TY - JOUR
T1 - Efficiency of thermal outgassing for tritium retention measurement and removal in ITER
AU - De Temmerman, G.
AU - Baldwin, M.J.
AU - Anthoine, D.
AU - Heinola, K.
AU - Jan, A.
AU - Jepu, I.
AU - Likonen, Jari
AU - Lungu, C.P.
AU - Porosnicu, C.
AU - Pitts, R.A.
PY - 2017/8/1
Y1 - 2017/8/1
N2 - As a licensed nuclear facility, ITER must limit the
in-vessel tritium (T) retention to reduce the risks of
potential release during accidents, the inventory limit
being set at 1. kg. Simulations and extrapolations from
existing experiments indicate that T-retention in ITER
will mainly be driven by co-deposition with beryllium
(Be) eroded from the first wall, with co-deposits forming
mainly in the divertor region but also possibly on the
first wall itself. A pulsed Laser-Induced Desorption
(LID) system, called Tritium Monitor, is being designed
to locally measure the T-retention in co-deposits forming
on the inner divertor baffle of ITER. Regarding tritium
removal, the baseline strategy is to perform baking of
the plasma-facing components, at 513. K for the FW and
623. K for the divertor. Both baking and laser desorption
rely on the thermal desorption of tritium from the
surface, the efficiency of which remains unclear for
thick (and possibly impure) co-deposits. This
contribution reports on the results of TMAP7 studies of
this efficiency for ITER-relevant deposits.
AB - As a licensed nuclear facility, ITER must limit the
in-vessel tritium (T) retention to reduce the risks of
potential release during accidents, the inventory limit
being set at 1. kg. Simulations and extrapolations from
existing experiments indicate that T-retention in ITER
will mainly be driven by co-deposition with beryllium
(Be) eroded from the first wall, with co-deposits forming
mainly in the divertor region but also possibly on the
first wall itself. A pulsed Laser-Induced Desorption
(LID) system, called Tritium Monitor, is being designed
to locally measure the T-retention in co-deposits forming
on the inner divertor baffle of ITER. Regarding tritium
removal, the baseline strategy is to perform baking of
the plasma-facing components, at 513. K for the FW and
623. K for the divertor. Both baking and laser desorption
rely on the thermal desorption of tritium from the
surface, the efficiency of which remains unclear for
thick (and possibly impure) co-deposits. This
contribution reports on the results of TMAP7 studies of
this efficiency for ITER-relevant deposits.
KW - Tritium retention
KW - ITER
KW - divertor
KW - outgassing
UR - http://www.scopus.com/inward/record.url?scp=85006804232&partnerID=8YFLogxK
U2 - 10.1016/j.nme.2016.10.016
DO - 10.1016/j.nme.2016.10.016
M3 - Article
SN - 2352-1791
VL - 12
SP - 267
EP - 272
JO - Nuclear Materials and Energy
JF - Nuclear Materials and Energy
ER -
