W. Daenner

Next European Torus in-vessel components

The Next European Torus (NET) in-vessel components consisting of first wall, divertor, and blanket/shield are subject to severe nuclear and thermal radiation. Their reliability and maintainability are crucial to the success of the NET mission. The requirements, options, preliminary design solution, and materials considerations for these components are described.

Status of fabrication development for plasma facing components in the EU

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.

Engineering design of the ITER blanket and relevant research and development results

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.

ITER primary first wall mock-up fabrication and testing for Be/Cu alloy joining development☆

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.

Progress on Design and R&D of ITER FW/Blanket

Abstract 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.

Manufacture of an ITER primary wall module shield prototype by powder hipping

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.

Design and R&D progress of blanket attachments

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.

Neutronics Shield Experiment for ITER at the Frascati Neutron Generator FNG

Publisher Summary 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.

Design and manufacturing of the ITER limiter

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.

Predesign and Feasibility Studies of the NET Shielding Blanket Segments

This paper summarizes the main conclusions of a study contracted by the NET Team to the CEA, with Framatome as subcontractor, to perform predesign and feasibility studies of the NET shielding blanket segments. A first phase of the study has been devoted to a comparative assessment of candidate blanket concepts, according to a list of criteria including the main design features, the operating performances and the relevant manufacturing process. A second phase of the study is currently in progress, with the dual objective of producing a detailed design of the selected concept for the updated NET configuration, and proposing a development programme for the qualification of the specific manufacturing methods and the verification of the target performances.