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A 24MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20MW of this power to the plasma for Heating and Current Drive (H&CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H&CD systems. An initial attempt has been developed at prioritizing the allocated H&CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (~12 MA), and Advanced (~9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems. [ABSTRACT FROM AUTHOR]

Abstract: Experiments at JET with Be as plasma-facing material demonstrate that although Be PFCs are easily damaged, the consequences for plasma operation are mild. L-mode plasmas up to a current of 7MA and heating power ∼30MW were obtained on heavily damaged Be limiter with ∼12MJ plasma energy, albeit with a very high Be content. The behaviour of H-mode plasmas in contact with a molten Be divertor target was studied. Despite the heavy damage to the target (∼3mm molten grooves) no plasma performance deterioration was observed for medium density ELMy H-modes. Type I ELMs, in these conditions, lead to ∼40μm melt layer formation but to an average melt layer loss of only ∼4%. Based on these results, conclusions for ITER limiter and divertor operation are extracted. [Copyright &y& Elsevier]

This paper presents the experimental characterization of pedestal parameters, edge localized mode (ELM) energy, and particle losses from the main plasma and the corresponding ELM energy fluxes on plasma facing components for a series of dedicated experiments in the Joint European Torus (JET). From these experiments, it is demonstrated that the simple hypothesis relating the peeling-ballooning linear instability to ELM energy losses is not valid. Contrary to previous observations at lower triangularities, small energy losses at low collisionality have been obtained in regimes at high plasma triangularity and q95∼4.5, indicating that the edge plasma magnetohydrodynamic stability is linked with the transport mechanisms that lead to the loss of energy by conduction during type I ELMs. Measurements of the ELM energy fluxes on the divertor target show that their time scale is linked to the ion transport along the field and the formation of a high energy sheath, in agreement with kinetic modeling of ELMs. Higher density ELMs, of a convective nature, lead to overall much longer time scales for the ELM energy flux, with more than 80% of the ELM energy flux arriving after the surface divertor temperature has reached its maximum value. On the contrary, for low density ELMs, of a conductive nature, up to 40% of the energy flux arrives at the divertor target before the surface divertor temperature has reached its maximum value. These large and more conductive ELMs may lead to up to ∼50% of the ELM energy reaching the main wall plasma facing components instead of the divertor target. The extrapolation to the International Thermonuclear Experimental Reactor of the obtained results is described and the main uncertainties discussed. [ABSTRACT FROM AUTHOR]

The author offers information on the 13th International Workshop onH-mode Physics and Transport Barriers held in Lady Margaret Hall College in Oxford, England in October 2011. He mentions that the series of meetings in the workshop were based on the study of transport barriers in fusion plasmas. He also informs that several topics have been discussed in the workshop including transport barrier control and plasma performance, H-mode transition physics and plasma turbulence.

The ITER project is under construction in Cadarache, France, to assemble and operate an experimental nuclear fusion facility. The aim of this paper is the description of the implementation of the French Order of February 7, 2012, concerning Basic Nuclear Installation (also called "INB" Order) [1] within the European Union Domestic Agency (EU-DA), specifically on the Electron Cyclotron Upper Launcher (EC UL). The EC UL will be used for the control and heating of the ITER plasma. The launcher includes the port plug and related in-vessel items (nuclear shielding, blanket module, and optics) and the ex-vessel first confinement items (diamond window, isolation valve, waveguides, miterbends, tapers and port plug back end). [ABSTRACT FROM AUTHOR]

Operating ITER in the reference inductive scenario at the design values of Ip = 15 MA and QDT = 10 requires the achievement of good H-mode confinement that relies on the presence of an edge transport barrier whose pedestal pressure height is key to plasma performance. Strong gradients occur at the edge in such conditions that can drive magnetohydrodynamic instabilities resulting in edge localized modes (ELMs), which produce a rapid energy loss from the pedestal region to the plasma facing components (PFC). Without appropriate control, the heat loads on PFCs during ELMs in ITER are expected to become significant for operation in H-mode at Ip = 6--9 MA; operation at higher plasma currents would result in a very reduced life time of the PFCs. Currently, several options are being considered for the achievement of the required level of ELM control in ITER; this includes operation in plasma regimes which naturally have no or very small ELMs, decreasing the ELM energy loss by increasing their frequency by a factor of up to 30 and avoidance of ELMs by actively controlling the edge with magnetic perturbations. Small/no ELM regimes obtained by influencing the edge stability (by plasma shaping, rotational shear control, etc) have shown in present experiments a significant reduction of the ELM heat fluxes compared to type-I ELMs. However, so far they have only been observed under a limited range of pedestal conditions depending on each specific device and their extrapolation to ITER remains uncertain. ELM control by increasing their frequency relies on the controlled triggering of the edge instability leading to the ELM. This has been presently demonstrated with the injection of pellets and with plasma vertical movements; pellets having provided the results more promising for application in ITER conditions. ELM avoidance/suppression takes advantage of the fact that relatively small changes in the pedestal plasma and magnetic field parameters seem to have a large stabilizing effect on large ELMs. Application of edge magnetic field perturbation with non-axisymmetric fields is found to affect transport at the plasma edge and thus prevent the uncontrolled rise of the plasma pressure gradients and the occurrence of type-I ELMs. This paper compiles a brief overview of various ELM control approaches, summarizes their present achievements and briefly discusses the open issues regarding their application in ITER. [ABSTRACT FROM AUTHOR]

Abstract: In the framework of the ITER project, an investigation has been conducted on an alternative vessel and blanket design, aimed at reducing cost and production risk. The modifications proposed have a strong impact on plasma control since they affect the main conducting structures surrounding the plasma column, providing passive stabilization but at the same time shielding the field generated by the active coils to control the plasma motion and shape. An extensive analysis was performed to assess the plasma vertical controllability and the modified requirements to the in-vessel vertical stability coils system as well as to the external Poloidal Field coils system. A similar analysis was aimed at assessing the performance of the shape control system in presence of the modified structures. The effect on plasma breakdown was also evaluated in terms of maximum initial loop voltage, quality of magnetic null and the flux loss related to the breakdown delay that was quantified under the same hypothesis employed by ITER for the baseline design. Furthermore, the modified design presents issues for the magnetic diagnostic system, related to the shielding of the probes by the eddy currents, which were analysed with a 3D model. The results of the analyses performed have some general interest in particular regarding the influence on plasma stability of 3D structures with close proximity to the plasma. The present paper aims at giving an overview of the analyses that have been carried out and a summary of the results in terms of impact of the modified design on plasma control and scenario, and in general an evaluation of the role of passive structure in plasma vertical stability and shape control. [ABSTRACT FROM AUTHOR]

Abstract: Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I p ^4. As a consequence, the achievement of the ITER objective of Q =10 requires the machine to be routinely operated at a nominal current I p of 15MA, and at full toroidal field BT of 5.3T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration. [Copyright &y& Elsevier]

To ensure optimal plasma performance at high Qfus for the baseline scenario foreseen for ITER, the fuelling requirements, in particular for non-stationary phases, need to be assessed by means of integrated modelling to address the special additional challenges facing plasma fuelling on ITER. The fuelling scheme needs to be adjusted to ensure robust divertor heat load control, avoiding complete detachment while still maintaining low divertor temperatures and heat fluxes to minimise W sputtering and contamination of the plasma by impurities. At the same time, the core density needs to be controlled to fulfil requirements for: the application of neutral beam heating with acceptable shine-through losses; a robust transition from L-mode to stationary fusion burn; the maximisation of the fusion yield; and a fast reduction in core particle content in the termination phase. Coupled core-edge-SOL transport calculations have been performed, simulating for the first time the entire ITER plasma evolution from just after the X-point formation until the late termination phase. These calculations are being exploited to find the most effective ways of fuelling and heating DT plasmas without exceeding ITER operational limits (e.g. divertor power density). The most efficient ways to fuel ITER with gas and/or pellet injection have been investigated self-consistently with the integrated core + edge + SOL transport suite of codes, JINTRAC, developed at JET (Romanelli et al 2014 Plasma Fusion Res. 9 3403023). Our modelling is exploited to study schemes for gas and pellet fuelling for main ion SOL and core density control, respectively, and for impurity seeding by Ne for the control of SOL radiation, that allow ITER to approach Qfus 10, with plasma evolution successfully controlled to respect major operational limits through all transients from the early ramp-up until the late ramp down phase. [ABSTRACT FROM AUTHOR]

TOKES simulations of shattered Ne pellets injection into the ITER discharge of 280 MJ thermal plasma energy was carried out taking into account shielding of the pellet. Injection was simulated to be performed through 3 upper port launchers or through 1 equatorial launcher. The shielding of pellets with 2 kPa m3 Ne has minor effects on the radiated power and on wall heat load because the injected amount is much larger than required for radiation. However, the minimum Ne amount, which radiates more than 90% of the core plasma energy (the threshold) is reduced by a factor of 2 compared to the result without shielding. The thresholds are 3 × 16 Pa m3 for upper injection and 62 Pa m3 for the equatorial injector. [ABSTRACT FROM AUTHOR]

• The design of the EC window unit, already optimized by FEM analysis, was advanced by ASME III-NC and the manufacturing aspects. • The design of the EC window unit was adapted to the ITER project decision to reduce the inner diameter of the waveguides from 63.5 to 50 mm. • The design of the EC window unit is sufficiently mature to start the prototyping and testing activity in view of the ITER final design review. The torus window unit is a very particular component of the ITER EC H&CD upper launcher aiming to provide the primary vacuum and confinement boundary between the vacuum vessel and the transmission lines (TLs). The high power 170 GHz millimeter-wave beams generated by the gyrotrons travel along the TLs and pass through the window units, before being quasi-optically guided into the plasma via the upper launchers. The design of the window unit shall thus meet stringent requirements to guarantee the safety function, the millimeter-wave beam transmission and the structural integrity during normal operation and off-normal events. The unit consists of an ultra-low loss CVD diamond disk brazed to two copper cuffs; this structure is then integrated into a metallic housing by welding. The compliance with the requirements shall be assured by applying the ASME Section III – Subsection NC code and a dedicated experimental qualification program. This paper reports the way in which the design of the unit, already optimized by FEM analyses against the ITER loading conditions, was further improved by the application of the ASME III-NC code, leading to a more feasible and simpler manufacturing and assembling sequence. In addition, the impact of the ITER project decision to change the inner diameter of the waveguide from 63.5 to 50 mm, to improve the beams' mode purity, was assessed and it is also discussed. Different materials for the metallic housing and in particular for the millimeter-wave inserts of the unit were compared using appropriate engineering criteria to mitigate the significant increase of the millimeter-wave thermal loads on the waveguides when the diameter is decreased. [ABSTRACT FROM AUTHOR]

The Electron Cyclotron diamond window which is located inside the port cell serves, together with an isolation valve, as primary vacuum boundary between the ITER vacuum vessel, the transmission lines and the port cell atmosphere and it functions as confinement barrier. The window consists of an ultra-low loss Chemical Vapor Deposition (CVD) diamond disk brazed into a metallic housing and it has to guarantee the compliance with very stringent nuclear safety requirements and an adequate transmission capability for high power mm-waves (1.31 MW at 170 GHz). The design of the window unit is approaching its final phase including design validation analyses, the development of a dedicated qualification program and prototyping activities. In preparation of the testing of complete window prototypes, pressure tests were performed on a mock-up formed by a diamond disk (D = 80 mm, d = 1.11 mm) brazed to two copper cuffs. The scope of the experiments was to show both the capability of the joining between the disk and the cuffs to keep the required vacuum tightness as well as the integrity of the disk when exposed to pressure loads. Based on the requirements defined for the pressure scenarios during normal operation and off-normal events, a test program was developed accounting for cyclic tests at low pressure differentials and overpressure tests up to 2 bar pressure difference over the disk to simulate severe accidental conditions. This paper gives an outline of the ongoing development of the overall qualification program of the ITER torus window and reports specifically on the experimental set-up and the successful outcome of the pressure tests of the brazed diamond disk mock-up. The test program for the complete window prototype needs discussion and approval by F4E and ITER and will directly determine the final qualification program for the series diamond windows. [ABSTRACT FROM AUTHOR]

Abstract Disruption mitigation by shattered Ne pellet injection (SPI) has been simulated for an H-mode ITER D-T discharge of 280 MJ stored energy using the 3D TOKES code. Heating of the first wall from the radiation flash during the thermal quench of the disruption initiated by the injected pellets has been assessed. The dynamics of Ne plasma expansion in the core with radiation from this plasma have been simulated for the series of pellet sizes from 5.4 × 1024 to 8.6 × 1022 Ne atoms. The threshold pellet size for irradiation of the core energy has been found. The results of SPI and massive gas injection simulations are compared. [ABSTRACT FROM AUTHOR]

Background Pre-clinical and clinical evidence suggests a rationale for the use of anti-angiogenic agents, including sorafenib, in recurrent and/or metastatic salivary gland carcinomas (RMSGCs). This study evaluates the activity of sorafenib in patients with RMSGCs and also investigates whether the activity of sorafenib could be related to its main tailored targets (i.e. BRAF, vascular endothelial growth factor receptor 2 [VEGFR2], platelet-derived growth factor receptor α [PDGFRα] and β, RET, KIT). Patients and methods Patients received sorafenib at 400 mg BID. The primary end-point was response rate (RR) including complete response or partial response (PR); secondary end-points included RR according to Choi criteria, disease control rate (DCR), overall survival (OS), and progression-free survival (PFS). Results Thirty-seven patients (19 adenoid cystic cancers, ACC) were enrolled. Six PRs were recorded. RR was 16% (95% confidence interval [CI]: 6–32; 11% in ACC and 22% in non-ACC). Choi criteria could be applied in 30 out of 37 cases with a RR of 50% (95% CI: 31–69%); DCR was 76% (95% CI: 59–88%). Incidence of ≥G3 adverse events was 29.7%. Median PFS and OS for the entire population were 5.9 months and 23.4 months, respectively. Median PFS and OS were 8.9 and 26.4 months for ACC versus 4.2 and 12.3 months for non-ACC patients. All the cases showed expression of PDGFRβ in the stroma and VEGFR2 in endothelial cells; PDGFRα positivity was found in the stroma of four (27%) cases. All except for two cases showed no PDGFRβ, VEGFR2 and PDGFRα expression in the tumour cells. KIT expression was restricted to ACC and a weak RET expression was limited to one adenocarcinoma, not otherwise specified (NOS). No BRAF mutation was found. No correlation was observed between the sorafenib activity and the expression of its markers although all six responders (two ACC, one adenocarcinoma, NOS, one salivary duct cancer [SDC], one high-grade mucoepidermoid [HG-MEC] and one poorly-differentiated cancer) are enriched in the stromal component showing a PDGFRβ immunodecoration. In ACCs, immunohistochemistry revealed MYB protein expression in 15/16 cases (94%) and the MYB-NFIB fusion oncogene was observed in 9/14 (64%). Conclusions Sorafenib is the first anti-angiogenic agent to demonstrate activity in RMSGC patients, particularly in some histotypes such as HG-MEC, SDC and adenocarcinoma, NOS. The PDGFRβ-positive rich stromal component characterising these histotypes and the lack of correlation between the activity of sorafenib and its targets suggests anti-angiogenic effect as the prevalent mechanism of action of sorafenib in SGCs. [ABSTRACT FROM AUTHOR]

The Joint European Torus (JET) is the largest tokamak currently in operation in the world. One of the greatest challenges of JET is the integrated commissioning of all its major plant systems. This is driven, partially, by the size and complexity of its operational infrastructure and also by the fact that, being an international environment, it has to address the issues of integrating, commissioning and maintaining plant systems developed by third parties. The ITER tokamak, now in construction, is a fusion device twice the size of JET and, being a joint effort between the European Union, China, India, Japan, South Korea, the Russian Federation and the USA, it will share on a wider scale all of the JET challenges regarding integration and integrated commissioning of very large and complex plant systems. With the scope of taking advantage from the history and experience of JET, Fusion for Energy (F4E) has worked together with the Culham Centre for Fusion Energy (CCFE), the host and operator of JET, for the provision of ITER relevant user experiences related to the integrated commissioning of the tokamak. This work presents and discusses the main results and the methods that were used to extract and translate the commissioning experience information into ITER requirements. [ABSTRACT FROM AUTHOR]

The SOLPS-ITER package represents a renewed coupling of the up-to-date parallelized EIRENE Monte-Carlo code and the B2.5 edge plasma solver. Each part of the fluid-kinetic code system has been dismantled and analysed. For SOLPS-ITER, full backwards compatibility with previous SOLPS4.3 ITER reference simulations is critical. We present the results of the ongoing comprehensive benchmark activity and discuss how SOLPS-ITER can be tuned to reproduce ITER simulations performed with SOLPS4.3. In particular in SOLPS-ITER a quantitative “metric” to measure the level of code convergence has been implemented, with corresponding code diagnostics carried over from SOLPS4.3, allowing code run times to be accurately compared. Good agreement is found between the two codes in the tests so far. Remaining discrepancies in the solution in front of the divertor targets are likely related to different formulations of the equations and the different discretization schemes used by the codes. [ABSTRACT FROM AUTHOR]

In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline. [ABSTRACT FROM AUTHOR]

In view of the preparation for the operation of the ITER tokamak it is necessary to develop the plasma scenarios taking into account all engineering constraints coming from the plant and including a realistic control system. It is important to consider that, due to the high energy of ITER plasmas, much more stringent requirements are posed on the control of transients in order to avoid machine damage. Several activities are performed in the EU focusing on one side on the scenario optimization from a physics point of view and on the other side on the design and modeling of a realistic plasma control system driving the plasma configuration throughout the whole pulse and suitable for implementation on a real machine. The issues related to the computation of the control feed-forward component are addressed. In particular, the possibility to trigger a feed-forward component to solve controllability problems arising in the transitions from plasma L to H and H to L modes is studied in detail with the support of linear and non-linear simulations. A control strategy is designed and tested on non-linear simulations of the whole pulse, including linear and non-linear effects due to controller switching, plasma shape reconstruction and power supplies. The paper reports on the results of the studies performed and discuss the proposed design of the plasma control system. [ABSTRACT FROM AUTHOR]

The design of the ITER Electron Cyclotron Heating and Current Drive (EC H&CD) system has evolved in the last years both in goals and functionalities by considering an expanded range of applications. A large effort has been devoted to a better integration of the equatorial and the upper launchers, both from the point of view of the performance and of the design impact on the engineering constraints. However, from the analysis of the ECCD performance in two references H-mode scenarios at burn (the inductive H-mode and the advanced non-inductive scenario), it was clear that the EC power deposition was not optimal for steady-state applications in the plasma region around mid radius. An optimization study of the equatorial launcher is presented here aiming at removing this limitation of the EC system capabilities. Changing the steering of the equatorial launcher from toroidal to poloidal ensures EC power deposition out to the normalized toroidal radius q≈0.6, and nearly doubles the EC driven current around mid radius, without significant performance degradation in the core plasma region. In addition to the improved performance, the proposed design change is able to relax some engineering design constraints on both launchers. [ABSTRACT FROM AUTHOR]

Highlights: [•] Front steering mirror design. [•] Plasma facing blanket shield module/first wall panel design. [•] Fixed frequency torus CVD diamond window serving as first tritium barrier. [•] Prototypes and tests of the above key components in the Launcher Handling and Testing Facility. [Copyright &y& Elsevier]

Highlights: [•] The massive gas injection (neon) is simulated with the two-dimensional tokamak code TOKES assuming the toroidal symmetry. [•] The neon injection, assimilation and transport of impurities through the entire plasma volume are modelled. [•] The output of TOKES is used by the melt motion code MEMOS to assess beryllium wall temperature and the regime with melting. [•] Complete plasma cooling occurs in minimum time of 5.7ms with avoiding Be melting at any point on the first wall. [Copyright &y& Elsevier]

Abstract: In the frame of the new grant started in November 2011 between Fusion for Energy (F4E) and the ECHUL-CA consortium, the development process of the Electron Cyclotron Heating and Current Drive (EC H&CD) upper launcher (UL) in ITER has moved a step toward the final design phase. Based on the 2009 preliminary design review version, the new configuration of the UL now features a thicker single-wall mainframe (up to 90mm), a recessed first wall panel (100mm, to reduce the impact of halo currents) and a new arrangement of the internal shield blocks. The main design drivers for the structural components are still the electromagnetic (EM) loads, which need to be reassessed for the new configuration of the UL. In this paper the results of a new EM 20° sector model of ITER, specialized for the UL, are shown. Six different disruption scenarios are considered in this work: upward linear (36ms) and exponential (36ms) vertical displacement events (VDE), upward linear (36ms) and exponential (16ms) major disruptions (MD), category II upward slow and slow–fast VDEs. Comparing the analyses’ results allowed to define a set of structural loads to be used as a reference for the forthcoming structural calculations. [Copyright &y& Elsevier]

The linear ideal magnetohydrodynamic (MHD) limits of the pedestal in ITER scenarios associated with the preparation and realization of the nominal fusion gain Q = 10 (inductive scenario at 15 MA/5.3 T, half-field/halfcurrent and intermediate H-mode scenario at 10 MA/3.5 T), as well as the hybrid scenario at 12 MA/5.3 T, are investigated in this work. The accessible part of the MHD stability diagram is determined by computing the bootstrap current and self-consistently evaluating the corresponding pedestal current. This procedure shows that only a small part of peeling-ballooning diagrams is physically accessible. Uncertainties about the foreseen plasma profiles motivate studies evaluating the impact of various parameters on the pedestal limits. We have addressed issues such as the pedestal width, the global performance, pressure peaking, edge current density, internal inductance and plasma shaping. A scaling law for the maximum pedestal pressure in the ITER scenarios is proposed, highlighting that the main dependences are on the plasma current, the edge safety factor, the pedestal width and the internal inductance. [ABSTRACT FROM AUTHOR]

The build-up of plasma parameters following the H-mode transition in JET has been analysed in view of its consequences for the alpha power evolution in the access to burning plasma conditions in ITER. JET experiments show that the build-up of plasma temperature both at the plasma core and the plasma edge occurs in timescales comparable to the energy confinement time. In contrast, the evolution of the edge and core densities differs strongly depending on the level of plasma current in the discharge and of the associated NBI penetration. For higher plasma current H-mode discharges (Ip > 2.0-2.5 MA, depending on plasma shape), with naturally higher plasma densities for which NBI penetration is poorer, the core density evolves in much longer timescales than the edge density leading to the formation of rather hollow density profiles. These hollow density profiles persist for timescales of several energy confinement times until they are usually terminated by a sawtooth. Modelling of the JET experiments with JETTO shows that the density build-up following the H-mode transition can be described with a purely diffusive model, despite the low collisionalities of high current H-mode plasmas at JET. The consequences of these JET experimental/modelling findings for the access to burning plasma conditions in the ITER QDT = 10 scenario are presented. [ABSTRACT FROM AUTHOR]

We present the results from a new fuelling scan database consisting of 14 high triangularity (δ ~ 0.41), type I ELMy H-mode JET plasmas. As the fuelling level is increased from low, (ΓD ~ 0.2 x 1022 el s-1, ne,ped,/nGW = 0.7), to high dosing (ΓD ~ 2.6 x1022els-1, ne,ped/nGW = 1.0) the variation in ELM behaviour is consistent with a transition from 'pure type I' to 'mixed type I/II' ELMs (Saibene et al 2002 Plasma Phys. Control. Fusion 44 1769). However, the pulses in this new database are better diagnosed in comparison to previous studies and most notable have pedestal measurements provided by the JET high resolution Thomson scattering (HRTS) system. We continue by presenting, for the first time, the role of pedestal structure, as quantified by a least squares mtanh fit to the HRTS profiles, on the performance across the fuelling scan. A key result is that the pedestal width narrows and peak pressure gradient increases during the ELM cycle for low fuelling plasmas, whereas at high fuelling the pedestal width and peak pressure gradient saturates towards the latter half of the ELM cycle. An ideal MHD stability analysis shows that both low and high fuelling plasmas move from stable to unstable approaching the ideal ballooning limit of the finite peeling-ballooning stability boundary. Comparison to EPED predictions show on average good agreement with experimental measurements for both pedestal height and width however when presented as a function of pedestal density, experiment and model show opposing trends. The measured pre-ELM pressure pedestal height increases by ~20% whereas EPED predicts a decrease of 25% from low to high fuelling. Similarly the measured pressure pedestal width widens by ~55%, in poloidal flux space, whereas EPED predicts a decrease of 20% from low to high fuelling. We give two possible explanations for the disagreement. First, it may be that EPED under predicts the critical density, which marks the transition from kink-peeling to ballooning-limited plasmas. Second, the stronger broadening of the experimental pedestal width than predicted by EPED is an indication that other transport related processes contribute to defining the pedestal width such as enhanced inter-ELM transport as observed at high fuelling, for mixed type I/II ELMy pulses. [ABSTRACT FROM AUTHOR]

The size and capability of JET to reach high plasma current and field enables a study of the plasma behaviour at ion Larmor radius and collisionality values approaching those of ITER. In this paper such study is presented. The achievement of stationary type I ELMy H-modes at high current proved to be quite challenging. As the plasma current was increased, it became more difficult to achieve stationary conditions. Nevertheless, it was possible to achieve stable operation at high plasma current (up to 4.5 MA) and low q95 (2.65-3) at JET. One of the main reasons to revisit the high plasma current experiments done in 1997 is the higher power available and the improvement of the pedestal diagnostics. Indeed, compared with previous results, higher stored energy was achieved but confinement was still degraded. The causes of this confinement degradation are discussed in the paper. [ABSTRACT FROM AUTHOR]

Understanding the origin of rotation in ion cyclotron resonance frequency (ICRF) heated plasmas is important for predictions for burning plasmas sustained by alpha particles, being characterized by a large population of fast ions and no external momentum input. The angular velocity of the plasma column has been measured in JET H-mode plasmas with pure ICRF heating both for the standard low toroidal magnetic ripple configuration, of about ∼0.08% and, for increased ripple values up to 1.5% (Nave et al 2010 Phys. Rev. Lett. 105 105005). These new JET rotation data were compared with the multi-machine scaling of Rice et al (2007 Nucl. Fusion 47 1618) for the Alfv'en-Mach number and with the scaling for the velocity change from L-mode into H-mode. The JET data do not fit well any of these scalings that were derived for plasmas that are co-rotating with respect to the plasma current. With the standard low ripple configuration, JET plasmas with large ICRF heating power and normalized beta, βN 1.3, have a very small co-current rotation, with Alfv'en-Mach numbers significantly below those given by the rotation scaling of Rice et al (2007 Nucl. Fusion 47 1618). In some cases the plasmas are actually counter-rotating. No significant difference between the H-mode and L-mode rotation is observed. Typically the H-mode velocities near the edge are lower than those in L-modes. With ripple values larger than the standard JET value, between 1% and 1.5%, H-mode plasmas were obtained where both the edge and the core counter-rotated. [ABSTRACT FROM AUTHOR]

Electron cyclotron (EC) heating and current drive (HCD), at 170 GHz, 20MW, is one of the heating systems foreseen to assist and sustain the development of various ITER scenarios since the early phase of ITER. It is usually assumed that EC is efficient only at magnetic fields operating around full field (5.3 T) and half field (2.65 T), and most of the analyses presented so far have focused on the ECRH&CD performances at flat-top. Here, the EC capabilities are investigated for different plasma parameters corresponding to different phases of the ITER plasma discharge, from current ramp-up up to burn, and for a wide range of magnetic fields, focusing in particular on the EC potential for heating and for L- to H-mode assist. It is found that the EC system can contribute to a wide range of heating scenarios during the ramp-up of the magnetic field, significantly increasing the applicable range as a function of magnetic field as compared with traditional views. [ABSTRACT FROM AUTHOR]

Abstract: As the ITER project moves forward in the procurement phase of major components, the baseline design is subject to a series of modifications due to the finalization of the design and interaction with industry. Some modifications have a potential impact on the performance and operation of the machine. A substantial activity is being carried out in EU in support to the design, aimed at assessing such impact. The effect of the 2010 design of the central solenoid on the operational space was assessed. A detailed analysis of the plasma breakdown and start-up phases was performed to assess the impact of withdrawing the booster converters from the baseline design and of the modifications to the power supply voltage ratings. In order to reach a clear conclusion, a review of the start-up strategy was also carried out starting from an analysis of the avalanche and breakdown conditions in ITER. The controllability of the null quality and position was verified in presence of uncertainties and measurement errors. The advantages of higher voltage rating were also quantified. General aim of the paper is to present an overview of the analyses performed and of the issues identified providing an assessment of the impact of some of the proposed changes on ITER operation. [Copyright &y& Elsevier]

Abstract: The present design of the ECH (Electron Cyclotron Heating) upper port launcher has been evaluated in light of the ITER remote handling (RH) requirements. Changes to the launcher design associated with the accessibility, maintainability and manageability of replaceable components are presented. Captive bolts were placed along the flange of the Blanket Shielding Module (BSM). A hinge mechanism was integrated to simplify the (dis-)mounting of the BSM and a frame with incorporated cooling and actuation lines was suggested for simplified mounting and replacement of the steerable mirrors. Rotating the upper port plug upside-down improves maintenance access and component handling. Tools are proposed for manipulation of the port plug and its sub-components. The RH compatibility analysis can improve a design. Early consideration of RH requirements and implementation of necessary features is therefore vital. [Copyright &y& Elsevier]

Abstract: The ITER electron cyclotron (EC) system having capability of up to 26MW generated power at 170GHz is being procured by 5 domestic agencies via 10 procurement arrangements. This implies diverse types of equipment and complex interface management. It also places a challenge on control system architecture to entertain the constraints of procurement slicing and meeting the overall functional requirement. The envisioned architecture is to use the local control units (supplied with each procurement) and a supervisory plant controller (by ITER). This offers a reliable control configuration for such delicate and complex EC plant system. The control system is envisioned to monitor the whole plant and perform automated tasks that are today performed via direct human intervention. For example, the automated gyrotron conditioning and active control of the EC plant to respond to requests from the plasma control system (PCS). This later aspect requires rapid shut down of the gyrotrons and power supplies, deviation of the actuators to direct the power from an equatorial to upper launcher and then restart of the power generation for rapid stabilization of the magneto hydrodynamic (MHD) instabilities that occur in high performance plasma operation. The plant controller will be designed for optimized performance with the PCS and the feedback control system used to actively control the power (with modulation capability up to 5kHz) and launching direction for MHD stabilization. [Copyright &y& Elsevier]

Abstract: Four ITER EC H&CD (Electron Cyclotron Heating and Current Drive) Upper Launchers will be installed in the ITER Tokamak to counteract plasma instabilities by injection of up to 20 MW of millimeter-wave power at 170GHz. Each Launcher features a structural system which is equipped with eight beam lines in a Front-Steering arrangement. The Launcher development has reached the status of a preliminary design, since the corresponding review meeting was held in November 2009 at the ITER site in Cadarache. All design work is performed by several EU associations being contracted by Fusion for Energy (F4E). The structural design of the Upper Launcher consists of three sub-components: First of all the Blanket Shield Module (BSM), which fills the gap between the regular blankets. The BSM dissipates about 80% of the nuclear heating and envelopes the front mirrors of the mm-wave system. Further the Launcher Mainframe, which provides a rigid structure for precise and secure integration of the mm-wave system to guarantee reliable operation under all potential scenarios. Finally the internals, such as dedicated support structures for the mm-wave system, shielding elements and components for gas and coolant supply. The most challenging design aspects are proper dissipation of nuclear heating in zones of high heat flux, the mechanical integrity during plasma disruptions, the integration of sufficient shielding material and the precise alignment of the mm-wave system under tight space conditions. Furthermore the definition of efficient manufacturing routes with respect to tolerance compliance requires substantial investigation and, though the Launcher is designed for ITER lifetime, potential repair by adequate remote handling procedures must be considered. This paper presents the recent status of the preliminary structural design and outlines future design approaches with the main focus on manufacturing methods, remote handling capability of the sub-components and optimum integration of the internals to bring the EC Launcher towards the final design. [Copyright &y& Elsevier]

Abstract: The RF heating and current drive (H&CD) systems that are to be installed in ITER during the construction phase, are the electron cyclotron (EC) and ion cyclotron (IC) systems. They are complex assemblies of high voltage power supplies (HVPS), RF generators, transmission lines and antennas. Their design and integration are constrained by many interfaces, both internal, between the subsystems, and external, with the other ITER systems. In addition, some components must be compatible with a nuclear environment and are classified as Safety Important Component. This paper describes the processes implemented in ITER to ensure proper integration. [ABSTRACT FROM AUTHOR]

Abstract: The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits. [Copyright &y& Elsevier]

Abstract: The Electron Cyclotron system for ITER is an in-kind procurement shared between five parties and the total installed power will be 24MW, corresponding to a nominal injected power of 20MW to the plasma, with a possible upgrade up to 48MW (corresponding to 40MW injected). Some critical issues have been raised and changes are proposed to simplify these procurements and to facilitate the integration into ITER. The progress in the design and the integration of the EC system into the whole project is presented in this paper, as well as some issues still under studies and some recommendations made by external expert committees. [Copyright &y& Elsevier]

Abstract: As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical stabilization system (VS) was performed. Recent results from experiments, aimed at the evaluation of the feasibility of the ITER reference scenarios, have raised several concerns regarding mostly the ramp-up and ramp-down phases of the pulse. The main issue is the value of the internal inductance li which may reach values outside the range 0.7–1, considered as reference for the ITER control system design. Similar concerns apply to the low current L-mode plasmas, needed to the exploitation of the machine towards the development of the 15MA pulse. The performance of the reference vertical stabilization system, under the revised conditions may be marginal, in particular if the effect of plasma generated noise on the velocity measurement is considered. A reliable and robust VS is mandatory to guarantee the operation of ITER at the reference elongation and plasma current values. To avoid de-scoping of the machine mission, several solutions have been proposed to improve the VS performances, ranging from an upgrade of the maximum voltage available to the present external coils system, to the introduction of in-vessel passive and/or active conductors. The paper presents an overview of the modelling and experimental effort aimed at the assessment of the baseline ITER VS and analyses the proposed solutions to improve the system performance. [Copyright &y& Elsevier]

Abstract: In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δB T =0.08% up to δB T =1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β N values above the ‘no-wall limit’ (β N ∼3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg. [Copyright &y& Elsevier]

Abstract: By using a fully three dimensional magnetic field orbit-following Monte-Carlo code, the energetic ion confinement was investigated for the current conceptual design of the ferromagnetic components in ITER which will be employed for reducing the toroidal magnetic field (TF) ripple. The ferromagnetic insert is effective in the reference standard scenario with Q =10 (Scenario No. 2) and steady state scenario with Q =5 (Scenario No. 4) to improve the energetic ion confinement. Over-compensation appears at half of the full toroidal magnetic field and its effect becomes stronger when the quantity of the ferromagnetic insert is increased in order to more reduce the TF ripple at the full toroidal magnetic field. Though the current design is acceptable, whether to increase the ferromagnetic insert to achieve lower TF ripple amplitude at the full field operation depends on how prospected are possibilities of lower field operations. Planned test blanket modules do not induce large loss (<1%) at the full field in Scenario No. 4. At the half field, however, the loss reaches ∼10% for the alpha particles due to localized large TF ripple. [Copyright &y& Elsevier]

Abstract: Engineering analysis has been carried out on a representative equatorial diagnostic port plug of ITER, outcome of which is given here. A preliminary overview of the work for a prototypical diagnostic upper port plug is also reported. To ensure that the port plug structure satisfies the requirements of the ITER environment, the following analyses have been performed: finite element (FE) analyses of static and dynamic behaviour under electromagnetic loads, FE vibration analysis, FE thermal-structural analysis and nuclear heating and radiation dose studies. The main outcomes from these analyses and consequent design developments are a revision of the stainless steel/water ratio in the plug neutron shielding modules and the incorporation of improved neutron shielding of the port duct, a modification of the port plug top plate arrangement, to increase torsional stiffness of the structure under disruption loads, and improved solutions for cooling arrangements. Manufacturing studies have also been performed, involving the assessment of methods for the production of plug flanges, the analysis of welding techniques for parts assembly and methods for the introduction of cooling features in plug components. Reference design solutions and possible modifications will be presented and discussed. [Copyright &y& Elsevier]

Abstract: The structural components of the ITER ECH upper port plug are presented which integrates the mm-wave beam lines of the front steering launcher for the control of plasma instabilities. The design of the main structure considers two alternatives. Firstly, a double wall configuration for the port plug frame responds to homogeneous baking requirements at 240°C. Secondly, an alternative frame is presented with a single wall in sections that do not require active cooling. The blanket shield module (BSM) is designed as a double-wall housing with a first wall panel (FWP) and individual shield blocks respecting the open space for the mm-wave beams. The cut-out sections to the design of the FWP and the neighbouring lower regular blanket module are defined for the beams of the actual front steering model. The thermo-mechanical stresses of the FWP and the BSM housing caused by nuclear heating and plasma radiation were analysed for the more severe case of the earlier remote steering (“3/8 RS”) reference model. As a result of the calculations for stationary loads and for transient loads during a typical plasma burn of 400s, no critical stresses have to be expected in the BSM housing. [Copyright &y& Elsevier]

Abstract: In order to exploit the capability of electron cyclotron waves (ECW) for localized heating and current drive over a large range of plasma radii in ITER, the electron cyclotron heating (ECH) and current drive (CD) upper port launcher needs a beam steering capability. In this paper the remote steering (RS) option is described as it is being studied by a number of European institutes. For the remote-steering approach a number of critical design issues were identified by the EFDA and ITER-IT teams. All technical design issues could be solved but the performance of the system, as far as stabilization of neoclasical tearing modes (NTMs) is concerned was not sufficient for all the required ITER plasma scenarios. In order to overcome this, new design efforts are ongoing in order to improve the performance of the RS setup at the relevant NTM positions. An increase of the square wave-guide length to around 9m is foreseen to enhance the current-drive capability to the minimum levels that are needed to stabilize the 3/2 and 2/1 NTM modes in nearly all ITER scenarios. [Copyright &y& Elsevier]

Abstract: The EU will be providing the largest contribution to the ITER electron cyclotron (EC) heating and current drive (H&CD) system (20MW, CW at 170GHz). The contribution includes one third of the H&CD gyrotrons, their associated power supplies and four upper port launcher antennas. In all areas of participation, the EU EC partnership (coordinated by the European Fusion Development Agreement) aims toward advancing the technology, while staying within a specified cost envelope. This is portrayed in the co-axial gyrotron development that offers the potential to double the output power per source (2.0MW), increasing the delivered power for a fixed number of auxiliary systems. The EU partnerships also attempt to increase performance for the entire EC system, in particular the launching antennas. The proposed front steering launcher design offers greater control of MHD activity than the previous remote steering design and opens up the possibility of an enhanced performance UL. The EC physics requirements are repartitioned between the upper and equatorial launchers for a synergetic balance, which increases the EC physics capabilities while relaxing some of the engineering requirements. [Copyright &y& Elsevier]