TY - JOUR
T1 - Tritium related studies within the JET fusion technology work programme
AU - Rosanvallon, S.
AU - Bekris, N.
AU - Braet, J.
AU - Coad, P.
AU - Counsell, G.
AU - Cristescu, I.
AU - Grisolia, C.
AU - Le Guern, F.
AU - Ionita, G.
AU - Likonen, Jari
AU - Perevezenstev, A.
AU - Piazza, G.
AU - Poletiko, C.
AU - Rubel, M.
AU - Weulersse, J.M.
AU - Williams, J.
AU - JET-EFDA Contributors,
PY - 2005
Y1 - 2005
N2 - The
JET Fusion Technology (FT) work programme was launched in 2000, in the
frame of the European Fusion Development Agreement, to address issues
related to JET and ITER. In particular, there are four topics related to
tritium being investigated. Based on the experience gained on the
existing tokamaks, first calculations indicate that in-vessel tritium
retention could represent a burden for ITER operation. Therefore
erosion/deposition studies are being performed in order to better
understand the layer co-deposition and tritium retention processes in
tokamaks. Moreover, testing of in-situ detritiation processes, in
particular laser and flash lamp treatments, should assess detritiation
techniques for in-vessel components in the ITER-relevant JET
configuration. To
reduce the constraints on waste disposal, dedicated procedures are
being developed for detritiation of metals, graphite, carbon-fibre
composites, process and housekeeping waste. During the operational and
decommissioning phases of a fusion reactor, many processes will produce
tritiated water. Key components for an ITER relevant water detritiation
facility are being studied experimentally with the aim of producing a
complete design that could be implemented and tested at JET. This paper
describes these topics of the FT-programme, the strategy developed and
the results obtained so far.
AB - The
JET Fusion Technology (FT) work programme was launched in 2000, in the
frame of the European Fusion Development Agreement, to address issues
related to JET and ITER. In particular, there are four topics related to
tritium being investigated. Based on the experience gained on the
existing tokamaks, first calculations indicate that in-vessel tritium
retention could represent a burden for ITER operation. Therefore
erosion/deposition studies are being performed in order to better
understand the layer co-deposition and tritium retention processes in
tokamaks. Moreover, testing of in-situ detritiation processes, in
particular laser and flash lamp treatments, should assess detritiation
techniques for in-vessel components in the ITER-relevant JET
configuration. To
reduce the constraints on waste disposal, dedicated procedures are
being developed for detritiation of metals, graphite, carbon-fibre
composites, process and housekeeping waste. During the operational and
decommissioning phases of a fusion reactor, many processes will produce
tritiated water. Key components for an ITER relevant water detritiation
facility are being studied experimentally with the aim of producing a
complete design that could be implemented and tested at JET. This paper
describes these topics of the FT-programme, the strategy developed and
the results obtained so far.
U2 - 10.13182/FST05-A925
DO - 10.13182/FST05-A925
M3 - Article
SN - 1536-1055
VL - 48
SP - 268
EP - 273
JO - Fusion Science and Technology
JF - Fusion Science and Technology
IS - 1
ER -