H.D. Pacher

Analysis of performance of the optimized divertor in ITER

The paper describes the results of a physics analysis of a modified divertor cassette for ITER. The issues addressed are the impact on the operational window, the effect of gas leaks through the broader gaps between the divertor cassettes and radiation power loading of different components of the cassettes. The analysis shows that the new design ensuring more flexibility for ITER operation remains acceptable within the framework of the usual trade-off between the target power loading and helium removal efficiency. The radiation load on the side walls of the cassette structures in the inter-cassette gaps is identified as a design constraint not previously considered.

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 issues on ITER and extrapolation to reactors

The current status of divertor modelling for ITER is presented, and major physics and technology constraints on the divertor operation are discussed in the paper. Extensive exploration of the operational window of the ITER divertor has lead to the emergence of simple scalings of the divertor plasma parameters with input power and plasma density, which are used here to make an educated guess on the divertor performance in a commercial reactor. The impact of fast transient events (ELMs), causing significant variation of the power loading, on the divertor operation and design in ITER is discussed and their implications for a reactor are shown. The issue of tritium co-deposition via hydrocarbons inside the vacuum vessel and in the pumping ducts is considered for ITER and projected to a reactor.

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.

ITER operation window determined from mutually consistent core–SOL–divertor simulations: definition and application

An operating window for ITER is defined based on mutually consistent core–SOL–divertor modelling, in which the core turbulent transport is based on the Weiland formulation as incorporated into the multi-mode model. The window consists of five limits, one of which is the edge-based density limit based on divertor detachment. The predicted operating space is ample for ITER to fulfil its mission, reaching a maximum Q ≈ 60 at Palpha ≈ 150 MW (65% of the edge-based density limit, 1.1 times the Greenwald limit), and a maximum Palpha of 220 MW at Q ≈ 15 (90% of the edge-based density limit, 1.45 times the Greenwald limit). This operating window takes into account physics constraints and the technical constraints imposed by the divertor system, i.e. peak power load and attainable pumping speed, but does not include further constraints arising from other technological aspects of the ITER design, such as first wall cooling or shielding, which may further limit operation at high fusion power. The operating window is still compatible with the ITER mission if the magnetic field were reduced by 5% or if the underlying core transport were GLF-like rather than Weiland-like. A moderate reduction in helium exhaust or in pumping speed could be accommodated. Other changes in the operating window resulting from different technical or physical hypotheses are also evaluated.

Modelling of DEMO core plasma consistent with SOL/divertor simulations for long-pulse scenarios with impurity seeding

The integrated core-pedestal-SOL model is applied to the simulation of a typical DEMO operation. Impurity seeding is used to reduce the power load on the divertor to acceptable levels. The influence on long-pulse operation of impurity seeding with various impurities is investigated. DEMO operation at acceptable peak power loads and long-pulse lengths is demonstrated.

Numerical estimates of the ITER first wall erosion due to fast neutral particles

Estimates of the ITER first wall steady-state sputtering due to fast atoms are made based on the B2-EIRENE modelling of the scrape-off layer (SOL) plasma. Models with fully diffusive and convective cross-field transport in the SOL are investigated. It is found that strong radial convection in the far-SOL has only a weak effect on the plasma profiles there. Effective yields of the physical sputtering of Be, C, Mo and W as well as their corresponding erosion rates at the first wall are calculated. The results are compared with available published data.

Scaling of ITER divertor parameters – interpolation from 2D modelling and extrapolation

Detailed modelling studies of the divertor plasma for ITER have been carried out. Using these results, scaling relationships are developed linking SOL power, density, throughput, pumping speed, peak divertor power load, and helium density for ITER conditions in order to systematise the results and to extrapolate them beyond the range presently covered by the simulations. The key parameter for the scalings is the neutral pressure in the divertor. Both peak power load and helium density vary as the square of the power at constant pressure. The inclusion of helium elastic collisions reduces the helium density and leads to a steeper reduction with increasing pressure. Variants of the input conditions, i.e. different geometry, no helium elastic collisions, carbon walls, are also discussed, the consistency of the edge modelling with conditions required in the core is treated, and extrapolation to higher power operation is carried out.

Impact of potential narrow SOL heat flux on H-mode access in ITER

The paper presents results of a first analysis of the divertor performance during the L–H transition in ITER. The integrated model consists of the SOLPS4.3 code suite for the SOL and divertor, and the ASTRA code for the core and pedestal regions. The results of SOLPS4.3 are parametrized and used as the boundary conditions for ASTRA, ensuring a consistent description of the plasma core and the edge. Boundary conditions switch from those for wide (L-mode) to narrow (H-mode) SOL once the transition criterion is met. The results show that, for conditions for which a full-power operational space with acceptable power loading of the targets exists, a transition from the initial L-mode operation to H-mode can be found for the same assumptions, i.e. the full-power H-mode regime is accessible.

ITER divertor performance in the low-activation phase

The paper presents results of SOLPS modelling of the edge plasma performance during the low-activation phase of ITER operation. The calculations show that the peak power loading of the divertor targets can reach the reactor-relevant level of 3 to 5?MW?m?2, even without the fusion reactions, rendering commissioning of the high heat flux components possible in this phase. Parametrization of the output of the SOLPS runs for the predominantly helium plasma concerned by the studies reported here is performed, thus providing the boundary conditions for modelling of the core and allowing efficient integration of the core and edge models. This approach, using the ASTRA code for core simulations, is applied to the analysis of hydrogen accumulation in helium plasmas due to H pellet injection. The latter is the only available option for early testing of ELM pace-making as an ELM control tool assuming H-mode in hydrogen will not be possible. Critical dilution with H down to 70% He in the core plasma can be reached in only 0.5 to 1?s or even shorter, depending on the assumptions made.