Improved ERO modelling of beryllium erosion at ITER upper first wall panel using JET-ILW and PISCES-B experience

ERO is a 3D Monte-Carlo impurity transport and plasma-surface interaction code. In 2011 it was applied for the ITER first wall (FW) life time predictions [1] (critical blanket module BM11). After that the same code was significantly improved during its application to existing fusion-relevant plasma devices: the tokamak JET equipped with an ITER-like wall and linear plasma device PISCES-B. This has allowed testing the sputtering data for beryllium (Be) and showing that the "ERO-min" fit based on the large (50%) deuterium (D) surface content is well suitable for plasma-wetted areas (D plasma). The improved procedure for calculating of the effective sputtering yields for each location along the plasma-facing surface using the recently developed semi-analytical sheath approach was validated. The re-evaluation of the effective yields for BM11 following the similar revisit of the JET data has indicated significant increase of erosion and motivated the current re-visit of ERO simulations.

The software and hardware architecture of the real-time protection of in-vessel components in JET-ILW

Otto Asunta, Patrick Blanchard, Yann Camenen, Stefano Coda, Basil Duval, Ambrogio Fasoli, Jonathan Marc Philippe Faustin, Davide Galassi, Javier García Hernández, Jonathan Graves, Jan Horacek, Samuel Lanthaler, Yves Martin, Mikhail Maslov, Federico Nespoli, Hamish William Patten, Alessandro Pau, Antonio José Pereira de Figueiredo, David Pfefferlé, Richard Pitts, Olivier Sauter, Cristian Sommariva, Duccio Testa, Nicola Vianello, Henri Weisen, Dalziel Joseph Wilson, Marco Wischmeier, Liang Yao, Yao Zhou

For the first time, the JET operation in deuterium-tritium (D-T) plasma, which is scheduled to take place on JET in 2020, will be performed in the ITER mix of plasma-facing component materials. In view of the preparation of the DT campaign (DTE2), several aspects of the plasma operation require significant improvements, such as a real-time protection of the first wall. The risk of damaging the metallic PFCs caused by beryllium melting or cracking of tungsten owing to thermal fatigue required a new reliable D-T compatible active protection system. Therefore, the future development of the JET real time first wall protection is focused on the D-T campaign and the ITER relevant conditions which may cause failure of camera electronics within the Torus hall. In addition to the technological aspect, the intensive preparation of the diverse software tools and real time algorithms for hot spot detection as well as alarm handling strategy required for the wall protection is in progress. This contribution describes the improved design, implementation, and operation of the near infrared (NIR) imaging diagnostic system of the JET-ILW plasma experiment and its integration into the existing JET protection architecture. To provide the reliable wall protection during the DTE2, two more sensitive logarithmic NIR camera systems equipped with new optical relays to take images and cameras outside of the biological shield have been installed on JET-ILW and calibrated with an in-vessel calibration light source (ICLS). Additionally, post-pulse data visualization and advanced analysis of all types of imaging data is provided by the new software framework JUVIL (JET users video imaging library). The formation of hot spots is recognized as a significant threat due to rapid surface temperature rise. Because it could trigger the protection system to stop a pulse, it is important to identify the mechanisms and conditions responsible for the formation of such hot spots. To address this issue the new software tool ` Hotspot Editor' has been developed.

2019

Material erosion and migration in tokamaks

Jan Horacek, Richard Pitts

The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the susbsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasmafacing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined.

2005

Erosion, screening, and migration of tungsten in the JET divertor

The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the applied magnetic configuration in the divertor, e.g. close divertor with good geometrical screening of W or open divertor configuration with poor screening. JET equipped with the ITER-like wall (ILW) provided unique access to the net W erosion rate within a series of 151 subsequent H-mode discharges (magnetic field: B-t = 2.0 T, plasma current: I-p = 2.0 MA, auxiliary power P-aux = 12 MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5-6 x 10(26) D+ m(-2) in the semi-detached inner and attached outer divertor leg. The comparison of W spectroscopy in the intra-ELM and inter-ELM phases with post-mortem analysis of W marker tiles provides a set of gross and net W erosion values at the outer target plate. ERO code simulations are applied to both divertor leg conditions and reproduce the erosion/deposition pattern as well as confirm the high experimentally observed prompt W re-deposition factors of more than 95% in the intra- and inter-ELM phase of the unseeded deuterium H-mode plasma. Conclusions to the expected divertor conditions in ITER as well as to the JET operation in the DT plasma mixture are drawn on basis of this unique benchmark experiment.

2019

The software and hardware architecture of the real-time protection of in-vessel components in JET-ILW

2019

Erosion, screening, and migration of tungsten in the JET divertor

2019