscope equipped with a
Focused Ion Beam (FIB), the presented method determines the local burnup from a nuclear fuel, where a
minimal amount of waste is produced. In this work, three samples were analyzed with different burnup
conditions: as received low enriched 19.8% U-235, intermediate burnup (~52% U-235 fissioned) and high
burnup (~69% U-235 fissioned) UeMo fuel. APT is used to quantify the isotopes of 235U, 236U, 238U, 239Pu
and 237Np for burnup calculation in the irradiated metallic Ue7Mo dispersion fuel. The equation used to
estimate the burnup of fuels is derived by considering that the initial counts of U is equal to the sum of
remaining atoms of U isotopes and all the U reactions undergone during irradiation. This method provides
U enrichment and local burnup with an unprecedented high spatial resolution based on quantification
of isotopic ratios of U.
KW - APT
KW - irradiated fuel
UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/36037450
U2 - 10.1016/j.jnucmat.2019.151853
DO - 10.1016/j.jnucmat.2019.151853
M3 - Article
SN - 0022-3115
VL - 528
SP - 1
EP - 9
JO - Journal of Nuclear Materials
JF - Journal of Nuclear Materials
IS - 151853
M1 - 151853
ER -