# Tungsten control in long pulse operation: feedback from WEST to ITER

Patrick Maget, P Manas, J Morales, K Afonin, C Bourdelle, J Bucalossi, G Ciraolo, R Coelho, Y Corre, J Dominski, et al.

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Patrick Maget, P Manas, J Morales, K Afonin, C Bourdelle, et al. Tungsten control in long pulse operation: feedback from WEST to ITER. Plasma Physics and Controlled Fusion, 2025, Special Issue Featuring the Invited Talks from the 50th EPS Conference on Plasma Physics, 8-12 July 2024, Salamanca, Spain, 67 (4), pp.045005. 10.1088/1361-6587/ads0d hal-05519118v2

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Tungsten control in long pulse operation : feedback
from WEST to ITER

P. Maget, P. Manas, J. Morales, K. Afonin, C. Bourdelle, J.
Bucalossi, G. Ciraolo, R. Coelho¹, Y. Corre, J. Dominski², R.
Dumont, A. Ekedahl, N. Fedorczak, T. Fonghetti, A. Gallo, J.
Gaspar³, C. Guillemaut, A. Grosjean⁴, J. Gunn, J. Hillairet, E.
Lerche⁵, R. Lunsford², N. Rivals, S. Silburn⁶, E. Tsitrone, T.
Wauters⁷, the WEST team* and the EUROfusion Tokamak
Exploitation Team#

CEA, IRFM, F-13108 Saint Paul-lez-Durance, France.

¹ Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa, Portugal.

² Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA.

³ Aix Marseille University, CNRS, IUSTI, Marseille, France.

⁴ Department of Nuclear Engineering, University of Tennessee, Knoxville, Tennessee 37996, USA.

⁵ LPP-ERM-KMS, Association EUROFUSION - Belgian State, TEC partner, Brussels, Belgium.

⁶ CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB, United Kingdom.

⁷ ITER International Organization, F-13067 St. Paul lez Durance, France.

* see http://west.cea.fr/WESTteam,

# see author list in [Joffrin et al., 2024]

E-mail: patrick.maget@cea.fr

**Abstract.** The full W environment that is now foreseen for ITER puts strong emphasis on experimental results obtained in present devices in similar conditions. In this context, the WEST tokamak is well equipped to bring key contributions to the preparation of ITER operation, thanks to its capability to perform long pulses in dominant electron heating, torque-free scheme based on RF systems, and its ITER-grade actively cooled divertor. Recent results of interest cover the understanding of tungsten contamination and the evaluation of conditioning methods, the feedback on the current ramp-up phase, and the ageing of the ITER divertor components in L-mode attached condition. New directions are being explored, in particular low divertor temperature operation that should be the normal operating regime of ITER. We also report on tungsten peaking physics, and the role of light impurity content.
# 1. Introduction

The choice was made in 2024 to equip the ITER wall with tungsten elements, instead of beryllium. This high-Z material offers better perspectives in terms of tritium retention, dust production and resilience to disruption loads, but its contamination level in the plasma core should be extremely low, with $c_W = \langle n_W \rangle / \langle n_i \rangle$ less than few $10^{-5}$, in order to limit radiative losses [Pütterich et al., 2019]. Following this change of the wall material, a revision of the ITER research plan was initiated, with a strong emphasis on experimental results supported by modelling from existing tokamaks, in order to mitigate the risks and prepare for plasma operation. The inputs from tungsten wall devices such as WEST, Asdex-Upgrade and EAST are therefore key contributions for this preparation. With its actively cooled ITER-grade divertor and long pulse capability, its dominant electron heating scheme without external momentum torque, WEST is well equipped for addressing topics like tungsten contamination and transport, wall conditioning and component ageing, as well as scenario optimization in a tungsten environment.

In the present work, we review the status of our understanding of the WEST results for tungsten contamination level and its mitigation with Glow Discharge Boronization (GDB) and Impurity Powder Dropper (IPD) injections (section 2), and the plasma current ramp-up optimization with high-Z (tungsten) and low-Z (boron-nitride, noted as BN in the rest of the paper) tile limiters (section 3). We present the outcomes of a high fluence campaign in L-mode attached divertor condition, where the ingress of W-rich flakes progressively hampered plasma operation (section 4). This campaign based on the repetition of a typical one-minute long discharge was followed by the development of long pulse operation with a preliminary record at about 6 minutes plasma duration and more than 1 GJ of injected energy, that can be extended over 1000s from a scenario point of view. New challenges appeared over the GJ range, related with the out-gassing of remaining far-off inertial elements, and with the increased probability of unforeseen events that require a proper control actuator (section 5). Plasma operation at low divertor temperature, using nitrogen as the edge radiating species in a controlled way, has been developed in Lower Single Null (LSN), Double Null (DN) and compact magnetic configurations, leading to an X-Point Radiator (XPR) regime (section 6). In addition to cancelling tungsten sources from the divertor, as required for limiting its erosion, this scenario shows improved confinement and could be extended to low loop voltage. This opens the perspective of achieving a dedicated experimental campaign where the ageing of the ITER divertor components could be explored in the operating regime foreseen in ITER. Finally, we report on the characterization of core tungsten peaking, with evidences on the role of light impurities (section 7). In the concluding section 8, we summarize the WEST plan for future contributions to ITER preparation.
## 2. Tungsten contamination

The tungsten contamination largely dominates radiative losses in WEST [Goniche et al., 2022], and the evaluation of the relative contributions of each plasma facing components is a key issue for implementing strategies minimizing the tungsten penetration. Radiative losses are measured with a 16 lines bolometer system [Devynck et al., 2021], and the tungsten density can also be determined for hot enough plasmas with a VUV spectrometry diagnostic [Guirlet et al., 2019]. In the following, we will use the reduced WEST database, constructed by isolating plateaus of constant plasma current and additional heating power, as described in [Ostuni et al., 2022], and we select L-modes plateaus only.

### 2.1. *Radiated fraction and wall conditioning*

The tungsten contamination of the plasma originates from the sputtering of tungsten plasma facing components (PFC) by incoming ions, either from the main plasma species, from light impurities, or from tungsten atoms already ionized. The conditioning of these PFCs is therefore an important aspect to mention in the first place (for a more detailed description, see [Gallo et al., 2024]). The conditioning of the WEST vacuum vessel performed before each experimental campaign consists of a baking at 170°C for the experiments reported here (a baking at 200°C was performed for the last campaign of 2024), usually followed by a Glow Discharge Boronization (GDB) for typically 5h, using typically 10g of boron. Other boronizations are performed during the campaigns when appropriate, e.g. for increasing the power crossing the separatrix and facilitating H-mode transition [Vermare et al., 2021], or for characterizing the boronization effect on particle retention (an experiment performed in 2024 that will be reported in a future publication). On a more regular basis during the experimental campaign, glow discharge cleaning discharges are performed. The constraints imposed by GDB (no magnetic field, a minimum of half a day without plasma operation) can be avoided by the injection of solid boron powder, using an Impurity Powder Dropper (IPD). Such a system was installed on WEST within a collaboration with PPPL, and it does evidence a conditioning of the Plasma Facing Components (PFCs) as a cumulative effect of boron injection - from 10 to 60 mg/s during up to 35 s. Reports from these experiments can be found in [Afonin et al., 2024, Lunsford et al., 2024], as well as from the associated modelling in [Afonin et al., 2023], but we will only focus on GDB in the present work.

The evolution of the boronization effect can be quantified with respect to the cumulated energy injected into the plasma. In figure 1, we show the distribution of the total radiative fraction (ratio of the total radiated power to the total input power, ohmic plus auxiliary) in diverted plasmas as a function of the cumulated energy injected in the plasma from the last boronization performed, ignoring points between the very first boronization of the campaign and the second one. It can be argued indeed that the effect of the very first boronization after an opening of the vacuum vessel can depend on the work done during the shutdown as well as on the baking. The distribution is normalized
to its maximum value on each energy range. The first experimental campaigns of WEST had W coated tiles on limiters, while the following ones up to mid-2024 had BN tiles in the equatorial plane of the limiters. With W coated limiters, the typical cumulated energy scale to dissipate the boronization effect is around 10 MJ, and the radiative fraction saturates then between 80 and 90% in the ohmic plateaus. After the installation of the BN tiles it takes around 0.1 to 1 GJ to reach the saturation level. Despite different evolution in time, the radiative fraction reaches the saturation level similar to what is obtained with the W limiters. This different dynamics could the combination of two different physics mechanisms, one related to the dissipation of the boronization, and the other to the evolution of the BN tile surface. Indeed, the analysis of the BN surface after the completion of a first experimental campaign showed the presence of dark deposits containing W, and the evolution of the radiated power on the very first 250ms indicated a slow increase [Bucalossi et al., 2024, Guillemaut and the WEST team, 2024]. Note that a similar dissemination of W atoms over all plasma facing elements of the vessel was also reported on JET [Baron-Wiechec et al., 2015]. The impact of this tungsten migration on the life-time of the boronization effect in presence of the BN tiles is however quite difficult to evaluate quantitatively, and this link with the deposits is only a possible explanation at this stage.

Filtering plateaus with more than 2 MW of additional power, the total radiated fraction saturates around 60% with the W tiles, and slightly lower (in the range 50-60%) with the BN tiles. It takes less than 100 MJ of injected energy following a boronization to reach this range of radiated fraction in the main heating phase, so that it constitutes a robust and resilient value for usual plasma operation in WEST. Numerical experiments using SOLEDGE [Bufferand et al., 2015] and ERO2.0 [Romazanov et al., 2017] provided first physics explanations for this resilience [Ciraolo and the WEST team, 2024, Genova et al., 2024]. Based on a realistic background plasma from WEST experiments, and using for the first time a 3D description of the limiters, these simulations demonstrate that the resilience originates from a strong coupling between the radiation in the plasma core and the erosion at the PFCs (mainly the antenna limiters in these simulations): for a given magnetic geometry, increasing the input power increases the power going to the Scrape-off layer (SOL) and the erosion of the PFCs. In turns, this increases the plasma contamination and radiation, leading to a reduction of the power crossing the separatrix. The value of the radiated fraction in these simulations is then independent from the input power. However, it is directly linked with the level of light impurities that is assumed in the plasma background, since the tungsten sputtering is dominated by the impact of the low-Z ions [Di Genova, 2023]. A concentration of 3% of oxygen is used in the simulations, as a proxy for all low-Z ions that are identified on WEST: B, C, O are identified from visible spectroscopy in the plasma edge [Grosjean et al., 2023b], but they are also seen in deposits over the divertor, embedded in tungsten layers [Balden et al., 2021, Hakola et al., 2021].
<figure><img src="image_2.png" /><figcaption>**Figure 1.** Distribution of the total radiated fraction of diverted plasmas as a function of the energy injected in the plasma from the previous boronization, excluding the phase between the first and the second boronizations of the campaign. Left column: with W coated limiters, right column: with BN tiles in the mid-plane area. Top: ohmic phases, bottom: auxiliary power above 2 MW.</figcaption></figure>

## 2.2. Role of plasma to wall gaps

The distance of the plasma to the PFCs impacts this self-regulated system. This has been investigated by varying the RF antenna limiters position in SOLEDGE-ERO2.0 simulations: an important reduction of the radiative fraction was reported, and this result is in agreement with WEST experiment. This is shown in figure 2 where the simulations points are superimposed to experimental points taken from the reduced WEST database. These points are filtered in order to have a magnetic configuration at the upper divertor that is similar to that considered in the simulations, i.e. with a secondary upper X-point positioned in close vicinity (less than 3mm) to the upper divertor. These experimental points are also more than 100 MJ away from a boronization, and with an additional power larger than 2 MW. The effect of boronization is therefore lost, as shown in figure 1. This filtering evidences a key actuator for reducing the radiated fraction, but it should be noted that in the experiment, the plasma-antenna distance cannot be increased too much, due to the coupling requirements of the LHCD and ICRH systems. In other words, the plasma-antenna gap effect participates to the dispersion of the database points in figure 1, but the constraints of RF power coupling is limiting somewhat the freedom to increase this gap in experiments, thus making
<figure><img src="image_3.png" /><figcaption>**Figure 2.** Radiative fraction as a function of the plasma-antenna gap, from SOLEDGE-ERO2.0 simulations [Ciraolo and the WEST team, 2024] and from the WEST reduced database filtered with a lower X-point at $65\pm2$ mm, an upper X-point close to the upper divertor target (as in the modelling), and RF coupled additional power above 2 MW.</figcaption></figure>

the optimization between the injected and the radiated powers complex. Another important player is the magnetic configuration at the upper divertor, that was fixed for the simulation and experimental points of figure 2. This aspect was highlighted by SOLEDGE-ERO2.0 modelling, where the tungsten flux from the upper divertor was found to be weakly screened [Genova et al., 2021]. In order to disentangle the contributions of the upper and lower divertors, an extensive scan in the magnetic configuration is required, including the separation between the primary and secondary separatrices. A step in this direction has been performed in the first 2024 experimental campaign, and the associated modelling is on-going [Grosjean et al., 2023a].

### 3. Plasma current ramp-up

The plasma current ramp-up phase is quite critical in a tungsten device, in particular during the limiter phase where the screening of tungsten erosion is low, and the cooling rate is large with a peak to be crossed around 1.5 keV. Core plasma radiation due to tungsten frequently limited the rise of the electron temperature, leading to what was called a "cold branch regime" [Ostuni et al., 2022], with tearing modes hampering a safe transition to the main heating phase [Maget et al., 2022a]. The lack of an ECRH system at the time favoured the search for alternative methods, and we have successfully used light impurity seeding for getting a more robust current ramp-up phase. The reason for the improved situation has been clarified using integrated modelling with physics based turbulent transport model. It is explained by the ITG-TEM turbulence reduction,
leading to significant rise of the core electron temperature [Maget et al., 2022a]. This
rise leads to a more peaked current profile, which in turns contributes as well to the
turbulence reduction thanks to a higher s/q value (with s the magnetic shear and q the
safety factor) [Fourment et al., 2003]. In addition, tearing mode triggering is avoided
thanks to a better MHD stability (lower Δ' parameter [Furth et al., 1963]). This method
has some analogies with ECRH application, since the core electron temperature is
increased, but it has also some limits due to the degradation of the power balance
that leads to uncontrolled detachment above a critical seeding rate.

From mid-2020, the installation of BN tiles at the equatorial mid-plane of the inner
and outer limiters led to a substantial reduction of the radiative fraction during the
limiter phase, from 70% down to about 30% [Bucalossi et al., 2024]. This is shown in
figure 3, where the total radiative fraction is plotted as a function of the volume averaged
electron density for experiments with W limiters, and with BN limiters, in the limited
and diverted phases. The database retains points obtained after more than 100 MJ of
energy has been injected since the previous boronization (see figure 1). Points in W-
limited configuration have a plasma volume around 15 $m^3$, and the plasma volume for
points with the BN tiles is shown in the colorbar. The increase of the radiated fraction
with the plasma volume that can be seen at low density could be related with the
increased area of interaction with the limiter, possibly accessing the upper and lower
parts that were still covered by W coated tiles. For diverted plasmas, the radiative
fraction is comparable for the campaign with full tungsten limiters or BN limiters, so
that the benefit gained in the limiter phase does not extend to the diverted phase.

From the second experimental campaign of 2024, the BN limiters have been replaced
by bulk W tiles properly equipped with thermocouples and Langmuir probes (the probes
installed in the BN limiters were degraded at the beginning of their operation, probably
due to arcing). Dedicated experiments in limiter configuration will be performed on
the bulk tungsten tiles, in order to document tungsten contamination at various levels
of plasma current and density, and provide constraints for the modelling of the ITER
limiter phase.

4. ITER divertor ageing under High particle Fluence: formation of deposits and their impact of plasma operation

A High Fluence campaign was performed in 2023 in order to assess the ITER-grade
divertor ageing in L-mode attached plasma condition, reaching after 3h of cumulated
plasma time a particle fluence of about $5 \times 10^{26} D/m^2$, more than an ITER PFPO
pulse [Fedorczak et al., 2024]. This was achieved by repeating, without intermediate
boronization, a reference scenario in Lower Single Null configuration, with a typical
duration of one minute, at a magnetic field of 3.7T, plasma current 0.4 MA ($q_{95} = 5.5$),
volume averaged density $\langle n_e \rangle = 3 \times 10^{19} m^{-3}$, and an effective ion charge around
1.5. The LHCD power was 3.8 MW, with a plasma-antenna gap about 4cm, and
a secondary X-point far from the upper divertor, at about 17cm. The electron
<figure><img src="image_4.png" /><figcaption>**Figure 3.** Total radiated fraction as a function of the volume averaged density in the ohmic plateaus of the WEST reduced database, obtained more than 100 MJ of injected energy after a boronization, with W coated limiters and BN tiles, in the limiter and diverted phases. The points in limiter phase on BN tiles are coloured with the plasma volume.</figcaption></figure>

temperature at the lower divertor was around 20 eV at both the inner and outer strike points, and the temperature of ions impacting the divertor is expected to be larger [Kocan et al., 2008, Li et al., 2023], leading to a significant tungsten erosion as measured with visible spectroscopy [Fedorczak et al., 2024]. A kinetic modelling of the tungsten sputtering and redeposition [Cappelli et al., 2023] shows that more than 75% of the sputtered tungsten is redeposited nearby the strike point location [Fedorczak et al., 2024]. We observed the progressive formation of deposits, and the ejection of flakes (named UFOs for Unidentified Flying Object in the following), that resulted in short but sometimes intense peaks of the radiated power, as shown in figure 4 with a pulse in the beginning of the campaign, after one and a half hour and after two and a half hours of cumulated plasma time. These impurity radiation events also impacted the proportion of pulses reaching the nominal duration of 60s, as illustrated in the figure.

The UFOs events were analyzed in detail, each one being verified manually in order to avoid confusion with MHD events or other types of perturbation from plasma-wall interaction [Gaspar et al., 2024]. This analysis benefited from the comprehensive Infra-Red survey of the WEST divertor, combining wide angle and very high resolution views [Courtois et al., 2019].

The origin of the UFOs has been presented in detail in [Gaspar et al., 2024] and we recall here the main conclusions for convenience. UFOs are mostly generated by the detachment of flakes due to thermal stress at the initiation of the additional power, but about 7% are debris from a preceding disruption. These debris contribute to about 27% of the number of UFOs leading to disruption. The UFOs are coming from the High
<figure><img src="image_5.png" /><figcaption>**Figure 4.** Lower Hybrid (LHCD) power and total radiated power as a function of time, for a pulse at the beginning (top), after one and a half hour (middle) and after two and a half hours of cumulated plasma time (bottom), during the High Fluence campaign.</figcaption></figure>

Field Side (HFS) region of the divertor, where thick deposits were already identified in previous campaigns [Hakola et al., 2021, Balden et al., 2021]. This localization of the deposits is consistent with the evaluation of tungsten erosion from visible spectroscopy, that indicates an erosion rate ten times larger at the inner strike point than at the outer one [Fedorczak et al., 2024].

Minor UFOs allowing a quick return to the reference scenario represent about 81% of the events (class 1), while larger ones inducing first a change of plasma state (transition to a cold branch and MHD) and leading to a disruption after more than 200 ms (class 2) represent about 17%. A third class distinguishes UFOs leading to disruption within 200ms (class 3) that contribute to the remaining 2%. The proportion of the different classes is summarized in table 1. The disruptive UFOs (classes 2 and 3) are mainly triggered at the start of the additional heating. The first 2 classes reach a stationary regime of 6 UFOs per minute of plasma, after a cumulated plasma time of about 100 min (see fig. 5). The class 3 was still growing at the end of the 3h of cumulated plasma time, as evidenced by the rate of UFO production in the right plot of figure 5. This means that the risk of disruption by UFO will grow ultimately above the 19% that was recorded at the end of the campaign.
<table><caption>**Table 1.** Proportion of UFOs per class, identified during the high fluence campaign.</caption><tr><td>UFO impact</td><td>transient</td><td>disruptive</td><td>disruptive</td></tr><tr><td>UFO class</td><td>class 1</td><td>class 2</td><td>class 3</td></tr><tr><td>%</td><td>81%</td><td>17%</td><td>2%</td></tr></table>

<figure><img src="image_6.png" /><figcaption>**Figure 5.** Left: Cumulated number of UFOs as a function of the cumulated plasma time (bottom scale) or D fluence (top scale). Right: UFO production rate averaged over 50 minutes of plasma, for the 3 different classes (see text).</figcaption></figure>

The radiative losses produced by the UFOs constitute the main reason for their impact on the plasma scenario, and with the growing number of large UFOs, the distribution of the radiation peaks also evolves. This is illustrated in figure 6 where we have divided the campaign in 3 phases of equal number of UFOs (300, in order to have enough statistics for each phase). This segmentation of the experimental campaign corresponds to an initial phase covering the first 87 min of cumulated plasma time, then between 87 and 134 minutes for group 2, and finally between 134 and 173 minutes for group 3. While during the first phase the radiation peak rarely exceeds 2 MW and always remains below 3 MW, the proportion increases in this range during the second period with few events in the 4 MW range, and finally class 3 UFOs become more frequent in the last period with numerous events where the radiation peaks are between 3 and 4 MW.

This evolution of the distribution of radiation peaks implies that the required power that would be needed to compensate for the UFO radiation increases with time. When considering separately the contributions of the two first classes, we find that although the class 1 UFOs have a steady time averaged peak around 0.5 MW, the class 2 UFOs generate an average peak that evolves from about 1.5 MW up to 3 MW at the end of the high fluence campaign, as shown in figure 7.

The distribution of the tungsten content of the UFOs, is evaluated from the radiative power and effective charge evolutions. It gives an averaged mass of 14µg for class 1 UFOs and 64 µg for class 2, equivalent to a sphere of 100µm and 200µm
<figure><img src="image_7.png" /><figcaption>**Figure 6.** Distribution of the radiated power peak amplitude along the high fluence campaign, divided in 3 groups of equal number of UFOs (300).</figcaption></figure>

respectively (fig. 8). Few events of class 3 could be analyzed, and give a mass exceeding 100µg. If we relate the distribution of the W mass and the production rate of UFOs, we find that class 1 and 2 mobilize comparable W mass per minute of plasma operation, with a total of about 140µg/min. This does not include class 3 UFOs, for which we do not have enough statistics, but their contribution could be as large or even higher as the other classes.

The variation of the effective ion charge $Z_{eff}$, measured from Bremsstrahlung emission, and of the radiated power during UFOs allows evaluating as well the balance between tungsten and light impurities contents. It was necessary to restrict our analysis to UFOs occurring during stationary phases in order to get reliable results. We find that most UFOs are mainly composed of W, but some of them are rich in light impurities
<figure><img src="image_8.png" /><figcaption>**Figure 7.** Amplitude of the radiation peak due to the UFOs, averaged over a sliding time window of 50 min, as a function of the particle fluence and cumulated plasma time, for all UFOs and for classes 1 and 2 separately.</figcaption></figure>

<figure><img src="image_9.png" /><figcaption>**Figure 8.** Distribution of the UFO W mass. A mass of 10µg corresponds to a W sphere of diameter 100µm, and 81µg to a diameter of 200µm.</figcaption></figure>
<figure><img src="image_10.png" /><figcaption>Figure 9. Change of $Z_{eff}$ as a function of the change in the bulk radiated power during an UFO. The contribution of tungsten to the $Z_{eff}$ increase is indicated by the color.</figcaption></figure>

(figure 9). Note that in this subset of the UFO database, the points with $\Delta P_{rad} > 1.5$ MW are all belonging to class 2 UFOs.

The evolution of the tungsten flux from the divertor during an UFO of class 1 gives a good illustration of the radiative fraction resilience. Visible spectroscopy [Meyer et al., 2016] combined with Langmuir probes measurements [Dejarnac et al., 2021] at the lower divertor allows determining the tungsten flux $\Gamma_W$:

$$ \Gamma_W = 4\pi R_W(\lambda)(S/XB)_W(n_e, T_e) \quad (1) $$

with $\Gamma_W$ in (ions/m²/s), $R_W(\lambda)$ the calibrated spectral radiance (Ph/m²/s/sr) and $(S/XB_W)$ the number of ionization per photon for the atomic transition corresponding to the W line at 400.9 nm, that depends on the local electron density and temperature. For a typical class 1 UFO, the W flux from the divertor is reduced by about 50% on both the inner and outer strike points, as shown in figure 10, before a return to the original state. This reduction is most likely a consequence of the strong reduction of the power going to the divertor, following the radiation peak of the UFO. Note that the similar amplitude of the power going to the divertor and the radiated power results from the radiated power fraction being around 50%.

For UFOs leading to a disruption, the impact on electron temperature is too large to avoid a bifurcation of the plasma state. Integrated simulations in interpretative mode, with plasma density and temperature taken from the experiment, but plasma current and LHCD deposition computed dynamically, have been performed with the HFPS suite of codes [Romanelli et al., 2014] for a class 1 and a class 2 UFOs. The LHCD deposition is computed with the simplified METIS model described in [Artaud et al., 2018] that has been coupled with the Heating and Current Drive module of the simulator. Results shown in figure 11 demonstrate that the large temperature excursion for class 2 UFO leads to a large reduction of the internal inductance, with a minimum safety factor crossing 3/2 and rising just below 2. The linear resistive MHD stability of the $n = 1$
<figure><img src="image_11.png" /><figcaption>**Figure 10.** Time trace of the radiated power and power going to the divertor (top), and variation of the tungsten flux measured at the Inner (ISP) and Outer (OSP) Strike Points (bottom) during a class 1 UFO.</figcaption></figure>

mode, computed with the MARS code [Bondeson et al., 1992] using the EQSTABIL workflow [Romanelli et al., 2020], confirms the destabilization of the (2,1) mode that is observed, and that leads to the disruption. The computation has been done at a fixed plasma resistivity ($\eta = 10^{-5}$ in MARS units), but the temperature fall increases the core resistivity by a factor $\sim 28$ so that the growth rate shown in the figure should be multiplied by a factor of $\sim 7$ during the UFO (the growth rate of resistive modes scales as $\eta^{3/5}$). For the class 1 UFO, the linear growth rate is barely affected. Note that the growth rate of the mode is always positive because the computation does not include diamagnetic effects that are known to be stabilizing [Meshcheriakov et al., 2012]. Due to the strong non-linearity of LHCD power deposition with temperature, the peak of the LHCD deposition moves outwards, further reinforcing the removal of the heat and current drive source from the plasma core. But the main trend of the safety factor evolution is maintained even without this LHCD specific effect, as shown in the figure with a simulation at fixed LHCD heating and current drive.

The evolution of the profiles illustrate this divergence of the scenario. The temperature, LHCD current and safety factor profiles before and after the class 2 UFO are shown in the right plot of figure 11. Although the temperature starts to recover 400ms after the UFO, the safety factor profile continues evolving away from the initial equilibrium. This situation can be interpreted as a metastable plasma regime that deviates for large perturbations towards an MHD unstable state, while for class 1 UFOs the plasma discharges remains in a resilient regime.

It is interesting to compare the WEST UFOs to observations reported in the JET-ILW experiments (with a tungsten divertor and beryllium walls). Despite the high heat
<figure><img src="image_12.png" /><figcaption>**Figure 11.** Left: time trace of the core electron temperature, internal inductance, minimum of the safety factor, linear growth rate of the resistive *n* = 1 mode and position of the maximum of the LHCD current, for a class 1 (#58358, $t_{UFO} = 30.1s$) and a class 2 (#58463, $t_{UFO} = 31.6s$) UFOs, modelled with the Integrated Simulation tool HFPS. The evolution of the internal inductance and safety factor in a simulation with a fixed LHCD deposition is also shown. Right: electron temperature, LHCD current density and safety factor profiles before during and after the UFO for the class 2 case.</figcaption></figure>

flux on the divertor targets, and the realization of a high fluence campaign reaching $2 \times 10^{25}$ D/m$^2$ [Brezinsek et al., 2019], impurity influx events are very different from WEST observations. In JET-ILW, Transient Impurity Events (TIE) originating from tungsten dusts were observed and analysed along several experimental campaigns. The mass of these W dusts is comparable to what we find on WEST, and it is equivalent to spheres of diameter 140-180 µm [Sertoli et al., 2014, Sertoli et al., 2015, Flanagan et al., 2014]. These tungsten dusts were presumably originating from the delamination of the W coating on the divertor, and were decreasing in number along the JET-ILW campaigns, therefore resulting in a negative net production rate [Widdowson et al., 2019]. In contrast with WEST observations, these dusts are therefore not correlated with the
growth and flaking of deposits. Such deposits are observed in the high field side region of the JET-ILW divertor, but they consist mainly of beryllium coming from the erosion of the first wall [Mayer et al., 2016]. Following some JET disruptions during which the plasma moved downwards, mobilisation of large numbers of particles from the horizontal surfaces on top of the outer divertor (JET tiles 8 and 9) was observed. These particles would then fall in to the divertor and cause impurity influxes and even disruptions at early times in the following pulses, in a similar way to the post-disruption WEST UFOs. The frequent use of strike point sweeping on JET-ILW to reduce the heat loads on the tungsten tiles [Silburn et al., 2017] may have mitigated to a certain extent the formation of deposits close to the erosion area. The divertor geometry at the inner strike point could also be a mitigation factor for the impact on plasma operation, with an horizontal inner divertor target on WEST that has a larger viewing angle towards the plasma than the vertical one of JET-ILW (and ITER). Finally, the WEST divertor is actively cooled at 70°C, whereas the JET-ILW divertor was inertial. The rise of the surface temperature of the inertial component during plasma operation could mitigate the thermal stress that leads to flake detachment on WEST. These three differences (material of the first wall, divertor geometry and divertor cooling), unrelated with the nature of the tungsten component that was tested, lead therefore to completely different conclusions on the operational impact of the ageing of the component.

**5. Toward Long Pulse and High Injected Energy Operation**

Following the high fluence campaign, Long Pulse Operation scenario was developed with a double control system on the loop voltage and the plasma current, using respectively the central solenoid and the LHCD power, comparable to what was developed for Tore Supra [Wijnands et al., 1997]. Operational diagrams were first derived using integrated simulations, in order to target relevant triplets for plasma current, density and LHCD power [Fonghetti and team, 2024, Fonghetti et al., 2024]. The progress in terms of duration and injected energy during the High Fluence campaign of 2023 (in the minute range), and the 2024 Long Pulse Operation (extending above 6 minutes) is illustrated in figure 12.

An MHD-stable scenario with a loop voltage $V_{loop} = 3mV$ was obtained at a plasma current of 270 kA and $B=3.7$ T. The maximum duration that was achieved is of 6 min 4 s, with 1.15 GJ of injected energy (figure 13), but this loop voltage allows a potential extension of the pulse duration above 1000s. This achievement in duration and energy puts WEST at a comparable level to Tore Supra records [Houtte et al., 2004, Litaudon et al., 2023]. While plasma duration was progressively extended, a water leak in one of the LHCD protection limiter interrupted this program, and it will be continued in the coming campaigns.

The lessons from these initial developments above the GJ energy range were twofold: first, we encountered a limitation in the injected energy due to the outgassing of remaining far-off inertial elements, with the evidence of a progressive conditioning
<figure><img src="image_13.png" /><figcaption>**Figure 12.** LHCD Energy injected per pulse as a function of the plasma duration for the 2023 campaign (including High Fluence discharges), and for the first campaign of 2024.</figcaption></figure>

<figure><img src="image_14.png" /><figcaption>**Figure 13.** Time trace of the plasma current, loop voltage, volume averaged electron density, LHCD power and radiated power for the record pulse in terms of duration and injected energy of the first campaign of 2024.</figcaption></figure>

from pulse to pulse (figure 14). Such an outgassing event can be seen at the end the record pulse shown in figure 13, with an uncontrolled density rise leading to an increase of the LHCD power required for sustaining the target plasma current. The plasma facing element causing this outgassing could not be determined, but this limitation is a soft one, imposing the rate of progress in the injected energy. Second, we had to face the rising probability of unpredictable events, such as UFOs and spontaneous Internal Transport Barrier (ITB) triggering, both generally leading to MHD instabilities and disruption. The sensitivity to MHD events, as well as to ITB formation, is related to the presumably hollow safety factor profile that is predicted by Integrated Modelling simulations [Fonghetti and team, 2024, Fonghetti et al., 2024],
<figure><img src="image_15.png" /><figcaption>**Figure 14.** LHCD Energy injected per pulse as a function of the LHCD energy injected from the beginning of the first campaign of 2024. Outgassing events are indicated by larger red circles. The threshold in injected energy leading to an out-gassing event increases progressively.</figcaption></figure>

and was already an issue for non-inductive discharges on Tore Supra [Houtte et al., 2004, Zabiégo et al., 2001, Peysson and the TORE SUPRA Team, 2000, Maget et al., 2004, Maget et al., 2005, Maget et al., 2006, Maget et al., 2007]. The on-going installation of an ECRH system [Delpech et al., 2023] will provide the missing actuator to mitigate these events in future campaigns, first by driving co-current on axis and avoiding hollow current profiles, and by compensating if needed the transient increase of core radiation following UFOs.

**6. Low divertor temperature operation**

The preservation of plasma facing components from excessive erosion requires a low plasma temperature at the interface with the divertor plates. Such conditions are met when edge radiation cools down the plasma below typically 5 eV, and this is obtained generally by injecting light impurities [Loarte et al., 1998, Guillemaut et al., 2014, Kallenbach et al., 2015, Komm et al., 2019, Dimitrova et al., 2020, Bernert et al., 2020, Bernert et al., 2023] - although a pure deuterium fuelling can also produce a similar transition [Dachicourt et al., 2013]. A radiative ring appears then above the divertor plates when the injection rate of light impurities exceeds a critical threshold [Bernert et al., 2023]. On WEST, the formation of this X-Point Radiator (XPR) regime is controlled using as sensor an interferometry line crossing the X-point region, and the nitrogen injection rate is used as actuator. The transition seen on the interferometry line is extremely fast (about 1 µs), as shown in the example in figure 15, where the control was applied for more than 12s. The target temperature falls below 5 eV but the particle flux is still important, so that the plasma is not detached [Rivals et al., 2024]. In addition, the upstream profiles are nearly unchanged, thus preserving the condition for a good
LHCD coupling. This example also shows that the electron and ion confinement are improved from the moment nitrogen is injected, as shown by the electron temperature, neutron rate and energy content increases, without relation with the XPR transition itself. A similar confinement improvement was reported in WEST [Yang et al., 2020], Asdex-Upgrade [Rathgeber et al., 2010] or JET [Bonanomi et al., 2018] during nitrogen seeding, as well as in WEST IPD experiments [Bodner et al., 2022, Bodner et al., 2024]. The tungsten concentration in the core is reduced, as expected due to the cancellation of the divertor source, but the reduction is only in the 40% range in this example. The fact that tungsten contamination is not drastically reduced when the source from the divertor is extinguished brings us back to the question of the plasma sources. It means that the main chamber sources are still dominant in this case, in line with numerical modelling results indicating that, even in a high divertor temperature regime, the main contributors are the antenna limiters and not the divertor plates [Genova et al., 2024]. A global picture on the tungsten erosion and migration processes can be drawn from the combination of experimental and numerical works: visible spectroscopy shows that the tungsten erosion is maximal at the inner strike point, ten times lower at the outer strike point, and twenty times lower at the outer limiters [Fedorczak et al., 2024]. But the divertor sources are well screened, and promptly redeposited, while only 20% of the antenna sources are redeposited [Fedorczak et al., 2024], and are weakly screened [Genova et al., 2024]. As a result, the main concern for the flaking of deposits and the generation of UFOs comes from the divertor, while the main concern for the plasma contamination comes from the outboard limiters.

The modelling of the XPR transition on WEST with SOLEDGE-EIRENE has been initiated, with a scan in the nitrogen seeding rate that shows a two steps process: a sharp fall of the temperature at the divertor at a critical seeding rate, while the formation of the radiative ring occurs at a higher seeding rate [Rivals et al., 2024]. This plasma state is found to be prone to a radiative collapse, and several mechanisms are identified that could lead to a stable regime. A first mechanism is an increase of the power entering the Scrape-off Layer (SOL), which is verified in the experiment thanks to the decrease of the core radiation. This indicates that core-edge integration is key for the sustainably of this regime. A second parameter is the impurity species, with a better stability obtained when replacing nitrogen by oxygen, probably due to their different radiation functions. More analyses on this topic are currently ongoing, but this shows that the impurity mix is an important aspect to consider as well (see [Rivals et al., 2024]).

An overview of the main results obtained with the XPR regime is shown in figure 16, with global parameters from the reduced database plotted as a function of the effective ion charge. Note that all points in the figure are in the XPR regime. Due to the failure of part of the bolometry lines of sight in the last experimental campaign, where Double-Null (DN) and Compact Radiative Divertor (CRD) configurations were tested, the bolometry data could not yet be fully validated and we have therefore ignored at this stage the corresponding data (bulk radiated power and tungsten profiles) for this campaign. With the light impurity seeding, radiative losses in the confined plasma are reduced from 50%
<figure><img src="image_16.png" /></figure>

**Figure 15.** LHCD power, line integrated density at the cord crossing the X-point region, core electron temperature, neutron rate, energy content, bulk radiated power, maximum electron temperature at the Outer Strike Point (OSP) and core tungsten concentration, with (#58089) and without (#58072) nitrogen seeding. The transition time to the XPR regime is indicated with a vertical dashed line.
<figure><img src="image_17.png" /><figcaption>**Figure 16.** Left column : bulk radiated fractions, core tungsten peaking and concentration. Right column : plasma volume, neutron rate, and core electron temperature, as a function of $Z_{eff}$ for the LSN, CRD and DN cases. All points are in the XPR regime, with $q_{95} \in [4, 5]$ and $\langle n_e \rangle \in [3, 4] \times 10^{19} \text{ m}^{-3}$.</figcaption></figure>

down to about 40%, with a plasma-antenna gap in this database in the range 2-2.5 cm. The core tungsten peaking is also reduced with $Z_{eff}$, but the core concentration remains approximately constant, thus leading to an apparent contradiction with the decrease of the radiative fraction. In fact, the strong increase of the electron temperature has a direct impact on the cooling rate of tungsten: with $T_{e0}$ going from about 2.5 keV to more than 4 keV, the cooling rate is reduced by a factor 2 (see [Ostuni et al., 2022]), thus explaining the reduction of the bulk plasma radiation.

The XPR regime has been developed in Lower Single Null (LSN) as well as in alternative magnetic configuration like the Double-Null (DN), allowing the investigation of different W sources [Moiraf et al., 2025], or in a Compact configuration (Compact Radiative Divertor or CRD [Lunt et al., 2023]) with the X-point positioned at the divertor target, allowing for a gain in the plasma volume (about 10% in WEST, see figure 16). In these alternative magnetic configurations, the performances appear to be rather similar to the LSN ones (see fig. 16). These recent experiments are now being studied in detail, and the results will be reported in future publications.

The development of the XPR regime on WEST opens the route to the study of the ITER divertor ageing at low electron temperature. The existing scenario has been evolved to lower plasma current where it could be maintained MHD-stable at a loop
<figure><img src="image_18.png" /><figcaption>**Figure 17.** Poloidal beta, loop voltage, normalized beta and internal inductance, as a function of the $Z_{eff}$ for the WEST reduced database of XPR pulses. The points are colored with $q_{95}$.</figcaption></figure>

voltage around 100 mV (see figure 17), allowing for its extension to about one minute discharges. The scenario that is obtained has a high poloidal beta around 1.5 and internal inductance $l_i \sim 1.4$, with a normalized beta $\beta_N \sim 0.9$. A future high fluence campaign at low divertor temperature could address questions like the growth rate of W deposits, but also the frequency and impact of UFOs from existing deposits. It will document as well the basis for a DEMO compact divertor, making this option attractive if an XPR regime on horizontal tiles can prevent UFOs.

## 7. Tungsten peaking

The relatively high tungsten content of the plasma requires a careful optimization of the electron heating source distribution to compensate for tungsten induced radiative losses. The off-axis deposition of LHCD power, or an insufficient confinement of the fast minority ions in Ion Cyclotron Resonance Heating scheme, can lead to a power imbalance in the core region and a radiative collapse [Ostuni et al., 2022, Goniche et al., 2022, Maget et al., 2023]. The threat of radiative collapses is also affected by the radial distribution of tungsten inside the plasma, due to the combination of collisional and turbulent transport. Over most of the plasma radius, turbulent transport is dominant and leads to rather flat profiles, both in experiment and in theoretical predictions. In the plasma core, where collisional transport can be dominant, the risk of strong tungsten peaking is avoided by the absence of external momentum sources [Yang et al., 2020]. This leads to a distribution of $n_W(0)/n_W(0.35)$, computed by bolometry inversion, that stays close to unity over the reduced WEST database, as shown in figure 18. We have filtered out points with $T_{e0} < 2$ keV to exclude the so-called cold branch cases for which the physical understanding is still incomplete [Bourdelle et al., 2023]. For the hot branch
<figure><img src="image_19.png" /><figcaption>**Figure 18.** Normalized distribution of the tungsten peaking $n_W(0)/n_W(0.35)$ from the WEST reduced database with $T_{e0} > 2$ keV (top). The average LHCD and ICRH powers are shown in the bottom plot.</figcaption></figure>

subset, a bump in the tungsten peaking distribution around 2 can be related to plasmas with a heating scheme combining LHCD and ICRH, as shown in the bottom plot. This is reminiscent of the larger W peaking observed in an ICRH pulse where a detailed analysis, combining stand-alone computations and integrated modelling, pointed to the role of the plasma rotation induced by fast ion losses [Maget et al., 2023].

Despite the modest W peaking observed on a statistical basis, we do have observations of large transient peaking ending in MHD or disruption, during radiative collapses [Ostuni et al., 2022, Maget et al., 2023]. The mechanism involved is the weakening of the ion thermal screening effect as the core temperature gradient falls. We illustrate this effect in figure 19 for the case studied in [Ostuni et al., 2022], by plotting the normalized tungsten gradient as a function of the normalized ion temperature gradient, around the radial position $\rho_{tor} = 0.25$. The bolometry inversion (black circles) indicates an increase of the tungsten peaking up to $R/L_{nW} \approx 10$ as the normalized ion temperature gradient decreases down to $\sim 1$. The neoclassical prediction (computed with the FACIT code [Maget et al., 2020a, Maget et al., 2020b, Maget et al., 2022b, Fajardo et al., 2022]) for a pure D plasma is however much more pessimistic, with a W peaking that should rise up to $R/L_{nW} \approx 60$. This observation can be understood by considering the dependencies of the collisional diffusion and convection terms with the main ion mass and charge. We have indeed the following dependencies in the Pfirsh-Schlüter regime that applies in this case:

$$D_{PS} \propto Z_i A_i^{1/2} \qquad (2)$$

$$V_{PS} \propto [Z_W \partial_r \ln n_i - (Z_i + Z_W (C_0^a - 1)) \partial_r \ln T_i] \qquad (3)$$

where $Z_i$ and $A_i$ are the charge and atomic number of the main ion, $Z_W$ the charge of tungsten ions, $n_i$ and $T_i$ the main ion density and temperature and $C_0^a \approx 1.5$ (see
[Maget et al., 2020a]). Note that in this context of pure LHCD heating, where toroidal rotation should be negligible, the poloidal distribution of tungsten ions is expected to be completely homogeneous at steady state, as shown in [Maget et al., 2020a]. Since in general $Z_W \gg Z_i$, the pinch velocity will be weakly dependent on the main ion charge, but the diffusion coefficient will increase as $Z_i A_i^{1/2}$. We have verified this by computing the tungsten peaking at steady-state using FACIT, assuming that the interaction between tungsten and light ions of D and N can be treated additively, i.e. the tungsten flux is computed as:

$$ \Gamma_W \approx \sum_{\text{ions } k \neq W} \Gamma_{W \to k} \qquad (4) $$

where the terms $\Gamma_{W \to k}$ are computed separately. The resulting diffusion and convection coefficient at steady-state ($\Gamma_W = 0$) are shown in figure 19 for a nitrogen concentration consistent with the effective charge of the plasma $Z_{eff} = 2.4$. We find that the diffusion coefficient is indeed well approximated using the effective ion charge and atomic number, and that the pinch velocity is weakly affected. As a consequence, the tungsten peaking $R/L_{nW} = -RV_W/D_W$ is strongly reduced and becomes consistent with experimental observations.

This mechanism mitigates tungsten peaking by increasing the collisional diffusion in presence of light impurities, and it plays favorably when the ion thermal screening effect drives a strong inward pinch. But it can also be deleterious when the ion thermal screening effect drives a strong outward pinch ($R/L_{nW} < 0$), as expected in the pedestal of ITER [Fajardo et al., 2024]. In this case indeed, it reduces the strength of the barrier isolating the plasma core from tungsten ions circulating in the separatrix region, as illustrated in figure 20.

The consequence of light impurity seeding on tungsten peaking is also visible in normal plasma operation, although with less dramatic effects. We will mention here two particular situations.

* The first one concerns the ohmic phase that is particularly critical in a W environment when an external electron heat source is not used. In the first campaigns of WEST, with W coated limiters, the situation has been improved by using nitrogen seeding, as reported earlier [Maget et al., 2022a]. The response of W peaking during this phase has been recently investigated using the XGC code [Ku et al., 2016, Dominski et al., 2024a], combining collisional and turbulent transport. In this case, it is found that nitrogen seeding does reduce the collisional peaking, but it also extends the core region where turbulent transport is low. The combination of the two effects leads to a more peaked tungsten profile in case of nitrogen seeding, in agreement with bolometry inversion [Dominski et al., 2024b].

* The second case concerns the XPR regime discussed in section 6, for which a reduction of the tungsten peaking is observed as well (see figure 16). The combination of collisional transport using FACIT, and turbulent transport using TGLF-SAT2 [Staebler et al., 2021], recovers this reduction of the tungsten
<figure><img src="image_20.png" /></figure>

**Figure 19.** Collisional diffusion (top), convection (middle) and local peaking (bottom) of tungsten, as a function of the normalized ion temperature gradient peaking, during a radiative collapse in WEST. The values are computed with FACIT for a pure D plasma ($Z_{eff} = 1$) and for a plasma containing nitrogen with $Z_{eff} = 2.4$. The diffusion coefficient rescaled with a factor $Z_{eff}A_{eff}^{1/2}$ is shown in the top plot (dashed black line), and the tungsten peaking deduced from bolometry inversion is shown in the bottom plot.

peaking with $Z_{eff}$ over the WEST reduced database of XPR discharges [Manas and team, 2024].

In these two situations, the deviation from a flat tungsten profile is moderate, but a proper understanding of these trends should allow more accurate predictions, and help designing optimized scenarios for WEST, as well as providing useful inputs for ITER preparation.

## 8. Conclusion and future plans

The WEST tokamak equipped ITER grade W divertor explores tungsten related issues on long durations, providing unique inputs to the preparation of ITER operation, in support of the new baseline with a full tungsten main chamber wall. The results obtained over the last 4 campaigns form the basis for a quantitative modelling and the validation
<figure><img src="image_21.png" /><figcaption>**Figure 20.** Local peaking of tungsten, as a function of the normalized ion temperature gradient peaking, extrapolated from the WEST case of figure 19 to large ion temperature gradient using eq. (2-3). While light impurity seeding mitigates tungsten peaking during radiative collapses, it is also expected to reduce the hollowness of tungsten profile in the pedestal region.</figcaption></figure>

of physics models describing tungsten source and transport. They show the resilience of the radiative fraction, standing around 50% on average, and the role of plasma-wall gaps that can reduce it down to 30%. Wall conditioning with boronization is shown to reduce the plasma radiation, but for a bounded cumulated energy injected in the vacuum vessel: in a full tungsten environment (including therefore limiters), the radiative fraction recovers a saturated value around 90% in ohmic plasmas, and around 60% with more than 2 MW of auxiliary heating, when the injected energy, cumulated since the last boronization, exceeds about 10 MJ. IPD conditioning provides a more flexible approach that will be further explored in future campaigns. Recent experiments on deuterium retention in boron layers are being analysed and will provide valuable inputs for the compatibility of this conditioning technique with tritium retention in ITER. Successful tests performed early 2024 on Ion Cyclotron Wall Conditioning (ICWC) will be further developed in future campaigns, adding new reference points to the existing database on the capabilities of this technique for ITER.

The limiter phase on W tiles and the ramp-up scenario improvement using light impurity seeding gives encouraging results, and this activity will be renewed at the end of 2024 with the installation of bulk W tiles at the main chamber limiters. The ageing of the ITER-grade divertor under high particle fluence in L-mode attached conditions (with $T_e \approx 20$eV, above the tungsten sputtering threshold) has been investigated and proved to form thick deposits around the inner strike point, with a significant and growing impact of detached flakes on plasma operation. On this topic, it has to be outlined that these results contrast with the JET-ILW experience, where tungsten erosion at the divertor did not result in the production of tungsten rich UFOs. This contrasting result
can be related with three main differences between the two experiments: the nature
of the first wall (W with BN tiles at mid-plane versus Be), the divertor geometry at
the inner strike point (open, nearly horizontal on WEST versus more closed and on
vertical tiles for the high fluence campaign of JET-ILW), and the divertor cooling only
present on WEST. The preparation for plasma operation with a mitigation of UFO
impact is on-going, with the development of a spatio-temporal convolutional neural
network able to detect UFOs from Infra-Red movies, with a success rate over 95% and
a first application in the second experimental campaign of 2024. In parallel, innovative
techniques based on laser ablation for removing the deposits on the divertor are being
tested for future application during shut-down phases. The development of a divertor-
preserving scenario, with a low electron temperature at the targets as expected in
ITER, has been undertaken under various magnetic configurations, providing a unique
experimental tool to evaluate the relative contributions of the upper and lower divertors,
as well as antenna limiters, to the tungsten contamination. The incomplete removal
of the tungsten contamination in this radiative regime indicates, in agreement with
numerical simulations, that the outboard limiters remain the dominant concern for the
plasma contamination, while the divertor is a major concern for the flaking of deposits
and the generation of UFOs. This low divertor temperature regime has been extended to
low loop voltage, allowing the evaluation of tungsten erosion/redeposition, and impact
of deposits, in a future high particle fluence campaign dedicated to the normal operation
regime of ITER. Plasma operation in the XPR regime with horizontal tiles as in WEST
could also allow for a compact divertor design in reactor-class devices such as DEMO.

Finally, core tungsten transport in conditions of dominant electron heating without external momentum torque is shown to be preserved from excessive density peaking, at the exception of radiative collapse events. Experimental results show that light impurity seeding is mitigating the tungsten peaking, as it increases the neoclassical diffusion. This mechanism would also mitigates as a side-by-side effect the tungsten pump-out in situations of large ion temperature screening, as in ITER pedestal for example.

Continuous integrated modelling effort using physics based turbulent transport
models is carried out, as well as SOLEDGE simulations with EIRENE and ERO
2.0 for clarifying the origin and impact of W sources, the role of gaps, and the
physics mechanism behind the XPR regime. This modelling validation is key
to extrapolate to ITER operation and DEMO design in conditions of dominant
electron heating without external momentum torque. The experiment program
of WEST will benefit from end 2024 from the installation of an ECRH system
[Delpech et al., 2023], recognized as an instrumental tool in metallic devices for dealing
with the multiple non-linear couplings introduced by heavy impurity core radiation,
such as radiative collapses, current diffusion and MHD stability, as exemplified in
WEST experiments. It will also relax the constraint on the plasma-antenna gap that
was pushed to low values for improving the coupling of LHCD and ICRH systems,
while increasing the tungsten contamination. The exploration of H-mode operational
domain, started few years ago [Vermare et al., 2021], will also benefit from this increased
flexibility. The completion of the Thomson diagnostic system installation covering the full plasma radius [Chelihi et al., 2022, Colledani et al., 2023, Colledani et al., 2024, Bouchand et al., 2024, Carole et al., 2024], from end 2024 as well, will enhance the outputs of the experimental program, with the pedestal physics exploration, and the quest for a causality chain in the coupling between the edge and core parameters [Bourdelle et al., 2023].

## Acknowledgments

This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them.

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