Recent analysis of key plasma wall interactions issues for ITER

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Abstract

Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q = 10 ITER discharge [G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4 ± 3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.

Introduction

Since the last PSI conference in 2006, the ITER Joint Implementing Agreement has been signed by the seven partners of the project, allowing to launch the construction of the machine [2], [3]. By end 2006, a design review process has been started, including discussion of urgent plasma wall interactions (PWI) issues, in particular those needing evaluation for the licensing authorities. The most critical PWI issues have been identified as:

  • lifetime of plasma-facing components (PFCs);

  • dust production from eroded PFCs;

  • tritium (T) inventory in the vacuum vessel.

This paper presents an assessment of these issues performed during the design review process through the European Plasma Wall Interaction Task Force (EU PWI TF) and, for the case of tritium retention in W, the US Burning Plasma Office (BPO).

In the evaluation of the above issues, which determine the choice of PFC materials (such as carbon fibre composite (CFC), tungsten (W), or beryllium (Be)) for reliable and safe operation of ITER, less emphasis was set on the detailed understanding of individual physical processes – previous reviews will be referenced throughout the paper – than on the consolidation of these individual processes in establishing robust predictions and associated uncertainty margins.

In Section 2 of this paper, ITER safety limits for PWI issues, such as T and dust inventories, are reviewed. In Section 3, input parameters used for the assessment, as well as material options considered, are described. Section 4 presents the assessment of erosion of PFCs, both from steady state and transient loads. Erosion rates derived in Section 4 are then used to evaluate dust generation in Section 5, and T inventory in Section 6. Different material options are addressed for comparison (CFC divertor/W baffles + dome/Be first wall, W divertor/Be first wall, full-W, full-C). Finally, consequences for the plasma scenarios and the PFC material choice are summarised in Section 7.

Section snippets

PWI related safety issues for ITER

Although not a concern in present day tokamaks, in vessel dust and tritium inventories have been recognised as a safety and operational issue for next step devices such as ITER [4], [5]. Safety related issues concerning mobilisable in vessel dust (size between 100 nm and 100 μm) inventory include:

  • contribution to the in vessel T inventory;

  • potential radioactive (mainly W) and toxic (Be) source term in case of accidental release in the environment;

  • potential hydrogen production from the reaction with

Input plasma parameters

Input plasma particle and energy fluxes, as well as surface temperatures, are taken for a reference 400 s Q = 10 ITER discharge as evaluated in [8] and used in [9]. The resultant fluxes are illustrated in Fig. 5 of Ref. [9].

In the divertor near the plasma strike point the typical ion and neutral fluxes reach values larger than 1024 m−2 s−1 (leading to a total fluence >1026 m−2 for each ITER pulse) with divertor plasma densities ∼1021 m−3 and plasma temperatures of ∼3 eV. This corresponds to a D+ impact

Lifetime of plasma-facing components

The first step in the chain of processes determining the PFCs lifetime, leading to dust generation and tritium retention by co-deposition, is the erosion of the wall material.

Dust generation

In tokamaks, dust can be produced during various operation phases:

  • Layer deposition and disintegration in steady state.

  • Disruptions.

  • Arcing [47], [48].

  • Operations during maintenance phases.

In this study, we will only consider the first two points. Dust is formed either directly by erosion processes leading to ejection of particulates or droplets, or by delamination of re-deposited layers. In both cases the formation rate is primarily determined by the respective erosion rate, which also represents

Tritium inventory

Tritium inventory accumulation in ITER has been the topic of a review published recently [9] using the same evaluation method as in the present paper. It will, therefore, be summarised here only shortly.

Consequences for plasma scenarios and material choice

From the estimates given above, the performance of different wall materials as well as limits on plasma scenarios can be discussed:

  • Transient wall loading by ELMs and disruptions, which are usual in present fusion devices, must be strongly limited in ITER. Experimental studies of ELM-like power loads in linear plasma devices [30] have shown that both potential divertor materials, CFC and W, will erode strongly when the ELM energy density exceeds 0.5 MJ/m2. Plasma scenarios with pellet pacemaking

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