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The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

This work presents the use of a collisional-radiative model to infer plasma parameters from 2D emissivities of several deuterium Balmer lines. The emissivities were obtained by MANTIS, an absolutely calibrated, ten-camera imaging polychromator with < 5 m m spatial resolution, up to 800 Hz frame rate, viewing the lower divertor tangentially. Our analysis of those image frames generates 2D maps of plasma parameters such as electron density, temperature, neutral atomic density and the reaction rates for ionisation, recombination and charge exchange as a function of time. The analysis is compared and validated against a SOLPS-ITER simulation accounting for drifts. The results are compared against the simulation in 2D, radial and poloidal profiles to probe the numerous effects of particle transport in the scrape-off layer, in particular in the approach to detached divertor leg conditions. The inferred inner divertor leg radial profiles of the electron density and temperature were consistent with the SOLPS-ITER predictions. A significant transport of particles to the private flux region is found experimentally, that is not captured in the simulation. The simulation diverges from the experiment at the outer divertor target, where the plasma emission appears to be consistent with the emission driven by plasma–molecule interactions. Our analysis also shows prospects for aiding the power exhaust control efforts by potentially providing an optical tracking of the particle balance in the divertor.

Plasma density is one of the key quantities that need to be controlled in real-time as it scales directly with fusion power and, if left uncontrolled, density limits can be reached leading to a disruption. On ASDEX Upgrade (AUG), the real-time measurements are the line-integrated density, measured by the interferometers, and the average density derived from the bremsstrahlung measured by spectroscopy. For control, these measurements are used to reconstruct the radial density profile using an extended Kalman filter (EKF). However, in discharges where ion cyclotron resonance heating (ICRH) is used, the measurements from the interferometers are corrupted and the reconstructed density is false. In this paper, the existing EKF implementation is improved, implemented and experimentally verified on AUG. The new EKF includes a new particle transport model in the prediction model RAPDENS as well as a new representation of ionization and recombination. Furthermore, an algorithm was introduced that is capable of detecting the corrupt diagnostics; this algorithm is based on the rate of change of the innovation residual. The changes to the RAPDENS observer resulted in better density reconstruction in ICRH discharges where corrupt measurement occur. The new version has been implemented on the real-time control system at AUG and functions properly in ICRH discharges.

In contemporary magnetic confinement devices, the density distribution is sensed with interferometers and actuated with feedback controlled gas injection and open-loop pellet injection. This is at variance with the density control for ITER and DEMO, that will depend mainly on pellet injection as an actuator in feed-back control. This paper presents recent developments in state estimation and control of the electron density profile for ITER using relevant sensors and actuators. As a first step, Thomson scattering is included in an existing dynamic state observer. Second, model predictive control is developed as a strategy to regulate the density profile while avoiding limits associated with the total density (Greenwald limit) or gradients in the density distribution (e.g. neoclassical impurity transport). Simulations show that high quality density profile estimation can be achieved with Thomson Scattering and that the controller is capable of regulating the distribution as desired.

In magnetic confinement thermonuclear fusion the exhaust of heat and particles from the core remains a major challenge. Heat and particles leaving the core are transported via open magnetic field lines to a region of the reactor wall, called the divertor. Unabated, the heat and particle fluxes may become intolerable and damage the divertor. Controlled ‘plasma detachment’, a regime characterized by both a large reduction in plasma pressure and temperature at the divertor target, is required to reduce fluxes onto the divertor. Here we report a systematic approach towards achieving this critical need through feedback control of impurity emission front locations and its experimental demonstration. Our approach comprises a combination of real-time plasma diagnostic utilization, dynamic characterization of the plasma in proximity to the divertor, and efficient, reliable offline feedback controller design., The exhaust of heat and particles is an important challenge in future nuclear fusion devices. Here the authors report the use of carbon emission as indicator for plasma detachment in a tokamak and its real-time feedback control.

In this paper we present a novel algorithm to extract the optical plasma boundary and radiation front for detached divertor plasmas. We show that reliable detection of the divertor leg and radiation front is possible using lightweight image processing tools. Using a non-tomographic approach, the detected divertor leg and radiation front can be mapped to the poloidal plane. This approach is fast and accurate enough for real-time control purposes, allowing in particular real-time plasma shape and detachment control, and post-shot detachment physics and dynamics analysis.

A model-based approach to real-time reconstruction of the particle density profile in tokamak plasmas is presented, based on a dynamic state estimator. Traditionally, the density profile is reconstructed in real-time by solving an ill-conditioned inversion problem using a measurement at a single point in time. This approach is sensitive to diagnostics errors and failure. The inclusion of a dynamic model in a real-time estimation algorithm allows for reliable reconstruction despite diagnostic errors. Predictive simulations show that the model can reproduce the density evolution of discharges on TCV and ASDEX-Upgrade after tuning of a few parameters. Offline reconstructions using experimental data from TCV show accurate estimation of the density profile and show examples of fault detection of interferometry signals.

Transient electron temperature measurements of a step power experiment at W7-AS are reassessed by direct comparison of the up- and downward responses of the electron temperature. The analysis shows that the response at some distance to the center behaves linearly and the model predicted responses based on a power-dependent diffusion coefficient that vary from the measured step responses.

The Magnum-PSI facility is unique in its ability to produce and even exceed the heat and particle fluxes expected in the divertor of a fusion reactor, combined with good access to the plasma-material interaction region for diagnostics and relatively easy sample manipulation. In addition, it is possible to study the effects of transient heat loads on a plasma-facing surface, similar to those expected during so called Edge Localized Modes. By virtue of a newly installed superconducting magnet, Magnum-PSI can now maintain these conditions for hours on end for truly long term tests of candidate plasma facing materials. The electron density and temperature in the plasma beam center as a function of different magnetic fields up to 1.6 T, gas flow and source current are determined: particle fluxes greater than 10 25 m −2 s −1 and heat fluxes of up to 50 MW m −2 are obtained. Linear regression and artificial neural network analysis have been used to gain insight in the general behavior of plasma conditions as a function of these machine settings. The plasma conditions during transient plasma heat loading have also been determined. These capabilities are now being exploited to reach fluence of up to 10 30 particles m −2 at ITER-relevant conditions, equivalent to a significant fraction of the divertor service lifetime for the first time.

The leading candidate for impurity seeding in ITER is currently nitrogen. To date, there have only been a few studies on the plasma chemistry driven by N2/H2 seeding and its effect on the molecular-activated recombination of incoming atomic hydrogen ions in a detached-like scenario. Numerical simulations are needed to provide insights into such mechanisms. The numerous plasma chemical reactions that may occur in such an environment cannot be entirely included in a 2- or 3-dimensional code such as Eirene. A complete global plasma model, implemented with more than 100 plasma chemical equations and 20 species, has been set up on the basis of the Plasimo code. This study shows that two main nitrogen-including recombination reaction paths are dominant, i.e. the ion conversion of NH followed by dissociative recombination, and the proton transfer between and N2, producing N2H+. These two processes are referred to as N-MAR (nitrogen molecular-activated recombination) and have subsequently been implemented in Eunomia, which is a spatially resolved Monte Carlo code designed to simulate the neutral inventory in linear plasma machines such as Pilot-PSI and Magnum-PSI. To study the effect of N2 on the overall recombination, three studies have been set up, and from a defined puffing location with a constant total seeding rate of H2 + N2, three N2 ratios were simulated, i.e. 0%, 5% and 10%. The parameter monitored is the density of atomic hydrogen, being the final hydrogenic product of any recombination mechanism in the scenario considered. The difference in H density between the 0% case and the 10% case is about a factor of three. The importance of NH as an electron donor is highlighted, and the N-MAR reaction routes are confirmed to enhance the conversion of ions to neutrals, making the heat loads to the divertor plate more tolerable. This work is a further step towards a full understanding of the role of N2-H2 molecules in a detached divertor plasma.

The heat flux is one of the key parameter used to quantify and understand transport in fusion devices. In this paper, a new method is introduced to calculate the heat flux including its confidence with high accuracy based on perturbed measurements such as the electron temperature. The new method is based on ideal filtering to optimally reduce the noise contributions on the measurements and piece-wise polynomial approximations to calculate the time derivative. Both methods are necessary to arrive at a heat flux and effective diffusion coefficient with high accuracy. The new methodology is applied to a measurement example using electron cyclotron resonance heating block-wave modulation at the Large Helical Device showing the merit of the newly developed methodology.

Old and recent experiments show that there is a direct response to the heating power of transport observed in modulated ECH experiments both in tokamaks and stellarators. This is most apparent for modulated experiments in the Large Helical Device (LHD) and in Wendelstein 7 advanced stellarator (W7-AS). In this paper we show that: (1) this power dependence can be reproduced by linear models and as such hysteresis (in flux) has no relationship to hysteresis as defined in the literature; (2) observations of 'hysteresis' (in flux) and a direct response to power can be perfectly reproduced by introducing an error in the estimated deposition profile as long as the errors redistribute the heat over a large radius; (3) non-local models depending directly on the heating power can also explain the experimentally observed Lissajous curves (hysteresis); (4) how non-locality and deposition errors can be recognized in experiments and how they affect estimates of transport coefficients; (5) from a linear perturbation transport experiment, it is not possible to discern deposition errors from non-local fast transport components (mathematically equivalent). However, when studied over different operating points non-linear-non-local transport models can be derived which should be distinguishable from errors in the deposition profile. To show all this, transport needs to be analyzed by separating the transport in a slow (diffusive) time-scale and a fast (heating/non-local) time-scale, which can only be done in the presence of perturbations.

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 (q95 = 5.5, ) 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 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 . 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 and Er allow for inter ELM transport analysis confirming that Er 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 have been characterised using -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 and 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 profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvénic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. "Article signat per més de 100 autors/es: H. Meyer, for the AUG Team: D. Aguiam, C. Angioni, C.G. Albert, N. Arden, R. Arredondo Parra, O. Asunta, M. de Baar, M. Balden, V. Bandaru, K. Behler, A. Bergmann, J. Bernardo, M. Bernert, A. Biancalani, R. Bilato, G. Birkenmeier, T.C. Blanken, V. Bobkov, A. Bock, T. Bolzonella, A. Bortolon, B. Böswirth, C. Bottereau, A. Bottino, H. van den Brand, S. Brezinsek, D. Brida, F. Brochard, C. Bruhn, J. Buchanan, A. Buhler, A. Burckhart, Y. Camenen, D. Carlton, M. Carr, D. Carralero, C. Castaldo, M. Cavedon, C. Cazzaniga, S. Ceccuzzi, C. Challis, A. Chankin, S. Chapman, C. Cianfarani, F. Clairet, S. Coda, R. Coelho, J.W. Coenen, L. Colas, G.D. Conway, S. Costea, D.P. Coster, T.B. Cote, A. Creely, G. Croci, G. Cseh, A. Czarnecka, I. Cziegler, O. D'Arcangelo, P. David, C. Day, R. Delogu, P. de Marné, S.S. Denk, P. Denner, M. Dibon, A. Di Siena, D. Douai, A. Drenik, R. Drube, M. Dunne, B.P. Duval, R. Dux, T. Eich, S. Elgeti, K. Engelhardt, B. Erdös, I. Erofeev, B. Esposito, E. Fable, M. Faitsch, U. Fantz, H. Faugel, I. Faust, F. Felici, J. Ferreira, S. Fietz, A. Figuereido, R. Fischer, O. Ford, L. Frassinetti, S. Freethy, M. Fröschle, G. Fuchert, J.C. Fuchs, H. Fünfgelder, K. Galazka, J. Galdon-Quiroga, A. Gallo, Y. Gao, S. Garavaglia, A. Garcia-Carrasco, M. Garcia-Muñoz, B. Geiger, L. Giannone, L. Gil, E. Giovannozzi, C. Gleason-González, S. Glöggler, M. Gobbin, T. Görler, I. Gomez Ortiz, J. Gonzalez Martin, T. Goodman, G. Gorini, D. Gradic, A. Gräter, G. Granucci, H. Greuner, M. Griener, M. Groth, A. Gude, S. Günter, L. Guimarais, G. Haas, A.H. Hakola, C. Ham, T. Happel, N. den Harder, G.F. Harrer, J. Harrison, V. Hauer, T. Hayward-Schneider, C.C. Hegna, B. Heinemann, S. Heinzel, T. Hellsten, S. Henderson, P. Hennequin, A. Herrmann, M.F. Heyn, E. Heyn, F. Hitzler, J. Hobirk, K. Höfler, M. Hölzl, T. Höschen, J.H. Holm, C. Hopf, W.A. Hornsby, L. Horvath, A. Houben, A. Huber, V. Igochine, T. Ilkei, I. Ivanova-Stanik, W. Jacob, A.S. Jacobsen, F. Janky, A. Jansen van Vuuren, A. Jardin, F. Jaulmes, F. Jenko, T. Jensen, E. Joffrin, C.-P. Käsemann, A. Kallenbach, S. Kálvin, M. Kantor, A. Kappatou, O. Kardaun, J. Karhunen, S. Kasilov,, Y. Kazakov, W. Kernbichler, A. Kirk, S. Kjer Hansen, V. Klevarova, G. Kocsis, A. Köhn, M. Koubiti, K. Krieger, A. Krivska, A. Krämer-Flecken, O. Kudlacek, T. Kurki-Suonio, B. Kurzan, B. Labit, K. Lackner, F. Laggner, P.T. Lang, P. Lauber, A. Lebschy, N. Leuthold, M. Li, O. Linder, B. Lipschultz, F. Liu, Y. Liu, A. Lohs, Z. Lu, T. Luda di Cortemiglia, N.C. Luhmann, R. Lunsford, T. Lunt, A. Lyssoivan, T. Maceina, J. Madsen, R. Maggiora, H. Maier, O. Maj, J. Mailloux, R. Maingi, E. Maljaars, P. Manas, A. Mancini, A. Manhard, M.-E. Manso, P. Mantica, M. Mantsinen, P. Manz, M. Maraschek, C. Martens, P. Martin, L. Marrelli, A. Martitsch, M. Mayer, D. Mazon, P.J. McCarthy, R. McDermott, H. Meister, A. Medvedeva, R. Merkel, A. Merle, V. Mertens, D. Meshcheriakov, O. Meyer, J. Miettunen, D. Milanesio, F. Mink, A. Mlynek, F. Monaco, C. Moon, F. Nabais, A. Nemes-Czopf, G. Neu, R. Neu, A.H. Nielsen, S.K. Nielsen, V. Nikolaeva, M. Nocente, J.-M. Noterdaeme, I. Novikau, S. Nowak, M. Oberkofler, M. Oberparleiter, R. Ochoukov, T. Odstrcil, J. Olsen, F. Orain, F. Palermo, O. Pan, G. Papp, I. Paradela Perez, A. Pau, G. Pautasso, F. Penzel, P. Petersson, J. Pinzón Acosta, P. Piovesan, C. Piron, R. Pitts, U. Plank, B. Plaum, B. Ploeckl, V. Plyusnin, G. Pokol, E. Poli, L. Porte, S. Potzel, D. Prisiazhniuk, T. Pütterich, M. Ramisch, J. Rasmussen, G.A. Rattá, S. Ratynskaia, G. Raupp, G.L. Ravera, D. Réfy, M. Reich, F. Reimold, D. Reiser, T. Ribeiro, J. Riesch, R. Riedl, D. Rittich, J.F. Rivero-Rodriguez, G. Rocchi, M. Rodriguez-Ramos, V. Rohde, A. Ross1, M. Rott, M. Rubel, D. Ryan, F. Ryter, S. Saarelma, M. Salewski, A. Salmi, L. Sanchis-Sanchez, J. Santos, O. Sauter, A. Scarabosio, G. Schall, K. Schmid, O. Schmitz, P.A. Schneider, R. Schrittwieser, M. Schubert, T. Schwarz-Selinger, J. Schweinzer, B. Scott, T. Semer, E. Seliunin, M. Sertoli, A. Shabbir, A. Shalpegin, L. Shao, S. Sharapov, G. Sias, M. Siccinio, B. Sieglin, A. Sigalov, A. Silva, C. Silva, D. Silvagni, P. Simon, J. Simpson, E. Smigelskis, A. Snicker, C. Sommariva, C. Sozzi, M. Spolaore, A. Stegmeir, M. Stejner, J. Stober, U. Stroth, E. Strumberger, G. Suarez, H.-J. Sun, W. Suttrop, E. Sytova, T. Szepesi, B. Tál, T. Tala, G. Tardini, M. Tardocchi, M. Teschke, D. Terranova, W. Tierens, E. Thorén, D. Told, P. Tolias, O. Tudisco, W. Treutterer, E. Trier, M. Tripský, M. Valisa, M. Valovic, B. Vanovac, D. van Vugt, S. Varoutis, G. Verdoolaege, N. Vianello, J. Vicente, T. Vierle, E. Viezzer, U. von Toussaint, D. Wagner, N. Wang, X. Wang, M. Weiland, A.E. White, S. Wiesen, M. Willensdorfer, B. Wiringer, M. Wischmeier, R. Wolf, E. Wolfrum, L. Xiang, Q. Yang, Z. Yang, Q. Yu, R. Zagórski, I. Zammuto, W. Zhang, M. van Zeeland, T. Zehetbauer, M. Zilker, S. Zoletnik, H. Zohm and the EUROfusion MST1 Team55

The interaction of intense plasma impacting on the wall of a fusion reactor is an area of high and increasing importance in the development of electricity production from nuclear fusion. In the Magnum-PSI linear device, an axial magnetic field confines a high density, low temperature plasma produced by a wall stabilized DC cascaded arc into an intense magnetized plasma beam directed onto a target. The experiment has shown its capability to reach conditions that enable fundamental studies of plasma-surface interactions in the regime relevant for fusion reactors such as ITER: 1023–1025 m−2s−1 hydrogen plasma flux densities at 1–5 eV for tens of seconds by using conventional electromagnets. Recently the machine was upgraded with a superconducting magnet, enabling steady-state magnetic fields up to 2.5 T, expanding the operational space to high fluence capabilities for the first time. Also the diagnostic suite has been expanded by a new 4-channel resistive bolometer array and ion beam analysis techniques for surface analysis after plasma exposure of the target. A novel collective Thomson scattering system has been developed and will be implemented on Magnum-PSI. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.

This work presents a novel, real-time capable, 10-channel Multispectral Advanced Narrowband Tokamak Imaging System installed on the TCV tokamak, MANTIS. Software and hardware requirements are presented together with the complete system architecture. The image quality of the system is assessed with emphasis on effects resulting from the narrowband interference filters. Some filters are found to create internal reflection images that are correlated with the filters' reflection coefficient. This was measured for selected filters where significant absorption (up to 65% within similar to 70 nm of the filter center) was measured. The majority of this was attributed to the filter's design, and several filters' performance is compared. Tailored real-time algorithms exploiting the system's capabilities are presented together with benchmarks comparing polling and event based synchronization. The real-time performance is demonstrated with a density ramp discharge performed on TCV. The behavior of spectral lines' emission from different plasma species and their interpretation are qualitatively described.

For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX Upgrade however, the desired frequency range in which the islands are to be tracked (100 Hz-10 kHz) is much larger than is possible with a PLL. In this contribution, an extended Kalman filter (EKF) and an unscented Kalman filter (UKF) are proposed for real-time frequency, phase and amplitude tracking of sinusoidal signals, based on noisy measurements. Compared to PLLs, the EKF and UKF are able to track sinusoidal signals in a much larger frequency range. The filters are applied on synthetic data and on experimental data from the TEXTOR and TCV tokamaks, from which we conclude that the UKF can be useful for real-time control of magnetic islands on ASDEX Upgrade. (C) 2013 Elsevier B.V. All rights reserved.

The ITER tokamak has a modular design, with port plugs, blanket modules and divertor cassettes. This set-up allows for maintenance of diagnostics, heating systems and first wall elements. The maintenance can be done in situ, or in the Hot Cell. Safe and effective remote handling (RH) will be ensured by the RH requirements and standards. Compliance is verified through remote handling compatibility assessments at the ITER Design Review milestones. The Remote Handling Study Centre at FOM Institute DIFFER is created to study ITER RH maintenance processes at different levels of complexity, from relatively simple situational awareness checks using snap-shots in the CAD system, time studies using virtual reality (VR) animations, to extensive operational sequence validation with multiple operators in real-time. The multi-operator facility mimics an RH work-cell as presently foreseen in the ITER RH control room. Novel VR technology is used to create a realistic setting in which a team of RH operators can interact with virtual ITER environments. A physics engine is used to emulate real-time contact interaction as to provide realistic haptic feed-back. Complex interactions between the RH operators and the control room system software are tested. RH task performance is quantified and operational resource usage estimated. The article provides a description and lessons learned from a recent study on replacement of the Steering Mirror Assembly on the ECRH (Electron Cyclotron Resonance Heating) Upper Launcher port plug.

Control of Neoclassical Tearing Modes (NTMs) requires an accurate and low latency detection of the mode position. For a burning H-mode ITER plasma, simulations are conducted for both ECE detected via the equatorial port plug and along the line-of-sight of the ECCD launchers. Simulated ECE is detected using synthetic radiometers, with settings chosen to meet the required accuracy. A video bandwidth of 2 kHz is used which allows for an intermediate frequency bandwidth of BIF = 400 MHz for ECE detected via the equatorial port plug. For ECE detected via the ECCD line-of-sight, an intermediate frequency bandwidth of 1.5 GHz and 1 GHz for the 2/1 and 3/2 NTM respectively suffices for accurate location detection. For both ECE systems, the latency requirements for NTM suppression are fulfilled.

A basic requirement for ITER equipment to meet is a high level of reliability, because ITER operation time is precious and radioactive operation leaves limited scope for repair. In order to reduce the risk of failure during ITER operation an effective risk mitigation strategy is necessary. This paper presents such strategy for the ITER electron cyclotron upper port launcher (ECUPL). A preliminary ECUPL risk analysis identifies possible failure modes. A probabilistic risk assessment quantifies the risk of failure using a 4 x 4 impact-likelihood matrix. Impact is quantified through technical, cost and schedule elements. Likelihood depends on the risk mitigation strategy adopted. A cost benefit analysis determines the most cost effective risk mitigation strategy. An essential element in risk mitigation is the testing of equipment prior to installation on the ITER machine. This paper argues the need for low- and highpower millimetre wave tests carried out on the fully assembled ECUPL It presents a conceptual design for a dedicated on-site test bed that can handle 2 of 8 microwave beams at 2 MW long pulse operation. (C) 2010 Elsevier B.V. All rights reserved.

Model-based real-time scenario monitoring for transient event prediction with RAPTOR

A new approach to real-time estimation and feedback control of the particle density profile in tokamak plasmas is presented, based on ideas from Kalman filtering and H∞ robust control synthesis. Traditionally, the density profile is reconstructed in real-time by solving an inversion problem using a measurement from a single time instant. Such an approach is sensitive to sensor errors and does not account for the dynamical evolution and spatial continuity of the density. The observer-based approach we presented here includes the system dynamics, which is realized by careful modeling of the particle density behaviour using a 1D PDE with a nonlinear source term and two ODEs, which are discretized in space and time to yield a finite-dimensional nonlinear model. The influence of other plasma quantities and operational modes on the transport dynamics are included in the control-oriented model as time-varying parameters. An extended Kalman filter estimates the density, additive random-walk state disturbances as well as fringe jumps (a specific type of sensor error) from measurements, for which special measures are needed. Offline reconstruction using tokamak measurements show accurate estimation of the density profile and show the quality of fringe jump detection. Moreover, a robust state feedback controller with anti-windup is designed based on the model to track a reference signal for the average density, with the estimate obtained from the observer. Closed-loop simulations show that the controller is able to track representative reference signals, with the performance mostly limited by the nonnegativity constraint of the control input.

The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which could damage the inner wall of the machine, while operating the device with larger margins against the operational limits would lead to a reduction of the electrical output power. Second, the performance of DEMO control is limited by space restrictions for the implementation of components (optimization of the tritium breeding rate), by lifetime issues for the front-end parts (neutron and gamma radiation, erosion and deposition acting on all components) and by slow, weak and indirect action of the available actuators (plasma shaping, heating and fuelling). The European DEMO conceptual design studies include the development of a reliable control system, since the details of the achievable plasma scenario and the machine design may depend on the actual performance of the control system. In the first phase of development, an initial understanding of the prime choices of diagnostic methods applicable to DEMO, implementation and performance issues, the interrelation with the plasma scenario definition, and the planning of necessary future R&D have been obtained.

Fast protons can react with tritons in an endothermic nuclear reaction which can act as a source of neutrons in magnetically confined fusion plasmas. We have performed an experiment to systematically study this reaction in low tritium concentration (approximate to 1%) plasmas in the Joint European Torus. A linear dependence is found between excess neutron rate and tritium concentration when the DT fusion rate is low. We discuss the properties of the neutron emission, including anisotropy, from the proton-triton reaction in a fusion reactor environment and derive simple models for the calculation of the neutron yield from this reaction in terms of tritium density, fast ion temperature and fast ion energy content.

Experiments designed for the attainment of the hybrid advanced mode of operation in JET revealed the existence of a new form of reconnecting modes with toroidal number n = 1 and poloidal number m = 1, which have the same radial structure as sawtooth precursors but are characterized by small growth rate and low saturation amplitude. These modes develop in repetitive cycles with typical growth times of 100 ms. Temperature perturbations are not in the form of sawteeth; in fact, there is slow and mild erosion of the central temperature without any crash. Comparison with theory of m = 1 modes indicates that the slow growth observed in the hybrid regime is caused by low magnetic shear (a measure of the radial variation of field line inclination) and possibly by diamagnetic effects in the plasma core.

The development of scenarios characterized by H-mode confinement and internal transport barriers (ITBs) in high triangularity, δ ~ 0.4–0.5, discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours H-modes which are ELM-free or develop type I edge localized mode (ELM) activity, which inhibits long lasting ITBs. The recent experiments reported here concentrate on integrated optimization of edge and core conditions. The stability of the edge pedestal was controlled using gas injection, deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, resulting in L-mode. In parallel, the conditions for triggering and sustaining ITBs encompassing a large proportion of the plasma volume (outer ITBs) were optimized, as opposed to less performing ITBs located closer to the plasma centre (inner ITB). These plasmas have deeply reversed target current profiles with qmin ~ 3 and a narrow inner ITB, located typically at a small normalized radius ρ < 0.5 and close to the reversed shear region, is routinely observed. Large radius outer ITBs are only triggered at an input power in excess of 20 MW, but they do not usually survive the transition into H-mode. The best results, in terms of sustained high confinement, have been obtained with neon injection; an outer ITB is triggered during the phase with L-mode edge and survives into H-mode for about 2 s, corresponding to ~10 times the global energy confinement time τE, at q95 = 7.5, H89βN ~ 3.5–4 and ~60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios.

A 800 kW, 140 GHz gyrotron has been mounted on TEXTOR. Initial operation was limited by arcs in the transmission line, and by pulse length restrictions due to the expected limited power handling of the window and the launching mirror. Measurements of power absorption in the window and launching mirror, however, showed them to be capable of handling full power, 3 s pulses, while arcing could be prevented by absorption of stray radiation at strategic position in the transmission line. This resulted in the realisation of full power, 3 s pulses on the plasma. Further measures were taken to prevent gyrotron operation in wrong modes or under conditions of low absorption in the plasma. Future improvements, expected to allow further extension of the pulse length to 10 s include: a CVD diamond window to replace the current quartz torus vacuum window, and a fast and accurate launcher.

Results are presented from the JET Trace Tritium Experimental (TTE) campaign using minority tritium (T) plasmas (n(T)/n(D) 2 MA) and monotonic q-profiles. In CH discharges the gamma-ray emission decay times are much lower than classical (tau(Ts) + tau(alpha s)), indicating alpha confinement degradation, due to the orbit losses and particle orbit drift predicted by a 3-D Fokker-Planck numerical code and modelled using TRANSP.

In preparation for next step burning plasma devices such as ITER, experimental studies of instabilities and confinement of energetic ions were performed on Joint European Torus (JET) and on Mega-Amper Spherical Tokamak (MAST) with innovative diagnostic techniques, in conventional and shear-reversed plasmas, exploring a wide range of effects for energetic ions. A compendium of recent results testing capabilities of the present-day facilities for burning plasma relevant study is presented in this paper. 'Alpha tail' production using 3rd harmonic ion-cyclotron resonance heating (ICRH) of He-4 beam ions has been employed on JET for studying He-4 of the megaelectronvolt energy range in a 'neutron-free' environment. The evolution of ICRH-accelerated ions of He-4 with E >= 1.7 MeV and D with E >= 500 keV was assessed from nuclear gamma-ray emission born by the fast ions colliding with Be and C impurities. A simultaneous measurement of spatial profiles of fast He-4 and fast D ions relevant to ITER was performed for the first time in positive and strongly reversed magnetic shear discharges. Time-resolved gamma-ray diagnostics for ICRH-accelerated He-3 and H minority ions allowed changes in the fast ion distribution function to be assessed in the presence of unstable toroidal Alfven eigenmodes (TAEs) and sawteeth. A significant decrease of gamma-ray intensity from protons with E >= 5 MeV was detected during the 'tornado' modes. This was interpreted as 'tomado'-induced loss of fast ions with the drift orbit width, Delta(f), comparable to the minor radius of tokamak a. Experiments performed in the opposite case, Delta f/a = 500 keV, have shown excitation of numerous Alfven eigenmodes without a significant degradation of the fast ion confinement. The stabilizing effect of fast particles on 'monster' sawteeth was experimentally found to fail in low-density plasmas with high power ion cyclotron resonance frequency (ICRF)-heating. The transition from the 'monster' to short-period 'grassy' sawteeth was investigated with different ICRF phasing, which controls the pinch-effect and radial distribution of ICRF-accelerated ions. Instabilities excited by super-Alfvenic beam ions were investigated on the spherical tokamak MAST. Due to higher values of beta and a higher proportion of fast ions on MAST than on JET, a wider variety of modes and nonlinear regimes for the Alfven instabilities were observed, including the explosive TAE-regimes leading to the formation of hole-clump pairs on the fast ion distribution function. The MAST and START data showed that TAE and chirping modes decrease both in their mode amplitudes and in the number of unstable modes with increasing beta.

We have investigated the effect of different ion cyclotron resonance frequency (ICRF) heating schemes, of error field modes, of the plasma shape and edge magnetic shear, and of the ion delB drift direction on the stability of Alfven eigenmodes (AEs). The use of multi-frequency or 2nd harmonic minority ICRF heating at high plasma density gives rise to a lower fast ion pressure gradient in the plasma core and to a reduced mode activity in the Alfven frequency range. Externally excited low-amplitude error fields lead to a much larger AE instability threshold, which we attribute to a moderate radial redistribution of the fast ions. The edge plasma shape has a clear stabilizing effect on high-n, radially localized AEs. The damping rate of n = 1 toroidal AEs is a factor 3 higher when the ion VB drift is directed towards the divertor. These results represent a useful step towards the extrapolation of current scenarios to the inclusion of fusion-born alpha particles in ITER, with possible application for feedback control schemes for the various ITER operating regimes.

Results from recent experiments to study the effects of divertor geometry and increased plasma shaping on the L–H transition threshold on JET are reported. Equivalent septum configurations run with the new septum replacement plate (SRP) in the MkII Gas Box divertor have shown that the presence of the septum lowers the L–H transition power threshold, Pth, by 20%. For X-point to virtual septum top distances of less than 6 cm, the SRP plasmas also demonstrate a significant decrease of the L–H Pth and pedestal electron temperature, Te with reduced X-point height. Although, the SRP plasma's Pth remains above that with the septum, there is no difference in the pedestal Te at the L–H transition. The influence of plasma shaping on the L–H transition has also been investigated for the first time on JET in a series of density scans at Ip/Bt of 2.5 MA/2.7 T. While keeping the lower triangularity, δlower, and divertor geometry constant, the upper triangularity, δupper, has been increased from 0.23 to 0.34 with no effect on Pth or the pedestal Te or Ti. In a separate edge ne scan, two configurations with different δupper/δlower values of 0.23/0.23 and 0.43/0.33 have been compared. A large difference in the transition threshold is observed at values of edge ne above 1.8 × 1019 m−3, with the higher δ plasmas characterized by Pth of up to 25% lower at the highest densities scanned. This is thought to be the result of lowered X-point and outer strike point heights with increased δ.

Ion cyclotron resonance frequencies (ICRF) mode conversion has been developed for localized on-axis and off-axis bulk electron heating on the JET tokamak. The fast magnetosonic waves launched from the low-field side ICRF antennas are mode-converted to short-wavelength waves on the high-field side of the He-3 ion cyclotron resonance layer in D and He-4 plasmas and subsequently damped on the bulk electrons. The resulting electron power deposition, measured using ICRF power modulation, is narrow with a typical full-width at half-maximum of approximate to30 cm (i.e. about 30% of the minor radius) and the total deposited power to electrons comprises at least up to 80% of the applied ICRF power. The ICRF mode conversion power deposition has been kept constant using He-3 bleed throughout the ICRF phase with a typical duration of 4-6 s, i.e. 15-40 energy confinement times. Using waves propagating in the counter-current direction minimizes competing ion damping in the presence of co-injected deuterium beam ions.

ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities.

In this paper, the estimation of the thermal diffusivity from perturbative experiments in fusion plasmas is discussed. The measurements used to estimate the thermal diffusivity suffer from stochastic noise. Accurate estimation of the thermal diffusivity should take this into account. It will be shown that formulas found in the literature often result in a thermal diffusivity that has a bias (a difference between the estimated value and the actual value that remains even if more measurements are added) or have an unnecessarily large uncertainty. This will be shown by modeling a plasma using only diffusion as heat transport mechanism and measurement noise based on ASDEX Upgrade measurements. The Fourier coefficients of a temperature perturbation will exhibit noise from the circular complex normal distribution (CCND). Based on Fourier coefficients distributed according to a CCND, it is shown that the resulting probability density function of the thermal diffusivity is an inverse non-central chi-squared distribution. The thermal diffusivity that is found by sampling this distribution will always be biased, and averaging of multiple estimated diffusivities will not necessarily improve the estimation. Confidence bounds are constructed to illustrate the uncertainty in the diffusivity using several formulas that are equivalent in the noiseless case. Finally, a different method of averaging, that reduces the uncertainty significantly, is suggested. The methodology is also extended to the case where damping is included, and it is explained how to include the cylindrical geometry.

Remote maintenance will determine the available uptime of future fusion plants such as ITER. Experience at predecessor JET showed that a human-in-the-loop tele-operated approach is crucial, although this approach entails drawbacks such as the unavoidable extensive operator training and relatively long execution times. These drawbacks are common knowledge, but little quantitative research is available to guide improvements (such as improved training methods, or active operator support systems). The aim of this paper is to identify the key areas for further improvement of tele-operated maintenance. This is achieved by a detailed task analysis based on recent maintenance at JET, using task logbooks and video data as well as interviews with experienced master–slave operators. The resulting task analysis shows the (sub)tasks that were most time-consuming and shows a large variance in time performance within operators, but also substantial differences between qualified operators with different levels of experience. The operator interviews indicate that intuitive (virtual) visual feedback and artificial (guiding) forces are promising directions for improvement. The results found in this study will be used for future research and development activities focusing on haptic guiding strategies, with the aim to further design and optimize RH maintenance systems for ITER and beyond.

Remote maintenance will determine the available uptime of future fusion plants such as ITER. Experience at JET showed that a human- in- the- loop tele- operated approach is crucial, although this approach entails drawbacks such as the required extensive operator training and relatively long execution times. These drawbacks are common knowledge, but little quantitative research is available to guide improvements (such as improved training methods, or active operator support systems). The aim of this paper is to identify the key areas for further improvement of tele-operated maintenance for ITER. This is achieved by a detailed task analysis based on recent maintenance at JET, using task logbooks and video data as well as interviews with experienced master- slave operators. The resulting task analysis shows the (sub)tasks that were most time- consuming and shows a large variance in time performance within operators, but also substantial differences between qualified operators with different levels of experience. The operator interviews indicate that intuitive (virtual) visual feedback and artificial (guiding) forces are promising directions for improvement. The results found in this study will be used for future research and development activities focusing on haptic guiding strategies, with the aim to further design and optimize RH maintenance systems for ITER and beyond.

Neoclassical tearing mode (NTM) control on ITER requires detection of the mode location to be accurate and with low latency. This paper presents a systematic way to evaluate mode detection algorithms for ITER using numerical simulations of electron cyclotron emission (ECE), taking into account the radial asymmetry in the temperature perturbation by a rotating magnetic island. Simulated ECE is detected using a synthetic radiometer, in the ITER equatorial port plug, and processed by two detection algorithms for the 2/1 and 3/2 NTMs for a burning H-mode ITER plasma. One of the algorithms also incorporates simulated Mirnov data. The video bandwidth is set at 2 kHz. This allows for intermediate frequency bandwidths of B IF = 400 MHz and B IF = 300 MHz for the two algorithms, respectively. The intermediate frequency bandwidth provides a trade-off between radial accuracy (low bandwidth) and low noise/latency (large bandwidth). 2/1 and 3/2 NTMs, seeded with widths up to 9 and 11 cm, are detectable with the required accuracy within 250 ms. With appropriate settings for the radiometer, the NTM detection using ECE is accurate and with low latency. The algorithm that incorporates both ECE and Mirnov data showed the lowest detection latencies.

Maintenance of the ITER EC H&CD Upper Launcher (UL) shall be performed through the use of Remote Handling (RH) in the ITER Hot Cell Facility (HCF). The UL design will have to be fully compliant with ITER RH maintenance requirements and the set of RH tooling and services available in the HCF. This paper describes the development of an overall maintenance strategy for the UL, starting from a listing of all conceivable maintenance operations, including hands-on tasks. Components for which design concepts are discussed in this paper are the Blanket Shield Module (BSM), the steering mirror (M4), the mid optics (M1, M2) and the waveguide (WG) feed-through plate. Aspects related to RH documentation, overall maintenance strategy and design concepts for optimizing the maintainability of the UL are presented. (C) 2013 Elsevier B.V. All rights reserved.

Full suppression of neoclassical tearing modes (NTMs) using electron cyclotron current drive (ECCD) should be reached before mode locking (stop of rotation) makes suppression impossible. For an ITER scenario 2 plasma, the similar time scales for locking and island growth necessitate the combined modelling of the growth of the mode and its slow down due to wall induced drag. Using such a model, the maximum allowed latency between the seeding of the mode and the start of ECCD deposition and maximum deviation in the radial position are determined. The maximum allowed latency is determined for two limiting models for island growth; the polarization model with W-marg = 2 cm, representing the worst case, and the transport model with W-marg = 6 cm, representing the best case. NTMs with seed island widths up to 9.5 cm and 12 cm for the 2/1 and the 3/2 NTM, respectively, are suppressible. The maximum allowed latency is 1.05 s and 2.95 s for the 2/1 and 3/2 NTM, respectively, for the worst case model. Radial misalignment should not exceed 7-10 mm for the 2/1 NTM and 5-16 mm for the 3/2 NTM depending on the model for island growth. As long as the alignment suffices, it does not reduce the maximum allowed latency. Mode locking has serious implications for any real-time NTM control system on ITER that aims to suppress NTMs by ECCD.

The sawtooth instability is a repetitive phenomenon occurring in plasmas of tokamak nuclear fusion reactors. Experimental studies of these instabilities and the effect they have on the plasma (notably the drive of secondary instabilities and consequent performance reduction) for a wide variety of plasma conditions is an important line of study in nuclear fusion research. Variations in the plasma conditions have a significant influence on the dynamical behavior of the sawtooth instability. Therefore, this paper presents the design of a sawtooth period controller which is robust against such variations. The controller is from a class of adaptive controllers better known as extremum seekers. In this technique, a cost function in terms of the desired sawtooth period is optimized on-line. The Extremum Seeking Controller (ESC) is model-free and is therefore inherently robust against model uncertainty. Simulations show that the controller is robust against variations in plasma parameters, delay in the sawtooth crash detection, and noise on the in- and outputs of the process. Because of its robustness, ESC is a promising candidate strategy for a wide range of fusion-related control problems with high model uncertainty. © 2012 AACC American Automatic Control Council).

We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations with wave vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated the ability to resolve these signatures in the CTS spectrum as well as their sensitivity to the ion species mix in the plasma. This paper shows that the plasma composition can be inferred from the measurements through forward modeling of the CTS spectrum. We demonstrate that spectra measured in plasmas consisting of hydrogen, deuterium and (3)He can be accurately reproduced by theory and yield inferred plasma compositions consistent with expectations. The potential to use CTS for measurements of plasma composition is of significant interest since CTS is well suited for reactor environments and since there is at present no established method to measure the fuel ion density ratio in the core of a burning fusion plasma.

The systematic design of a robust adaptive control strategy for the sawtooth period using electron cyclotron current drive (ECCD) is presented. Recent developments in extremum seeking control (ESC) are employed to derive an optimized controller structure and offer practical tuning guidelines for its parameters. In this technique a cost function in terms of the desired sawtooth period is optimized online by changing the ECCD deposition location based on online estimations of the gradient of the cost function. The controller design does not require a detailed model of the sawtooth instability. Therefore, the proposed ESC is widely applicable to any sawtoothing plasma or plasma simulation and is inherently robust against uncertainties or plasma variations. Moreover, it can handle a broad class of disturbances. This is demonstrated by time-domain simulations, which show successful tracking of time-varying sawtooth period references throughout the whole operating space, even in the presence of variations in plasma parameters, disturbances and slow launcher mirror dynamics. Due to its simplicity and robustness the proposed ESC is a valuable sawtooth control candidate for any experimental tokamak plasma, and may even be applicable to other fusion-related control problems.

A demonstration of real-time feedback control for autonomous tracking and stabilization of m/n = 2/1 tearing modes in a tokamak using Electron Cyclotron Resonance Heating and Current Drive (ECRH/ECCD) is reported. The prototype system on TEXTOR combines in the same sight-line an Electron Cyclotron Emission (ECE) diagnostic for tearing mode sensing and a steer-able ECRH/ECCD antenna. The mode location is retrieved from the ECE measurements and serves as input for a control loop, which aligns the ECRH/ECCD deposition with the tearing modes by steering of a launcher mirror. The alignment is achieved by matching the mode location in the sensor spectrum with the fixed ECRH/ECCD actuator frequency. The control response is dominated by the response of the mechanical launcher. Analysis of the launcher dynamics receives special emphasis in the control design. In addition, the ECRH/ECCD power is modulated in phase with the rotation frequency of the O-point of the tearing modes using a feedback loop, which extracts the modes frequency and phase from the ECE data. The 'line-of-sight' concept has advantages over conventional tearing mode suppression schemes as it provides information of the modes location and phase in the direct vicinity of the EC-deposition. It avoids the use of ray-tracing in the control loop and guarantees accurate alignment in the presence of disturbances. The experimental results demonstrate the capabilities of the control system to track and suppress tearing modes in real-time. A relatively simple control design suffices to meet the performance requirements demanded for effective tearing mode suppression. In addition, a control-oriented model for the tearing mode control problem is being developed. This model includes tearing mode and plasma dynamics, models for the actuators, diagnostics, data-processing and control algorithms. Simulation results for typical TEXTOR conditions will be discussed and validated with experimental results.

Modelling of the experiments on TEXTOR on tearing mode suppression by electron cyclotron resonance heating and current drive based on the generalized Rutherford equation (GRE) is presented. The comparison between the model and the experimental data provides a satisfactory agreement taking into account the experimental uncertainties. Both the model and the experimental observations confirm that in TEXTOR heating is the dominant suppression mechanism above that of current drive. As a conclusion, these experiments provide a positive benchmark for the stabilizing term in the GRE arising from the localized heating. (Some figures in this article are in colour only in the electronic version)

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Virtual Reality (VR) simulation can be used to study, improve and verify ITER maintenance operations during preparation. VR can also improve the situational awareness of human operators during actual Remote Handling (RH) operations. Until now. VR systems use geometric models of the environment and the objects being handled and kinematic models of the manipulation systems. The addition of dynamic effects into the VR simulation was investigated. Important dynamic effects are forces due to contact transitions and the bending of beams under heavy loads. A novel dynamics simulation module was developed and introduced as an add-on to the VR4Robots VR software. Tests were performed under simplified test conditions and in the context of realistic ITER maintenance tasks on a benchmark product and on the ECRH Upper Port Launcher Plug (UPL). The introduction of dynamic effects into VR simulations was found to add realism and provide new insights in procedure development. The quality of the haptic feedback depends strongly on the haptic device used to "display" haptic feedback to the operator. Dynamic effect simulation can also form the basis for real-time guidance support to operators during the execution of maintenance tasks (augmented reality). (C) 2011 Elsevier B.V. All rights reserved.