S Chiocchio

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.

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.

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.

Design, Materials and R+D Issues of Innovative Thermal Contact Joints for High Heat Flux Applications

Abstract Plasma facing components in fusion machines are designed with a layer of sacrificial armour material facing the plasma and a high-conductivity material in contact with the coolant. One of the most critical issues associated with making the proposed design concept work, from a power handling point of view, is achieving the necessary contact conductance between the armour and the heat sink. This paper presents a novel idea for the interface joint between the sacrificial armour and the actively cooled permanent heat sink. It consists of a thermal bond layer of a binary or more complex alloy, treated in the semi-solid region in such a way as to lead to a fine dispersion of a globular solid phase into a liquid matrix (rheocast process). The alloy in this “mushy state” exhibits a time-dependent, shear rate-dependent viscosity, which is maintained reversibly when the material is solidified and heated again in the semi-solid state. The function of the thermal bond layer is to facilitate heat transfer between the replaceable armour and the permanent heat sink without building up excessive thermal stresses, as in conventional brazed joints, and allow an easy replacement whenever needed without disturbing the coolant system. No contact pressure is required in this case to provide the desired heat transfer conductance, and the reversible thixotropic properties of the rheocast material should guarantee the stability of the layer in the semi-solid conditions. Key design, material and testing issues are identified and discussed in this paper with emphasis on specific needs for future research and development work. Examples of suitable material options which are being considered are reported together with some initial heat transfer analysis results.

Design of the ITER EDA plasma facing components

Abstract The design of the plasma facing components (PFC) in ITER has evolved with the detailed design of the reactor. The structures exposed to the plasma have different requirements according to their functions. The primary wall, surrounding most of the plasma along the last closed magnetic surface, is exposed to a moderate heat flux (0.5 MW m−2) but has to withstand the highest neutron load. The baffle wall is exposed to a peak heat flux of 3 MW m−2 and to severe erosion from neutral particles due to their high neutrals pressure in the divertor. The limiter is subjected to the same loads as the primary wall during plasma burn conditions and a higher peak heat flux (depending on its location) during the start-up and shut down phases when the plasma is leaning on its surface. The divertor vertical targets intercept the open magnetic flux surfaces near the separatrix and have to withstand the highest heat flux and erosion in their lower part. The divertor dome is located directly below the null point and works in conditions similar to the baffle. The divertor wings receive similar thermal loads as the dome but can be subjected to high heat shocks and electromagnetic forces during plasma disruption. The paper describes the solutions adopted for the PFC and the results of analyses performed to validate the design. The description is focused on the part of the PFC which is exposed to the plasma.

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.

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.

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.