Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20° or 30°, on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

During the preparation of the procurement specifications of ITER for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain more structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inlet–outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R&D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and non-destructive tests for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods.

The problem of joining the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) sectors, considering the tolerance requirements of the blanket attachments, and the time required for TIG welding, continues to stimulate EU R&D into power beam welding techniques which can yield fewer passes, less welding time and lower distortion. The previous work on reduced pressure e-beam welding showed that penetration varied with position, fit-up, distance and pressure and single-pass weld control was deemed to be not reliable enough so the work direction changed to an all-e-beam welding procedure where the root weld is carried out with rest-current-control and the fill passes by wire-fill. In addition, a novel method of increasing the possible single-pass weld thickness for overhead positions is investigated demonstrated and now patented. Another solution may be offered with wire-fill NdYAG laser welding, which has demonstrated useable and stable results and proved improved performance over TIG. Preliminary work has shown even further advantages with the introduction of hybrid MIG/Laser welding.

The ITER blanket is built of small regular modules attached to the vacuum vessel by an insulated mechanical support system, an electrical connector for grounding and a hydraulic connector for the coolant manifolds. An attachment has been developed which satisfies the demanding load conditions and which includes the handling features to enable the removal of the module for repair in the hot cells. Latest improvements are presented with the comprehensive R&D activities used to validate the design.

The International Thermonuclear Experimental Reactor (ITER) tokamak has been designed to have two limiter modules located outboard in horizontal ports. During the start-up and shut-down phases of the Tokamak operating cycle, the limiter defines the plasma boundary and protect the main first wall (FW) and the RF antennas from the direct contact with the plasma. Furthermore during the flat-top phase of a plasma pulse, the limiter has the same functions of a primary wall blanket module, namely: to provide a plasma-compatible surface, to withstand the radiation and charge particle flux, and to provide adequate shield to the structures behind. The paper presents the status of the limiter design with particular attention to the explanation of the reasons behind the several design and technology choices.

This paper presents the main results obtained so far from the development work performed in Europe to define the joining conditions between beryllium (Be) tiles and the dispersion strengthened copper alloy (DS-Cu) heat sink material for the ITER primary first wall (PFW). Two Be/DS-Cu joining techniques were investigated: hot isostatic pressing and furnace brazing. Six PFW mock-ups have been thermal fatigue tested so far. One PFW mock-up with HIPped Be tile was tested at 2.5 MW/m(2) for 1000 cycles without any indication of failure. On two other mock-ups, Be tiles detached at or above 2.7 MW/m(2). Two others were tested at 0.7 MW/m(2) for 13 000 cycles also without any indication of failures. A first PFW mock-up with a furnace brazed Be tile was tested at 1.6 MW/m(2) for 1000 cycles. These results should be compared with the operation conditions of the ITER PFW, namely 0.5 MW/m(2) peak heat flux and off-normal events up to 1.4 MW/m(2). Thermal fatigue testing of other mock-ups is still in progress and the development programme is continuing to further increase the engineering margins while decreasing the fabrication cost of the PFW panels. (C) 2002 The European Commission. Published by Elsevier Science B.V. All rights reserved.

The electromagnetic (EM) load on the first wall (FW) panel during disruptions is reduced by slots penetrating the copper layer and the SS backing plate. The maximum stress in the central beam is within the allowables under the most significant load induced by halo currents. In the recent ITER R&D, full-scale FW panels have been manufactured successfully by hot isostatic pressing (HIP) as the reference method. The shield block cooling scheme consists of front water headers that distribute the coolant in radial channels. The shield block is composed of four flat forged blocks electron-beam (EB) welded together at the rear side. Recently, full-scale shield blocks were fabricated by drilling/machining and plugging/welding of flat forged blocks, and assembled with a FW panel with a central beam. Detailed design has progressed on the blanket attachments. Buckling tests, fatigue tests and dynamic load tests have been performed on the T-alloy flexible support (550 kN). Mechanical and thermal fatigue tests, and electrical tests in a solenoid coil, have been carried out on the electrical connection (280 kA). Feasibility of the blanket sub-components has been demonstrated through the R&D.

This paper summarises the European R&D efforts for the manufacture of shield modules and divertor cassettes for the International Thermonuclear Experimental Reactor (ITER), including their plasma facing components. The various development steps are described as they had to be taken to resolve the fabrication issues, and to keep track with the evolving design requirements and solutions. For all components, the manufacturing feasibility has been demonstrated on prototype scale which puts Europe in the position to start the procurement as soon as the decision about ITER construction is taken. The time period remaining until then is used to optimise the fabrication processes and to develop more cost effective alternatives.

The Shield Prototype manufactured by the European Home Team is representative of the Primary Wall Module No. 11 as designed for the ITER 1998 Design. This module was selected because it had the most complicated shape to fabricate due in particular to its double curvature in poloidal and toroidal directions. It consists of a stainless steel Shield block of about 4 tons, equipped with eight penetration holes through it and with all the features at the rear and side walls required for the module attachment system. All these requirements led to a complex cooling channel arrangement inside the module, making the fabrication by powder Hot Isostatic Pressing (HIPping) very attractive and competitive. This paper describes the main steps of the Shield Prototype manufacture and the results of the non-destructive examination. These results, which shall be confirmed by destructive examination, have shown so far the manufacturing feasibility by powder HIPping of the Shield of ITER Primary Wall Modules.

The ITER modules are connected to the vacuum vessel wall mechanically, electrically and hydraulically in a way suitable for remote handling with access from the frontside only. In order to validate the design and operation of the attachment systems for the 1998 ITER design, EFDA directed three Associations (ENEA, Brasimone, (ÖAW Seibersdorf and CEA, Arceuil) in tasks including process investigations and the design, procurement and testing of three ITER-representative mock-ups. ENEA, Brasimone procured and provides the site for the largest rig, the Blanket Module Carrier, which provides 6-axis movements of a 4 ton module mock-up to simulate the mounting of a typical ITER module to a vertical wall. ÖAW Seibersdorf designed an ITER remote-handling-compatible, water-hydraulically-operated bolting tool, which ENEA, Brasimone procured and used in a test programme. CEA, Arceuil developed and tested a special bore tool using a NdYAG laser for cutting and welding the hydraulic connectors. The overall programme demonstrated the concept viability and highlighted problem areas of the assembly procedure of the blanket module-to-wall attachment. Aspects of the work still require further attention to refine and finally validate the techniques especially with regard to the ITER-FEAT design.

This paper presents the main achievements of the European Home Team (EU HT) test programme of ITER primary wall small scale mock ups. It describes briefly the fabrication method of the mock ups, the test conditions and the main results obtained with high heat flux and thermal fatigue tests of Cu alloy/stainless steel and beryllium/Cu alloy/stainless steel mock ups. The results obtained so far show very good thermal fatigue performance and operation margin of the ITER primary first wall concept.

The design of the ITER blanket is presented together with the related technology which has been developed. The evolution of this component since the beginning of the EDA is explained in relation to the developing understanding of the thermal deformations and of the electromagnetic forces. These loads lead to a system composed of compact modules protecting a continuous support shell called a backplate. The backplate is a stiff double wall construction which conveys the coolant to the modules. The supports of the module are flexible and allow relative thermal expansions. They are connected and disconnected to the backplate by bolts operated through holes in the front face of the module. The coolant connections and the electrical straps located on the back of the modules are reached similarly. The first wall is integral with the module and cooled in series. A research and development program on materials and joining methods defined the construction path which has been tested in prototypes. The main body is built of stainless steel by forging and drilling or powder hot isostatic pressing (HIP), depending on the complexity of the shape. The first wall includes a dispersion strengthened copper heat sink which is hot isostatic pressed onto the steel body. Beryllium is the basic plasma facing material and is attached by HIP to the copper. Prototypes of the module attachment have been built and are under integrated tests.

Significant improvements of the ITER blanket were possible with the better definition of the thermal and electromagnetic effects. A supporting double wall backplate with reinforcement about the ports has been defined. The module has been simplified and the intrusion of the connections reduced. The attachment has been assessed by system analyses. The cooling system is configured to simplify the leak detection

A first primary wall small scale mock up with beryllium as the armor material was manufactured by hot isostatic pressing (HIPing) by CEA and tested at the electron beam test facility JUDITH. The mock up consisted of a 9.4 mm thick beryllium armor joint onto a 20 mm thick dispersion strengthened copper (DS-Cu) alloy plate. Type 316L Stainless Steel tubes, 10/12 mm in diameter were embedded in the DS-Cu, which was subsequently joined onto a 30 mm thick 316LN stainless steel plate. The mock up was tested under a surface heat flux of 2.5 MW m−2 for 100 preparatory cycles and 900 cycles of 30 s heating and 30 s cooling time. At the end of the 1000 cycles, the surface and the Be/DS-Cu joint of the mock up did not show any damage due to the fatigue test.

This paper summarises the main results obtained so far in the frame of the EU Home Team test programme of ITER primary wall small scale mock ups. It describes briefly the fabrication method of the mock ups, the test conditions and the main results obtained with high heat flux and thermal fatigue tests of Cu alloy/stainless steel and beryllium/Cu alloy/stainless steel mock ups. The results obtained so far show good thermal fatigue performance and operation margin of the ITER primary first wall concept.

Radiation loads are critical issues for the International Thermonuclear Experimental Reactor (ITER) shielding design. Limits on radiation induced displacement damage, insulator dose, and He production are imposed for all components assumed to be permanent or semi-permanent, that is, the toroidal field coil, the vacuum vessel, the blanket backplate, and the coolant manifold. Limits on the nuclear heating of the toroidal field magnet are also imposed. Radiation loads to ITER components are routinely calculated by using current state-of-the-art neutron, photon transport codes, and nuclear data, such as the Monte Carlo code and the fusion evaluated nuclear data library cross-section data file. A neutronics bulk shield experiment has been initiated by Frascati Neutron Generator with the main objective to validate the shielding performance of the ITER predicted by calculations. This chapter discusses the results of this experiment, that provide experimental data on the relevant nuclear quantities on a mock-up of the ITER inboard shield where the critical loads are found.