The particle exhaust system plays a pivotal role in fusion reactors and is essential for ensuring both the feasibility and sustained operation of the fusion reaction. For the successful development of such a system, density control is of great importance and some key design parameters include the neutral gas pressure and the resulting particle fluxes. This study presents a simplified conductance-based model for estimating neutral gas pressure distributions in the particle exhaust system of fusion reactors, focusing specifically on the sub-divertor region. In the proposed model, the pumping region is represented as an interconnected set of reservoirs and channels. Mass conservation and conductance relations, appropriate for all flow regimes, are applied. The model was benchmarked against complex 3D DIVGAS simulations across representative operating scenarios of the Wendelstein 7-X (W7-X) stellarator. Despite geometric simplifications, the model is capable of predicting pressure values at several key locations inside the particle exhaust area of W7-X, as well as various types of particle fluxes. The developed model is computationally efficient for large-scale parametric studies, exhibiting an average deviation of approximately 20%, which indicates reasonable predictive accuracy considering the model simplifications and the flow problem complexity. Its application may assist early-stage engineering design, pumping performance improvement, and operational planning for W7-X and other future fusion reactors.

The International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source (IFMIF-DONES) will be an installation capable of qualifying materials for use in future fusion power reactors. To do so, a linear accelerator will deliver high-intensity deuteron beam to a liquid Li loop, creating a flow of neutrons that will produce material damage equivalent to that expected in a fusion reactor. The vacuum system must maintain high vacuum conditions during operation with high reliability. For this reason, a careful study of the vacuum has been assessed and implemented in the accelerator project. Vacuum simulations have been performed in order to have a first estimation of the expected pressure profile, and documentation is being produced to standardize the subsystem. Additionally, valuable knowledge from the prototype installation, IFMIF/Engineering Design and Engineering Validation Activities (EVEDA), is being taking into account in engineering design activities. Moreover, two vacuum related prototypes are being manufactured; the Multipurpose Vacuum Accident Scenarios (MuVacAS) prototype will enable experimental studies of air inrush in the last section of the accelerator; and the Quick Disconnecting System (QDS) prototype is being developed to study the feasibility of the remote handling of the interface located between the accelerator beam line and the target chamber. The paper presents the status of the design of the vacuum subsystem of the accelerator and discusses the future challenges.

The qualification of in-vessel components for a fusion power plant requires a test environment with a high flux of 14 MeV neutrons over a sufficiently large surface and volume. Performance testing and qualification of the complex design and technologies of fusion nuclear components is needed, in particular that of the tritium breeding blanket (BB). Testing in relevant conditions over a relevant time will also allow gaining the necessary confidence regarding the build-up and control of tritium inventories inside the BB, which will be an important radioactive source. An option of such a volumetric neutron source (VNS) is a beam-driven tokamak. A feasibility study of the main machine components and associated plant systems is described in this article. The machine has a major radius of 2.53 m, a single-null divertor configuration, and four tangential 120 keV beamlines that generate a fusion power of approximately 30 MW and provide current drive for a steady-state plasma scenario. The plasma is small with a minor radius of a = 0.55 m to maximize the neutron wall load, up to 0.5 MW/m(2), similar to what is targeted in ITER. Approximately 25 m(2) are available for blanket testing including 4 port plugs, which offer flexibility regarding the test module operating conditions and the implementation of instrumentation. Given the small plasma, much of the tokamak's volume is made up by the neutron shielding structures that are similarly sized as in ITER. To reduce the construction risk, ITER-like concepts were adopted for many components. In some cases, however, lessons learned from ITER led to the development of customized or innovative concepts. Due to the modest fusion power the plasma will burn <1 kg of tritium per year, which can be provided from external sources.

An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

Recently the design of the divertor tokamak test (DTT) Facility divertor has been modified and consolidated. The new divertor design presents significant geometrical differences compared to the previous ITER-like one, including the presence of a more flattened dome shape. This paper presents a complete 3D numerical analysis of the neutral gas dynamics inside the DTT subdivertor area for the latest divertor design. The analysis has been performed based on the direct simulation Monte Carlo method by applying the DIVGAS simulator code. SOLEDGE2D-EIRENE plasma simulations have been performed for a deuterium plasma scenario at the maximum additional power in partially detached condition achieved by neon impurity seeding and the extracted information about the neutral particles has been imposed as incoming boundary conditions. The pumping efficiency of the DTT divertor is examined by considering various cases with respect to the pumping probability and the effect of the toroidal and poloidal leakages is quantified. The results show that a significant percentage of the incoming flux of neutrals returns back to the plasma site through the entry gaps (60% for deuterium and 40% for neon), and, consequentially, only a small percentage (∼2%–15%) of the incoming flux can be pumped out from the system. The toroidal leakages affect significantly the pumping performance of the divertor causing a significant decrease in the pumped flux (and also in the pressure at the pumping opening) about 37%–47% and 43%–56% for deuterium and neon respectively. It is discussed how many pumping ports are needed depending on the achievable pumping performance per port. The number can be reduced by closing the toroidal gaps. The analysis shows that enlarging the poloidal gaps by a factor of two causes a significant increase in the poloidal flux losses by a factor 1.7. It is also illustrated how the presence of the cooling pipes leads to conductance losses.

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98( p,y ) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’ q -profile) were established on AUG and characterised concerning their plasma performance.

In the DEMO Pre-Concept Design Phase, a novel fuel cycle architecture has been developed that aims at reducing the tritium inventory of the plant. In contrast to previous designs, the central role thereby is taken by an additional direct internal recycling loop. By already separating a significant proportion of the fuel at the pumping duct and providing this purified gas almost directly to the injection systems, the load on the tritium plant as well as the overall plant tritium inventory is reduced substantially. This paper summarizes the activities conducted in regard of integration and presents the results (aka the key design integration issues) for the design and feasibility of a pumping concept based on tritium direct recycling. The focus is on the aspects of integration of such a fuel separation function with the main technologies involved. In the functional position foreseen, these environment conditions mainly are space constraints, strong magnetic field, and the limited amount and restrictions in occupation of suitable ports. Two technical implementations for providing the fuel separation function have been identified: A metal foil pump relying on the physical phenomena of superpermeation of hydrogen through a metal membrane is able to provide continuous gas transfer. A multi-stage cryopump utilizing gas specific cryogenic trapping on several stages. Relying on gas binding, it only can operate intermittently. Within this activity, these variants are specifically reviewed for their suitability also to provide criteria for a future down selection.

An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L–H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced D α H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts.

In the present work, the role of plasma facing components protection in driving the EU-DEMO design will be reviewed, focusing on steady-state and, especially, on transients. This work encompasses both the first wall (FW) as well as the divertor. In fact, while the ITER divertor heat removal technology has been adopted, the ITER FW concept has been shown in the past years to be inadequate for EU-DEMO. This is due to the higher foreseen irradiation damage level, which requires structural materials (like Eurofer) able to withstand more than 5 dpa of neutron damage. This solution, however, limits the tolerable steady-state heat flux to ~1 MW/m2, i.e. a factor 3–4 below the ITER specifications. For this reason, poloidally and toroidally discontinuous protection limiters are implemented in EU-DEMO. Their role consists in reducing the heat load on the FW due to charged particles, during steady state and, more importantly, during planned and off-normal plasma transients. Concerning the divertor configuration, EU-DEMO currently assumes an ITER-like, lower single null (LSN) divertor, with seeded impurities for the dissipation of the power. However, this concept has been shown by numerous simulations in the past years to be marginal during steady-state (where a detached divertor is necessary to maintain the heat flux below the technological limit and to avoid excessive erosion) and unable to withstand some relevant transients, such as large ELMs and accidental loss of detachment. Various concepts, deviating from the ITER design, are currently under investigation to mitigate such risks, for example in-vessel coils for strike point sweeping in case of reattachment, as well as alternative divertor configurations. Finally, a broader discussion on the impact of divertor protection on the overall machine design is presented.

The outgassing rates of various ceramic materials were systematically investigated. This paper intercompares the most relevant ceramic material classes used in vacuum technology, namely, oxide (five different types investigated) and non-oxide based ceramics, including nitrides (four different types) and carbides (three different types). For this purpose, the Outgassing Measurement Apparatus, which uses the difference method, was applied. Besides time, also, temperature dependence of outgassing was studied to check whether a previous heat treatment of the ceramics is necessary. Additionally, the outgassing species were qualitatively determined by a quadrupole mass spectrometer. In total, relatively large differences were found between the investigated ceramic materials for their initial outgassing rates as well as for their behavior at elevated temperatures. All investigated types of carbide ceramics showed very low outgassing rates of less than 3 × 10−8 Pa m3/(s m2) after 100 h at room temperature, whereas rather high outgassing rates of about 10−5 Pa m3/(s m2) were found for magnesia as well as the tested nitride ceramic types. Thus, depending on the ceramic material, a heat treatment is strongly recommended before use in ultrahigh vacuum applications.

IFMIF-DONES (International Fusion Materials Irradiation Facility - DEMO Oriented Neutron Source) is currently being developed in the frame of the EUROfusion Early Neutron Source work package (WPENS) and will be an installation for fusion material testing, that will generate a flux of neutrons of 10(18) m(-2)s(-1) with a broad peak at 14 MeV by Li(d,xn) nuclear reactions thanks to a 40 MeV deuteron beam colliding on a liquid Li flow. The accelerator system is in charge of providing such high energy deuterons in order to produce the expected neutron flux. The High Energy Beam Transport line (HEBT) is the last subsystem of the accelerator and its main functions are to guide the deuteron beam towards the Lithium target and to shape it by the use of magnetic elements to the reference beam footprint at the Lithium Target. The present work summarizes the current status of the HEBT design, including beam dynamics, vacuum, radioprotection, diagnostics and remote handling studies performed, along with the layout and integration of the line.

The aim of this paper is to present a signal processing algorithm that, applied to the raw Locked Mode signal, allows us to obtain a disruption indicator in principle exploitable on different tokamaks. A common definition of such an indicator for different machines would facilitate the development of portable systems for disruption prediction, which is becoming of increasingly importance for the next tokamak generations. Moreover, the indicator allows us to overcome some intrinsic problems in the diagnostic system such as drift and offset. The behavior of the proposed indicator as disruption predictor, based on crossing optimized thresholds of the signal amplitude, has been analyzed using data of both JET and ASDEX Upgrade experiments. A thorough analysis of the disruption prediction performance shows how the indicator is able to recover some missed and tardy detections of the raw signal. Moreover, it intervenes and corrects premature or even wrong alarms due to, e.g., drifts and/or offsets.

In ITER and DEMO, various control objectives related to plasma control must be simultaneously achieved by the plasma control system (PCS), in both normal operation as well as off-normal conditions. The PCS must act on off-normal events and deviations from the target scenario, since certain sequences (chains) of events can precede disruptions. It is important that these decisions are made while maintaining a coherent prioritization between the real-time control tasks to ensure high-performance operation. In this paper, a generic architecture for task-based integrated plasma control is proposed. The architecture is characterized by the separation of state estimation, event detection, decisions and task execution among different algorithms, with standardized signal interfaces. Central to the architecture are a plasma state monitor and supervisory controller. In the plasma state monitor, discrete events in the continuous-valued plasma state are modeled using finite state machines. This provides a high-level representation of the plasma state. The supervisory controller coordinates the execution of multiple plasma control tasks by assigning task priorities, based on the finite states of the plasma and the pulse schedule. These algorithms were implemented on the TCV digital control system and integrated with actuator resource management and existing state estimation algorithms and controllers. The plasma state monitor on TCV can track a multitude of plasma events, related to plasma current, rotating and locked neoclassical tearing modes, and position displacements. In TCV experiments on simultaneous control of plasma pressure, safety factor profile and NTMs using electron cyclotron heating (ECH) and current drive (ECCD), the supervisory controller assigns priorities to the relevant control tasks. The tasks are then executed by feedback controllers and actuator allocation management. This work forms a significant step forward in the ongoing integration of control capabilities in experiments on TCV, in support of tokamak reactor operation.

In this paper we validate the finite element code RAPLICASOL, which models radiofrequency wave propagation in edge plasmas near ICRF antennas, against calculations with the TOPICA code. We compare the output of both codes for the ASDEX Upgrade 2-strap antenna, and for a 4-strap WEST-like antenna. Although RAPLICASOL requires considerably fewer computational resources than TOPICA, we find that the predicted quantities of experimental interest (including reflection coefficients, coupling resistances, - and -matrix entries, optimal matching settings, and even radiofrequency electric fields) are in good agreement provided we are careful to use the same geometry in both codes.

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q(95) = 5.5, beta(N) <= 2.8) at low density. Higher installed electron cyclotron resonance heating power <= 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with P-sep/R <= 11 MW m(-1) with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently H-H98(y,H-2) <= 1.1. This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of T-i and E-r allow for inter ELM transport analysis confirming that E-r is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to n/n(GW) <= 0.7 have been characterised using beta-feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle-measured for the first time-or the cross-phase angle of T-e and n(e) fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow T-e profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvenic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.

Ammonia formation was studied in a series of dedicated nitrogen seeded H-mode discharges at ASDEX Upgrade.The evolution of ammonia formation was investigated with a reference phase before the seeding, and a long, stable flat-top nitrogen-seeded phase.It was monitored with divertor spectroscopy and analysis of the exhaust gas.The amount of the detected ammonia increased continuously over the course of five discharges with the same nitrogen seeding rate.The same trend was observed in the nitrogen density in the core plasma, as measured by charge exchange recombination spectroscopy and other signals, linked to the effects of nitrogen seeding.The results show that the rate of ammonia formation exhibited the same trend as the nitrogen density in the plasma.This density, in turn, was strongly influenced by the nitrogen wall inventory.The spatial distribution of the detected ammonia suggests that a significant contribution to the net ammonia formation is made in plasma-shaded areas, through surface reactions of neutral species.

In vacuum applications, polymers often need to be used in spite of their relatively high outgassing rates, e.g., as sealing or spacer material. For some polymers—like Viton®, which is used for O-rings—data on outgassing rates can be found in literature, while for others, in particular, the most relevant ones that can be used at higher temperatures, only few results are available. Therefore, different high-temperature resistant polymers were measured at the Outgassing Measurement Apparatus, which uses the difference method. These polymers were polyimides like Kapton®, Vespel®, and Meldin®, as well as a polyetheretherketone and a polybenzimidazole. In order to investigate the temperature behavior of the outgassing of the polymers, the measurements were performed at room temperature, 100 °C, and 200 °C. The outgassing species were determined by a mass spectrometer. In addition, the dependence of the outgassing on the material thickness was analyzed. The paper shows that the outgassing rates of the examined polymers are higher than 10−4 (Pa m3)/(s m2) after 10 h at room temperature, except for the thin Kapton samples. Thus, further treatment of the materials will be necessary depending on the application. It is shown that a vacuum bake-out at not less than 200 °C is favorable. The main outgassing species is water, as expected, but also different hydrocarbons with higher mass can be seen for the polyimides. Concerning different thicknesses, the outgassing rate of a thin sample generally decreases more steeply than the outgassing rate of the thicker sample. Nevertheless, the steepest decrease was found for the thickest Kapton sample, with a difference of nearly factor 500 between 10 and 100 h at room temperature.

In ASDEX Upgrade, the propagation of cold pulses induced by type-I edge localized modes (ELMs) is studied using electron cyclotron emission measurements, in a dataset of plasmas with moderate triangularity. It is found that the edge safety factor or the plasma current are the main determining parameters for the inward penetration of the Te perturbations. With increasing plasma current the ELM penetration is more shallow in spite of the stronger ELMs. Estimates of the heat pulse diffusivity show that the corresponding transport is too large to be representative of the inter-ELM phase. Ergodization of the plasma edge during ELMs is a possible explanation for the observed properties of the cold pulse propagation, which is qualitatively consistent with non-linear magneto-hydro-dynamic simulations.