Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for a significant share of the overall procurement of internal components inside the vacuum vessel, the largest ones ...being as follows: (1) 48% of the ITER blanket first wall (FW), (2) the divertor cassette bodies and (3) the divertor inner vertical target. Procurement strategies have been implemented by the In-Vessel Project Team at F4E aiming at mitigating technical and commercial risks for the procurement of these ITER internal components, promoting as far as possible competition among industrial partners. These procurement strategies have been supported by extensive research and development (R&D) work programmes, implemented over more than 15 years in Europe, to develop various fabrication technologies especially for plasma facing components. They include in particular the manufacture and testing of small-scale mock-ups, medium-scale mock-ups and full-scale prototypes of blanket FW panels and divertor inner vertical targets. In these R&D work programmes, significant efforts have been devoted to the development of reliable armour-to-heat sink materials joining techniques. More recently, full-scale prototypes of the divertor cassette body have also been fabricated.
This paper presents the procurement strategies implemented by F4E for the supply of the European contribution to the procurement of the above ITER internal components, together with the main outcome of the R&D work programmes and the state of progress of the respective qualification programmes.
► The WEST programme is a unique opportunity to experience the industrial scale manufacture of tungsten plasma-facing components similar to the ITER divertor ones. ► In Tore Supra, it will bring ...important know how for actively cooled W divertor operation. ► This can be done by a reasonable modification of the Tore Supra tokamak. ► A fast implementation of the project would make this information available in due time. ► This allows a significant contribution to the W ITER divertor risk minimization in its manufacturing and operation phase.
The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.
Understanding heat flux deposition processes is essential for the design of the plasma facing component allowing reliable high power steady state plasma operations. Misalignments up to δ=0.2mm ...between two adjacent CFC tiles have been reported on the Toroidal Pump Limiter of Tore Supra. Heat flux impinging the top and leading edge of the protruding tile are characterized with both IR thermographic system and numerical modelling using 2D particle-in-cell simulations that accounts for the Larmor radius smoothing effect of incident ions. Numerical heat loads are coupled with a 2D thermal model of the tile and with a specific sensor correction to simulate spatial-resolution related effects (necessary here since the tile misalignment is smaller than the spatial resolution of the IR system). In the experiment depicted here, with a misalignment smaller than an ion gyro-radius, the Larmor radius smoothing effect is maximum and overheating of the leading edge is reduced by a factor of two.
•Mechanical deformation of CuCrZr in case a thermal barrier layer has been formed due to impurity content in the cooling water.•Crack formation at the W/Cu interface starting at the block ...edge.•Porosity formation in the pure Cu interlayer.•Microstructural changes in tungsten down to the W/Cu interface, which indicates also high temperatures for the pure Cu interlayer.•Macrocrack formation in tungsten which is assumed to be ductile at the initiation point and brittle when proceeding toward the cooling tube.
High heat flux tested small-scale tungsten monoblock mock-ups (5000 cycles at 10MW/m2 and up to 1000 cycles at 20MW/m2) manufactured by Plansee and Ansaldo were characterized by metallographic means. Therein, the macrocrack formation and propagation in tungsten, its recrystallization behavior and the surface response to different heat load facilities were investigated. Furthermore, debonding at the W/Cu interface, void formation in the soft copper interlayer and microcrack formation at the inner surface of the CuCrZr cooling tube were found.
•JADA has demonstrated the feasibility of manufacturing the full-W plasma-facing units (W-PFU).•The surface profiles of the W monoblocks of the W-PFU prototypes on the test frame to mimic the support ...structure of the ITER OVT were examined by using an optical three-dimensional measurement system. The results show the most W monoblock surface in the target part locates within+0.25mm from the CAD data.•The strict profile control with the profile tolerance of ±0.3mm is imposed on the OVT to prevent the leading edges of the W monoblocks from over-heating.•The present full-scale prototyping demonstrates to satisfy this requirement on the surface profile.•It can be concluded that the technical maturities of JADA and its suppliers are as high as to start series manufacturing the ITER divertor components.
Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20MW/m2 for 10s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.
•Manufacturing technologies for the ITER internal components have been developed.•The Blanket System successfully went through its Final Design Review in April 2013.•The decision to start operation ...with a Divertor with a full-W armour has been taken.
The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R&D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.
The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which ...include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor.
F4E undertook the qualification of so-called “additional suppliers” in order to enhance competition among the potential bidders and to secure the procurement of the ITER Divertor Inner Vertical ...Target. In order to assess the performances of W armoured plasma facing components under the conditions expected in the divertor target strike point region, a total of 36 W monoblock mock-ups were manufactured by Atmostat (F) by diffusion bonding, and by CNIM (F) and Research Instruments GmbH (D) by brazing, from which fifteen mock-ups were High Heat Flux (HHF) tested at IDTF (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facility.
The basic HHF testing program foresaw the performance of 5000 cycles at 10 MW/m2and 300+700 cycles at 20 MW/m2. The test results showed some significant improvements, in particular on the issues of W monoblocks macro-cracking and heat sink thermo-mechanical fatigue performances, identified during previous HHF tests campaigns. Some enhanced HHF thermal fatigue testing and critical heat flux experiments were also performed.
Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the procurement of the divertor inner vertical target (IVT) and the cassette body (CB). Both IVT and CB have been ...supported by extensive Research and Development (R&D) work programmes, implemented over more than 15 years in Europe, to validate design, develop various fabrication technologies, assess thermo-mechanical and thermohydraulics performances and RH installation and maintainability.
This paper presents the work performed and the technical challenges associated to the fabrication of the CB prototypes, the results of the prototype qualification as well as the lessons learned.
Historically developed for cooling of klystron electronic tubes, metallic hypervapotron® prototypes with different width were manufactured for high heat flux plasma facing components (PFC) ...applications. They were critical heat flux (CHF) tested on the European 200 kW electron beam facility (FE200), the 54 measured values have shown their good performances—up to 25–30 MW/m
2 at low axial velocities (2–6 m/s)—interesting for ITER divertor dome and vertical target design. An important conclusion is that CHF decreases when the width increases.
