Prime pagine RA2010FUS:Copia di Layout 1 - ENEA - Fusione

technology programme (cont’d.)
progress report
2010
061
The ITER blanket and shield system is the
innermost part of the reactor; it is directly
exposed to the plasma and provides the main
thermal and nuclear shielding to the vacuum
vessel and external reactor components. Its
concept is based on a modular configuration with
blanket modules consisting of water–cooled
austenitic stainless steel shield blocks and
separable first–wall (FW) panels, mechanically
attached to the shield blocks. The blanket
modules have typical dimensions of
1m× 1.5 m × 0.5 m and are mechanically
attached to the VV. The water coolant is supplied
to the modules by a set of inlet and outlet
manifolds attached to the inner wall of the VV.
Figure 3.10 – Proposed design and detail of the FEM of the
inboard ITER shielding blanket NMC including the inner vessel
shell, the modules #6 and #7 and part of the single pipe
assembly
This work was performed to evaluate the eddy currents and the forces induced by fast variations of the
magnetic fields due to plasma off–normal events in the NMC where the original welded structure supplying
the cooling water to each module has been replaced by single pipes. This evaluation was needed to address
some concerns raised during the ITER design review about the capability of repairing the components
remotely and the high operating stresses and possible water leakages that might be produced at some locations
on the manifolds.
The FEMs developed for these analyses (fig. 3.10) include a 10° toroidal sector of the ITER machine with the
double shell vessel, the divertor, the blanket modules and a quite detailed model of the pipe manifold where
each pipe is insulated from the other pipes, the blanket and the vessel except in the points where the pipes are
bundled together and attached to the vessel.
The 3D EM load analyses, performed by using the ANSYS code, allowed the time evolution of the current
and bending force to be evaluated per unit length in the whole bundle as well as the current sharing between
the whole pipe bundle and the vessel, for several plasma off–normal events. The major disruption type II with
36 ms linear current quench was proved to be the most severe event as far as the force per unit length on a
single pipe, even if the total force on the manifold is larger during the thermal quench, because the eddy
current is not shared among the pipes but is concentrated in the first part of a single pipe at each manifold
section. The EM loads induced on the NMC are lower as compared to the reference concept, but this
advantage has to be confirmed in terms of mechanical stresses due to the significant differences among the
mechanical structures. In any case, the large resulting EM loads must be carefully considered in the detailed
design of the manifold components and their supporting systems.
Development of method for highly tritiated water handling in ITER tritium plant
A process based on a combination of permeator catalyst (PERMCAT) (Pd–based membrane reactor) and
vapor phase catalytic exchange (VPCE) has been studied for processing higly tritiated water (HTW) (ITER
contract ITER–CT–09–4300000087). The simulation of the PERMCAT and VPCE systems has been carried
out and the process flow diagram has been prepared [3.3]. As a main advantage, the process being proposed
permits to combine the PERMCAT and VPCE in several ways accordingly to the characteristics of the
different HTW streams to be processed. Furthermore, both PERMCAT and VPCE can use the mixture of
hydrogen isotopes produced by the electrolyzers of the water detritiation system as sweep gas, thus reducing
the impact on the isotopic separation system and avoiding the generation of secondary wastes.
Training activities
ENEA Frascati laboratories have had in charge a Trainee (EURATOM Research Training Network
"Preparing the ITER Fuel Cycle" – Contract No. 042862 (FU 06)) to be prepared in the area of the fuel cycle
of ITER. All the activities have been completed by achieving the milestones of the task: study of cold–rolling
and diffusion welding techniques, long–term testing of Pd/Ag membranes, participation to the CAPER R&D
program, tritium confinement study [3.4], analysis of tritium release from the neutral beam injector [3.5] and
inactive tests for new PERMCAT prototypes [3.6].