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Effort on Design of a Full Tungsten Divertor for ITER

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Effort on Design of a Full Tungsten Divertor for ITER 1 ITR/P5-08 Effort on Design of a Full Tungsten Divertor for ITER F. Escourbiac 1, T. Hirai 1, S. Carpentier-Chouchana 1, A. Fedosov 1, L. Ferrand 1, S. Gicquel 1, T. Jokinen 1, V. Komarov 1, A. Kukushkin 1, M. Merola 1, R. Mitteau 1, R. A. Pitts 1, P.C. Stangeby 2, M. Sugihara 1 1 ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance, France, 2University of Toronto Institute for Aerospace Studies Toronto, M3H 5T6, Canada e-mail address of main author: [email protected] Abstract The design status of an all-tungsten divertor for ITER, currently under development at the ITER Organization, is reported in this paper. Neutronic analysis has been performed on a pre-detailed design and its output is used for running structural analysis, aiming at verifying that the design is acceptable according to the ITER Standard and Design Criteria. A new, more detailed divertor heat load specification has been developed as the main input for the design of the Plasma Facing Units. A key feature with respect to the current baseline (CFC/tungsten) design is the implementation of shaping to avoid as much as possible the appearance of any leading edges which can severely compromise the performance and lifetime of the divertor when using metallic plasma-facing components. The final design solutions when incorporating such shaping is a trade-off between reduction of wetted area (and hence power handling) and an increase of manufacturing costs due to increased complexity. In accordance with deadlines fixed by the ITER Council, the final design phase is planned to be completed by the end of 2013. 1. Introduction In the reference ITER divertor strategy, carbon fibre composite is planned for plasma-facing material in the strike point (high heat flux) regions throughout the non-active (hydrogen- helium) campaigns. This divertor is to be replaced by a full tungsten (W) variant before nuclear operations begin [1]. The issue of divertor material choice is one topic of controversy within the magnetic fusion community, notably as a consequence of the relative lack of experience of tokamak operation on tungsten in comparison with carbon together with materials and technology issues. Technical aspects aside, there is a strong argument in favor of tungsten: substantial cost reductions could be achieved if a single divertor would be installed from the start of ITER operations and survive well into the nuclear phase. Since the current ITER strategy is in any case one in which W is used as sole divertor material for nuclear phase operations (principally in order to avoid the high levels of tritium accumulation expected to occur if carbon is present with DT fuel), starting with W has considerable merit if it can be demonstrated that the increased risks to plasma operations of using W are acceptable and if the required technology developments can be demonstrated before procurement begins. This cost reduction strategy was proposed in mid-2011 by the ITER Organization (IO) and, following a recommendation from the ITER Council, is currently under study with the objective that a decision on the material choice could be taken near the end of 2013. This paper reports on the status of the design development at the IO approximately one year after the beginning of the design effort. In parallel with the design activities, physics R&D [2], material R&D and a technology qualification programme [3] are being undertaken within laboratories of the ITER Members and with the concerned Domestic Agencies (DAs) to provide as much as possible of the necessary information required to make an informed decision on material choice. 2 ITR/P5-08 2. Main design features The ITER Divertor consists of 54 divertor cassettes assembly (see FIG.1) positioned in the bottom part of the Vacuum Vessel (VV) via a nose and knuckle in contact with inner and outer toroidal rails. In order to minimize impact on Cassette Body (CB) design, diagnostic interfaces and running Procurement Arrangements (PA), the main features of the full-W divertor remain largely unchanged in comparison with the existing CFC/W reference design [1]: the three main plasma-facing components (PFC), namely the Inner Vertical Target (IVT), the Dome and the Outer Vertical Target (OVT) are mounted onto the CB. Water cooling is distributed to the PFCs via the CB which also provides neutron shielding for the VV and acts as the main support for the PFC. FIG.1: Divertor cassette assembly (full-W design) The IVT and OVT are comprised of actively cooled poloidal plasma facing units (PFU), themselves constituted of monoblock chains which act to shield the CB from the plasma heat and particle fluxes. The monoblock concept, which has proven its robustness in the past with regard to heat and neutronic loads, is retained with both unchanged tolerances at plasma-facing surfaces and attachment concept. For the dome design, changes with respect to the baseline design are limited to modification of the tilting angle incorporated to ensure shadowing of leading edges from cassette to cassette. The principal modifications to the design are being made in order to mitigate as much as possible the effects of tungsten melting under transient heat pulses (notably due to the disruptions and downward vertical displacement events (VDE)) – see Section 5. This is considered the greatest risk associated with the use of W in the divertor, since it is known that a surface deformed by melting does not repair under plasma exposure and can compromise operation of the entire device [2]. Tungsten is particularly difficult in the sense that only very small concentrations can be tolerated in the core of a burning plasma. The design activity is being pursued in 3 phases: - The pre-detailed design phase : to provide the design models for cost estimation by the DAs and for neutronic and structural analysis. The pre-detailed design includes the main features of the final design, in particular the presence of PFU shaping is implemented to assess its effects on neutronic shielding and electromagnetic loads. The shaping will be refined during the final design phase; 3 ITR/P5-08 - The preliminary design phase: to finalize the structural design such that it is consistent with the results of the neutronic and structural analysis; - The final design phase : to deliver the design and 2D general assembly drawings of the structural parts and PFUs, including the finalized shaping, all in accordance with the results from R&D activities of the qualification programme [3]. 3. Supporting analysis Neutronic analysis The pre-detailed design model has been integrated into the latest ITER MCNP-5 Monte Carlo Code in 3D geometry to perform the neutronic analysis. Nuclear heating, damage and helium production have been estimated [4]. Regarding nuclear heating, the conservative results obtained for a fusion power of 500 MW are comparable with previous analysis for the CFC/W baseline: maximum values of 10- 11 MW/m 3 in the W part of the PFU and 3-4 MW/m 3 in the copper alloy and steel supporting structures are expected. The shielding capabilities of the updated dome design are reduced in comparison with the previous design analysed in 2007, leading to an additional nuclear heating generation in the CB plates located below the dome of 1 MW/m 3 (to be compared to 0.5 MW/m 3 in the past). This should have a negligible impact on the CB design, but must be confirmed by the structural analysis. Similarly, it is not expected that the full-W divertor design will lead to an appreciable increase of the nuclear loads on vacuum vessel and toroidal field coils, since the relative contribution of the divertor region remains low. Assuming that the first divertor set will be replaced after the first nuclear campaigns (DD phase plus first major DT campaign), it would receive 18% of the cumulated 3x10 27 neutrons expected during the whole life of the ITER machine. Under this assumption, the maximum level of nuclear damage is estimated at ~0.5 dpa inside the copper alloy at OVT baffle region. The re-weldability of the CB radial pipes is not a concern even assuming irradiation during the whole ITER lifetime: the He-production (0.14-0.16 appm) is well below both the design target (<1 appm) and the criterion for re-weldability (<3 appm). Following this criterion, the re-weldability of pipes for the refurbishment of the CB with new PFCs is still possible after the first nuclear campaigns. Structural analysis Since ITER specific conditions (high energy flux neutron radiation, fast and slow variation of electromagnetic fields, high heat fluxes, vacuum, etc.) are not all covered by a single existing industrial code (e.g. ASME, Harmonised European Standards, RCC-MR), Structural Design Criteria for Internal Components (SDC-IC) was selected as the design code for the divertor [5]. During the pre-detailed and preliminary design phase, the design development of the structural part is supported by structural analysis aiming at estimating its static and cyclic strength following the SDC-IC code. The structural loads used as input for this exercise are derived from the ITER Load Specification [6], which includes inertial loads (associated with gravity and seismic events); hydraulic pressure loads, electromagnetic loads, thermal loads (due to nuclear heating, water temperature in different regimes and plasma radiation but not from conductive/convective plasma heat loads, the structure being protected by the PFUs) and assembly loads, typically due to preloads imposed on the CB during assembly. From this specification and past results 4 ITR/P5-08 obtained on the CFC/W divertor, an envelope for load combinations essential for divertor stress analysis has been defined.
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