Medium-scale Mock-ups Research Articles

EU contribution to the procurement of blanket first wall and divertor components for ITER

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.

EU contribution to the procurement of the ITER blanket first wall

Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel.This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

The heat removal capability of actively cooled plasma-facing components for the ITER divertor

Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m−2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m−2 for the CFC-armoured tiles and 15 MW m−2 for the W-armoured tiles, respectively.

Testing of high heat flux components manufactured by ENEA for ITER divertor

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested. A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process. The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France). The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m 2, 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m 2, 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m 2 Critical Heat Flux was measured at relevant thermal–hydraulics conditions at the end of the testing campaign. This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up. These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).

Optimization of armour geometry and bonding techniques for tungsten-armoured high heat flux components

Joining of tungsten with copper-based cooling structure and armour geometry optimization are the major aspects in development of the tungsten-armoured plasma facing components (PFC). Fabrication techniques and high heat flux (HHF) tests of tungsten-armoured components have to reflect different PFC designs and acceptable manufacturing cost. The authors present the recent results of tungsten-armoured mock-ups development based on manufacturing and HHF tests. Two aspects were investigated—selection of armour geometry and examination of tungsten–copper bonding techniques. Brazing and casting tungsten–copper bonding techniques were used in small mock-ups. The mock-ups with armour tiles (20×5×10, 10×10×10, 20×20×10, 27×27×10) mm 3 in dimensions were tested by cyclic heat fluxes in the range of (5–20) MW/m 2, the number of thermal cycles varied from hundreds to several thousands for each mock-up. The results of the tests show the applicability of different geometry and different bonding technique to corresponding heat loading. A medium-scale mock-up 0.6-m in length was manufactured and tested. HHF tests of the medium-scale mock-up have demonstrated the applicability of the applied bonding techniques and armour geometry for full-scale PFC's manufacturing.

Manufacture and first wall joining for an ITER primary wall module prototype: results of a medium scale mock-up manufacturing

In the frame of the Blanket development programme for ITER, the design envisaged a blanket-shield constructed from modules. Primary Wall Modules consisted of a water-cooled austenitic stainless steel (S.S.) Shield Block and a First Wall, as an integral part of it. The Primary First Wall uses a DS-copper alloy as the heat sink material bonded to the shield block and a protection material such as Beryllium, to cope with the plasma/wall interaction. The module was designed to sustain a peak heat flux of 0.5 MW/m 2. A manufacturing program was implemented to demonstrate the feasibility for joining the Primary First Wall onto the Shield, which included the manufacture of small scale and medium scale mock-ups, before the manufacture of a prototype. The aim of the paper is to present the different steps and results of a Medium Scale Module manufacturing.

ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield bodyto which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly testsof a blanket module have demonstrated the basic feasibility of its installation and removal.

Progress in the design of the HCPB test blanket module mock-ups for the HE-FUS3 facility

One of the ENEA contributions in the activity planned for the European Fusion Technology Programme for 1997–1998 is the design and testing of helium cooled pebble bed blanket (HCPB) mock-ups for the DEMO fusion reactor. The reference medium scale mock-up, called HEXCALIBER (HE-FUS3 experimental cassette of lithium beryllium pebble beds), will reproduce a portion of the HCPB module with the L 4SiO 4 ceramic breeder and the beryllium neutron multiplier, both in the form of pebble beds heated by flat heaters. The feasibility of the final design and the manufacturing solutions will be first evaluated by testing a reduced portion of the mock-up called HELICA (HE-FUS3 lithium cassette) filled with L 4SiO 4 breeder pebbles. Aimed at defining the thermal map of the mock-up and the related stresses, some preliminary FEM theoretical calculations were carried-out by ANSYS code. This paper presents the design of two mock-ups, the preliminary calculations, the raw material behaviours and the qualification of the manufacturing procedures. The experimental tests will be carried-out on the HE-FUS3 facility at ENEA Brasimone.

Development of first wall/blanket structure by hot isostatic pressing (HIP) in the JAERI

In the ITER shielding blanket design, first wall coolant tubes of type 316LN stainless steel (SS) are embedded in a heat sink of Al 2O 3 dispersion-strengthened Cu (DSCu), which are joined to a SS shield block. Development of fabrication technology using hot isostatic pressing (HIP) for this structure, including testing, is currently being performed at JAERI. Recent progress in this development is reviewed in this paper. A simultaneous joining of DSCu/DSCu, DSCu/SS and SS/SS, especially at the first wall and the first wall to the shield block, by a solid HIP technique, has been pursued. Following the optimization of HIP conditions, mechanical tests of the HIP joints and fabrication of first wall mock-ups were performed. High heat flux tests of the first wall mock-ups resulted in their sufficient heat removal capability and integrity of the HIP joints. Small- and medium-scale mock-ups of the shielding blanket, integrated with the first wall, were fabricated to examine the applicability of the HIP technique and develop a fabrication route. As a result of these activities, fabrication technologies and the route to be applied for the ITER shielding blanket and sufficient thermo-mechanical performance of the HIP bonded structure were foreseen.