E. Martin

Development of a long reach articulated manipulator for ITER in vessel inspection under vacuum and temperature

This project takes place in the EFDA Remote Handling (RH) activities for the fusion reactor International Thermonuclear Experimental Reactor (ITER). The aim of the R&D program is to demonstrate the feasibility of in-vessel RH intervention by a long reach, limited payload manipulator which penetrates the first wall using the six IVVS penetrations. Potential activities for this device include close inspection of the plasma facing surfaces and leak detection. The work includes the design, manufacture and testing of a demonstrator articulated manipulator called the In-Vessel Penetrator (IVP). The first part of this work concerned the analysis of the requirements and resulted in the development of the conceptual design of the overall manipulator, comprising a 5 module, 11 d.o.f robot based on a parallelogram structure. A scale one mock up of a representative segment was manufactured and tested. In parallel, a feasibility study of the IVP operation was made and provided recommendations to modify the design for intervention under vacuum and temperature. Some technologies were selected and analysed to determine their suitability to the IVP application and items identified for further validation. This paper presents the whole robot concept, the results of the test campaign on the prototype demonstrator and the vacuum and temperature technologies study.

Divertor Development for ITER

The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented.

Overview of the divertor design and its integration into RTO/RC-ITER

The design of the divertor and its integration into the reduced technical objectives/reduced cost-international thermonuclear energy reactor (RTO/RC-ITER) is based on the experience gained from the 1998 design of international thermonuclear energy reactor (ITER) and on the research and development performed throughout the engineering design activities (EDA). This paper gives an overview of the layout and functional design of the RTO/RC-ITER divertor, including the integration into the machine and the remote replacement of the divertor cassettes. Design guidelines are presented which have allowed quick preparation of divertor layouts suitable for further study using the B2-EIRENE edge plasma code. As in the 1998 design, the divertor is segmented into cassettes, and the segmentation, which is three per sector, is driven by access through the divertor level ports. Maintaining this access and avoiding interference with poloidal field coils means that the divertor level ports need to be inclined (7°). This opens up the possibility of incorporating inboard and outboard baffles into the divertor cassettes. The cassettes are transported in-vessel by making use of the toroidal rails onto which the cassettes are finally clamped in position. Significant reduction of the space available between the X-point and the vacuum vessel results in re-positioning of the toroidal rails in order to retain sufficient depth for the inner and outer divertor legs. This, in turn, requires some changes to the remote handling (RH) concept. Remote handling (RH) is now based on using a cantilevered articulated gripper during the radial movement of the cassettes inside the RH ports. However, the principle to use a cassette toroidal mover (CTM) for in vessel handling is unchanged, hence maintaining the validity of previous EDA research and development. The space previously left below the cassettes for RH was also used for pumping. Elimination of this space has led to re-siting of the pumping channel between the plasma facing components (PFC) and the cassette body (P. Ladd, C. Ibbott, G. Janeschitz, E. Martin, Design of the RTO/RC-ITER primary pumping system, this conference). This gives a somewhat better conductance from the private flux region to the pumping ports than in the previous design. Diagnostic access in the divertor now also uses the cut-outs provided for pumping instead of the space below the cassettes. Developments, in particular in the area of the PFCs, aimed at reducing the cost of the divertor are reported in C. Ibbott, A. Antipenkov, S. Chiocchio, G. Federici, H. Heidl, G. Janeschitz, E. Martin, R. Tivey, Design issues and cost implications of RTO/RC-ITER divertor, this conference.

Engineering and design aspects related to the development of the ITER divertor

Abstract The adaptation of the high-recycling divertor concept for ITER is not possible because, owing to the increase in fusion power to 1.5 GW, the resulting loads on the divertor target plates are in excess of 35 MW m−2. Therefore a novel concept is proposed, based on observations of power reduction by about one order of magnitude in existing tokamaks at high-density operation. It will allow for a viable approach to the problems of energy and particle exhaust, helium pumping and impurity control at an adequate lifetime of the divertor. The concept is based on radiation and momentum exhaust along the divertor channel, and relies on the fact that a substantial fraction of the incoming power can be radiated as long as the plasma pressure is balanced not only by plasma and neutral particles in front of the target plates but also dynamically by high fluxes of recycling particles. Therefore this operating regime is called the dynamic gas target divertor. This paper describes the design and the development of the ITER divertor and shows how the physics requirements have been translated into engineering solutions, and how the additionally existing constraints (such as the space limitations, neutron effects, electromagnetic loads, compatibility with other components, easy maintainability, etc.) have been taken into account.

The ITER divertor

The ITER divertor design is based on the high density exhaust concept of reducing peak heat fluxes falling on the divertor plates. The power is transferred from the edge plasma to the walls of the divertor chamber by atomic and molecular processes such as radiation, charge exchange, recombination, and gas conduction before it can reach tile divertor plates. Although the feasibility of this concept for ITER has not yet been fully established, experiments in present day tokamaks show that such a configuration can be sustained and preliminary results indicate that the present concept is promising, even for the very demanding power levels of ITER. Simulation codes are not yet complete and more progress in modelling high density, high power experiments is still needed. Model development and validation continues to be one of the main efforts in the development of the ITER divertor concept, in particular on modelling the implication of transients on the divertor. This paper summarises the ITER divertor concept, discusses the state of supporting experiments and models and focuses on design requirements approaches.

Design of the RTO/RC ITER primary pumping system

The primary pumping system is needed not only to exhaust helium ash resulting from the DT reaction but also excess fuelling gas injected during the fusion burn, which can extend for 100s to 1000s of seconds, and to perform a variety of other functions. The prevailing environmental conditions, principally nuclear radiation, tritium exposure, magnetic fields, and the need for containment, have a significant impact on the design and selection of equipment. This paper presents the design of the Reduced Technical Objectives/Reduced Cost (RTO/RC) ITER primary pumping system with particular emphasis on the nuclear aspects of the design. Component selection and equipment layout issues to meet established requirements for the system are reviewed together with the R&D that is being undertaken to support the design. In addition, serviceability and maintainability issues related to this system are also discussed.

Remote Maintenance of In-Vessel Components for ITER

ITER in-vessel components must be remotely maintained due to neutron activation. Components that require maintenance include the blanket shield modules, divertor cassettes and ancillary systems mounted in the vacuum vessel (VV) ports. Maintenance is predominantly accomplished by component removal and transfer to the hot cell facility for repair or waste processing. Component transfer between the VV and the hot cell is performed in sealed casks that dock to the VV ports. An overview of the in-vessel remote maintenance requirements, techniques and equipment is presented.

Overview of the engineering design of ITER divertor

Abstract The ITER divertor cassettes, which support plasma-facing components (PFCs), provide the flexibility to adapt to changing design requirements imposed by the evolving understanding of divertor physics. This paper gives an overview of how the divertor has been adjusted to meet the most recent changes in the ITER design also describing some cost saving design simplifications. Furthermore, the paper deals with an overview of the engineering R&D, shared amongst the Home Teams, that is underway in support of PFC development. In particular the paper summarises the ‘hot’ liner R&D. This R&D has produced interesting laboratory results and their relevance in mitigating carbon–tritium deposition in ITER is discussed.

Performance and remote maintenance of attachment schemes for Plasma Facing Components

The divertor design for the ITER-FEAT fusion reactor is based on cassettes which comprise a reusable body and three sacrificial Plasma Facing Components (PFCs) expected to be replaced in a hot-cell a number of times during machine lifetime. Central to this maintenance approach are the PFC-to-cassette attachments which must be readily assembled/disassembled by remote handling methods and withstand severe mechanical and thermal loading conditions during machine operation. This paper describes the facilities, equipment and methods used to carry out extensive testing of two attachment schemes, shear keys and multi-links, in order to assess their in-service performance and suitability to remote maintenance operations.

Design issues and cost implications of RTO/RC-ITER divertor

Abstract This paper reports on the conceptual divertor design developed for the reduced technical objectives/reduced cost-international thermonuclear experimental reactor (RTO/RC-ITER). The cost drivers are discussed and a number of cost-reducing measures identified. Scaled costs, based on industrial estimates of the 1998 ITER design (Technical Basis for the ITER Final Design Report, Cost Review and Safety Analysis (FDR). ITER EDA Documentation Series No. 16. IAEA, Vienna, 1998), give for the RTO/RC-ITER ≈60% of the FDR costs. Plasma facing components (PFCs) account for 75% of the total divertor costs. Hence, PFC design simplifications are outlined in the paper showing the possibility of achieving a cost reduction of 50%. The design proposals, outlined in the paper, focus on minimising the number of sub-components and simplifying the manufacturing cycle. These changes contribute to improved reliability based on a more robust coolant design layout. The reduced space allocated to the divertor (G. Janeschitz, A. Antipenkov, V. Barabash, S. Chiocchio, G. Federici, C. Ibbott, E. Martin, R. Tivey, Overview of the Divertor Design and its Integration into RTO/RC-ITER, this conference) requires changes to the design that minimise the cassette body thickness, relocate the cassette attachments and revise the remote handling philosophy. Results of supporting electro-magnetic, neutron shielding, thermo-hydraulic and pumping conductance analyses are reported, qualifying the cassette design. A reduction in the coolant inlet temperature to 100–120°C is discussed in terms of thermal-hydraulic performance and fatigue life of the heat sink. Finally, an RD and (2) to demonstrate the reliability of the chosen technologies.