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: frederic.escourbiac@iter.org

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. The analysis is underway to verify the design.

4. Heat loads specifications

The heat load specifications are the most important driver for the plasma-facing surface design. Specifications exist for the previous divertor, as part of the general ITER Heat and Nuclear Load Specification [7], but new, more detailed analysis has been performed in support of the full-W divertor design. Surface heat loads for steady-state, slow transient regimes and transient events due to downward VDEs and major disruptions have been estimated using the most recent simulations from the DINA and SOLPS codes [2]. An important additional detail has been the inclusion of a more rigorous analysis of the likely number and type of tokamak discharges which a full-W divertor installed from Day 1 and being replaced at the end of the first DT campaign would experience. This also provides input to the technology qualification programme, allowing the required heat flux testing cycle numbers to be more quantitatively defined [3].

Surface heat fluxes at steady-state and slow transient regimes

Assuming that a full-W divertor installed from the beginning of operations should survive at least until the end of the first full DT campaign, according to the current ITER Physics Research Plan [8], it will be exposed to two helium/hydrogen campaigns, a single deuterium campaign and one full DT campaign, at the end of which the first mission goal of long pulse operation at high fusion gain (Q = 10) would be achieved. This amounts to some ~25000 plasma pulses. Analysis concludes that the divertor installed throughout this period should expect to experience around 1500 cycles in which the peak divertor heat flux would transiently exceed 10 MWm -2, 3700 cycles to a peak stationary heat flux of 10 MWm -2 and approximately 300 cycles in which a peak value of 20 MWm -2 could be achieved on a slow transient timescale (namely over period of several seconds, typically at most 10 s) [2].

Surface heat loads during downward VDE and major disruptions

With regard to heat loads, component lifetime is determined both by thermal cycling and the very short duration, high energy density transients resulting, for example, from VDEs and major disruptions. In the case of W, melting driven by transients is recognized as one of the most outstanding risks associated with its use in the ITER divertor. Such melting, in addition to the erosion caused by material loss, can lead to surface topology modifications, which may be aggravated by subsequent plasma operation (steady state or transient). Dust created by melt transients may also settle on plasma wetted surfaces, posing a threat to plasma operation given the very low levels of W impurity which can be tolerated in the core of a burning plasma. A particular concern for the W divertor is the high power downward VDE, which is expected to impact and quench on the outer baffle region [2]. When this occurs, the plasma will be in limiter configuration so that magnetic field lines no longer attack with a common direction (as in the case of a divertor plasma). The consequence is that both toroidal extremities of the baffle must be shaped to avoid thermal plasma deposition on leading edges. The more detailed heat load specification includes parallel power flux or energy densities and power widths for all disruptive transient cases, including a breakdown of approximate event numbers for the different operational phases of ITER up to the end of the first DT campaign [2]. It is therefore possible to derive the expected 3D thermal load mapping 5 ITR/P5-08 on PFU envelopes (an example is shown in Fig.1) with 3D field line tracing calculations, using magnetic equilibrium reconstructions provided by the DINA code.

2 Einc (J/m ) 2 front surface: Einc max ~ 5 MJ/m

2 chamfer: Einc max ~ 28.7 MJ/m

leading edges: shadowed

FIG.1: 3D incident energy density distribution on the outer divertor cassette surface during an 2 unmitigated downward VDE expected in the He-H II phase: E // = 130 MJ/m and λe = 0.03m at the outer mid-plane. The model includes a variant of the baffle shaping being proposed to mitigate the effects of this transient impact (Section 5)

At the moment of the VDE thermal quench, DINA simulations find that the inner vertical target and divertor dome areas extremely rarely impacted [2]. They do not therefore require any particular shaping modification compared with the existing CFC variant.

5. Shaping

Tilting

In the current design, the PFCs are tilted when assembled onto the CB (see FIG. 2), ensuring that any misalignment during the assembly of divertor cassette onto the toroidal divertor rails in the VV does not produce leading edges which would be directly intersected by particles following magnetic field lines incident at glancing angles in the high heat flux (HHF) areas during steady-state operation. Typically a tilting angle of 0.4º creates a step of 3-4 mm at the level of the straight part of the OVT and IVT (the primary strike point regions), which ensures full shadowing of the PFC leading edge taking account the assumed tolerances on cassette assembly (+/- 2mm) and on the surface of the PFU itself (profile tolerance of 0.5 mm is required to avoid leading edges melting during slow transient regimes). The high (HFS) and low field sides (LFS) of the dome umbrella are toroidally tilted in opposite directions (around a centred vertical axis – see FIG. 2) to account for the possibility of loss of control situations in which the outer strike point falls on the LFS or the inner strike impacts the HFS of the umbrella (thus on unprotected leading edges - assuming fixed magnetic field helicity corresponding to standard ITER operation with negative toroidal field and plasma current directions). The reflector plates are also tilted (to protect against loss of 6 ITR/P5-08 control situations in which the strike points fall onto the plates, where leading edges will again be exposed), with the axis aligned along the plasma-facing surfaces (FIG. 2).

FIG. 2: Tilting axis of target (IVT as example with particle incidence indicated for steady-state operation) and Dome

Roof-shaping

As discussed in Section 4, limiter-like impact on the OVT baffle during downward VDEs would place very severe transient heat loads on exposed edges during high performance phases. These loads can be somewhat mitigated by adopting the same “roof-shaping” principle (see FIG. 3) used in the ITER First Wall panel design [9]. Toroidal “set-backs” (typically 10-20 mm, currently under optimization) are proposed on each side of the baffle to completely shadow any direct leading edge to heat fluxes arriving on either side during limiter impact, whilst optimizing the surface heat load spreading over the remaining wetted surface. However, designing the baffle in this way introduces the need for a transition from global roof-shaping to local fish-scale shaping (see paragraph below) in the HHF regions and is one of the main difficulties of the current full-W design. Additionally, this transition region must accommodate a monoblock armour thickness variation, from 8 mm on the straight part of the target (design value to ensure high thermal performance, to be confirmed by results of qualification programme) to 13 mm at the thickest part of the chamfered monoblock in the baffle region (see FIG. 3).

Fish-scaling

The third shaping issue introduced by the use of W concerns the requirement to systematically hide monoblock edges on the HHF regions of the IVT and OVT. To counteract a potential misalignment due to assembly tolerance of neighbouring PFUs (required at maximum 0.3 mm), local monoblock plasma-facing surface shaping is mandatory (see FIG. 4). A simple “fish-scaling” solution, in which each monoblock is chamfered only in the toroidal direction (typically with chamfer height 0.5 mm) is the concept currently retained, offering the best 7 ITR/P5-08 compromise between manufacturing complication (only one additional cut per block) and wetted area reduction of ~ 25%. The reality is that an optimum must be sought between perfect shadowing protecting all possible misalignments, the resulting reduction of the wetted area (and thus increased surface heat loads) and the increase of manufacturing costs due to complex shapes. This optimization will be finalized during the final design phase study, using the 3D thermal load mapping taking into account the full magnetic equilibrium geometry.

FIG. 3: Set-back at OVT Baffle part

FIG. 4: Fish-scaling at vertical target (OVT as example) . 6. Summary

Design of a full-W divertor for ITER is in progress following a proposal in 2011 by the ITER organization to modify the divertor replacement strategy and consider only a single, full-W component from the beginning of plasma operations. Neutronic analysis has been performed on a pre-detailed design and the results are now being used for structural analysis, aiming at verifying that the design is acceptable with regards to the ITER Structural Design Criteria. A new, more detailed divertor surface heat load specification based on physics analysis has been 8 ITR/P5-08 produced and is the principal input for design of the plasma-facing components. A key feature of the new design, specific to the use of tungsten, is the requirement to eliminate all possible leading edges which may arise due to relative misalignment of components, either cassette to cassette, or between individual heat flux elements on any given target surface. The solutions proposed are a compromise between optimization of wetted area and minimization of manufacturing costs (with respect to the baseline CFC/W design) engendered by the increased complexity. Three levels of shaping are currently thought to be necessary: component tilting, the addition of global “roof-top” like shaping and the introduction of local fish shaping on individual heat flux elements in high heat flux areas. Optimization is in progress and the final design phase is planned to be completed by the end of 2013.

7. References

[1] T. Hirai et al., "Design and Integration of ITER Divertor Components”, Advances in Science and Technology, Vol. 73 (2010) 1-10. [2] R. A. Pitts et al., “A full tungsten divertor for ITER: physics issues and design status”, J. Nucl. Mater., in press. [3] T. Hirai, et al. “ITER Tungsten Divertor Design Development and Qualification Program”, SOFT2012 [4] R. Villari et al., “Neutronic Analysis of the ITER full-tungsten Divertor”, SOFT 2012 [5] G. Sannazzaro et al., “Structural load specification for ITER tokamak components”, Proc. of SOFE 2009, 1-5 June [6].G. Sannazzaro et al., “Development of Design Criteria for ITER In-Vessel Components”, SOFT 2012 [7] R. Pitts et al., Journal of Nuclear Materials 415 (2011) S957–S964

[8] D. J. Campbell et al., “Challenges in Burning Plasma Physics: the ITER Research Plan” 24 th IAEA Fusion Energy Conference 2012 in San Diego [9] P. Stangeby et al., “Analytic expressions for shaping analysis of the ITER outer wall”, Nucl. Fusion 51 (2011) 103015

The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.