P. Lorenzetto

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.

A European proposal for an ITER water-cooled solid breeder blanket

Abstract The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m−2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m−2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3–5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200°C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using blanket test modules during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved.

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.

Manufacture of two primary first wall panel prototypes with Beryllium armor for ITER

The aim of this paper is to present the results of a manufacturing program that was implemented to demonstrate the feasibility for manufacturing the primary first wall panels, including a part of R&D work concerning the joining of Beryllium plates onto a Glidcop heat sink by HIPing or brazing.

Evaluation of mock-ups before manufacture of a shield block prototype by powder HIP

Abstract Hot isostatic pressing (HIP) of powder is proposed as a possible process for manufacturing primary wall blanket modules for the International Thermonuclear Experimental Reactor (ITER). The method is cost-effective and offers very good design flexibility. It also provides required, uniform material properties, and the material is easy to check with ultrasonic techniques. Bodycote Powdermet conducted a development programme as a part of a NET (Next European Torus) contract to manufacture a shield block for the ITER primary wall module prototype. The aim was to test the influence of geometry of the tube gallery on the tolerances of tube positions after HIP of one-eighth scale (724×318×299 mm3) prototype mock-ups. It also intended to verify results obtained earlier on the influence of fabrication parameters. Five mock-ups were produced. Four HIP cycles were tested using two different designs of the cooling galleries. Non-destructive testing was used and measurements on sections of the mock-ups were performed. The location of tubes facing the first wall is most important. Investigation showed very small differences between the locations of those tubes in the five mock-ups. For these mock-ups, the tolerances between the surface and the tubes could be kept within ±2 mm.

Dynamical and Thermal Testing of a FW Panel Attachment System for the ITER Vessel

The objective of this paper is to describe and evaluate mechanical tests of the First Wall (FW) panel Attachment System (AS) for the blanket system of the fusion reactor ITER according to the 2001 design. The tests were performed in SKODA VYZKUM, Ltd. with FE simulation support from Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering. The goal of the tests was to check the stiffness, strength limit, and fatigue behavior of the bolted joint under loads simulating conditions during off-normal plasma operations in the reactor. The FW panels are attached to a bottom thick shield block by means of ten special studs located on a shaped key-way on a shield block surface. A special device for a long-lasting test of stud tensile pre-load relaxation over of 30 000 temperature cycles between 100° and 200°C was developed. Two methods were used to determine the real stud pre-load force drop during such temperature cycling. An optimum procedure for pre-loading of the AS studs has been developed. Four panel mock-ups (1080×250×50 mm) and one massive shield block having all the features of the real AS were fabricated from 316L stainless steel at VITKOVICE Research and Development, Ltd. The panel screwed to the shield with stud preload from 45 to 100 kN and was then loaded alternatively by 2500 cycles of radial moment (±24.5 kN·m), poloidal (longitudinal to panel axis) force (±108 kN), or poloidal moment (±53 kN·m) at room temperature. The stud bending stresses, stud pre-load relaxation, cyclic deformation leading to undesirable radial gap opening at the key-way and a possible plastic deformations of AS were studied. Additional FE simulations were used for better interpretation of measurements. The experimental results have shown that thermal cycling leads to a stud pre-load drop from 100 to 60 kN, whereas the dynamic cycling itself does not cause an additional loss of the pre-load. The individual loads applied do not cause a loss of radial contact in the keyway or a damage of AS even under a low stud’s pre-load of only 54 kN. A small radial gap in the key way was observed only under maximum poloidal moment with an extremely low stud pre-load of 45 kN.Copyright

EUROPEAN COMMUNITY (EC) DESIGN FOR THE ITER DRIVER BLANKET

Within the frame of the ITER conceptual design phase a Water Cooled Ceramic Blanket (WCCB) design has been developed by ENEA/ANSALDO, in close collaboration with the NET Team, for application as ITER Driver Blanket. Such a design has been selected as European design proposal for the ITER Driver Blanket and is now under comparison with the Japanese, United States and Soviet Union design proposals. An out of pile experiment is being performed to investigate the thermal-hydraulic and mechanical performance of a typical blanket module. Status of design and experimental activities is reported together with a summary of the main features and performances of the EC blanket, based on detailed analyses performed in several fields of investigation (Neutronics, Thermal-hydraulics, Thermomechanics, Tritium Dynamics, Safety, Industrial Feasibility, etc) during the past two years.

Radiolysis in EC Aqueous Lithium Salt Driver Blanket

In the NET ALSB a lithium hydroxide solution provides both breeding and cooling, resulting in a mechanically simple and robust blanket. This blanket operates below 305 C and can be switched between breeding and non breeding operation without blanket removal by adding or removing the salt. The total breeding ratio expected for this design is close to unity. Computer simulations have been performed in order to quantify the water decomposition under NET irradiation conditions. The formation of radiolytic products is given for several cases, which are relevant to NET-cooling pure water and NET-aqueous lithium salt blanket. With pure water or 2.1 M LiOH cooling, radiolytic water decomposition could be suppressed by addition of less than 10−3 M H2.