R. Tivey

Critical heat flux analysis and R&D for the design of the ITER divertor

The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW:m 2 over :10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R&D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R&D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of : 1.4 or more for all divertor components.

Assessment and selection of materials for ITER in-vessel components

Abstract During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)–IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti–6Al–4V alloy and two copper alloys, CuCrZr–IG and CuAl25–IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

Armor and heat sink materials joining technologies development for ITER plasma facing components

An extensive program on the development of the joining technologies between armor (beryllium, tungsten and carbon fibre composites) and copper alloys heat sink materials for ITER plasma facing components (PFCs) has been carried out by ITER home teams. A brief review of this R&D program is presented in this paper. The critical problems related to these joints are described. Based on the results of this program and new requirements on the reduction the manufacturing cost of ITER PFC, reference technologies for use in ITER have been selected and recommended for further development.

Erosion of Plasma-Facing Components in ITER

Abstract Erosion of ITER plasma-facing components (PFCs) is analysed with a strike-point carbon divertor target and metallic walls (Be in the main chamber, and W divertor baffles), for a ‘semi-detached’ edge plasma regime, with Type I ELMs and off-normal events (e.g. disruptions). This paper builds on earlier studies [J. Nucl. Mater. 290–293 (2001) 260], refines the assessment of erosion during full DT operation with long cumulative burn time, and discusses erosion effects expected during early phases of operation with H and DD plasmas. The present analysis should be viewed as more reliable for indicating trends, rather than providing firm quantitative predictions. Outstanding issues are identified and recommendations are made for further urgent R&D.

Design and Material Selection for ITER First Wall/Blanket, Divertor and Vacuum Vessel

Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

ITER divertor, design issues and research and development ☆

Abstract Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R&D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R&D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m−2 and tungsten armour >10 MW m−2. Analysis and experiment show that a CfC armour thickness of ∼20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within ∼6 months.

Critical plasma–wall interaction issues for plasma-facing materials and components in near-term fusion devices

Abstract The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to todays experimental facilities. These will give rise to important plasma-physics effects and plasma–material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R&D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R&D work is urgently needed.

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.