K. Ioki

FW/Blanket and vacuum vessel for RTO/RC ITER

Abstract The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

Abstract Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal–hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R&D and a brief description of future R&D effort to address remaining issues are also included.

Beryllium Application in ITER Plasma Facing Components

Abstract Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R&D program are highlighted.

Improved modules for the blanket of RTO/RC ITER

This paper describes innovative design aspects that are considered to optimise the blanket modules for the reduced technical objective/reduced cost international thermonuclear experimental reactor. The blanket modules have a vertical straight profile facing the plasma, and the first wall is built in small and flat panels. Copper may be applied only in front of the first row of cooling passages. The radial cooling of the shield block avoids a complex by-pass at the back and opens up the possibility to use cast instead of forged steel. Slits in the shield block and in the first wall reduce the electromagnetic forces enough to allow the support of the modules on the vessel and the mechanical attachment of the first wall panels.

Improvements in the design and manufacture of the ITER FEAT first wall towards cost minimization

The ITER FEAT first wall (FW) has been designed with modular flat panels that are separable from the massive blanket shield, in order to reduce costs, to allow repair or substitution, and to reduce nuclear waste. Thermal stresses and induced electromagnetic loads in the structure of the FW are reduced well below the allowable values by introducing poloidal slots in its heat sink. The paper describes the reasons behind recent design modifications of the ITER FW and reports the main results of thermo-mechanical and structural analyses performed to verify the improved design.

Design and analysis of the vacuum vessel for RTO/RC-ITER

Abstract Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a ‘tight fitting’ configuration is required to correctly position the modules’ first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible.