Annual Report(PDF file)

Executive Summary

1. Mission of the Large Tokamak Implementing Agreement and relevance to the international fusion programme

The objective of this Implementing Agreement (IA) is to enhance the scientific and technological achievements of the Large Tokamaks (LT) by means of co-operative actions for the advancement of the tokamak concept. This IA is one of the largest co-operations among the fusion IA’s under the IEA. The achievements of the large tokamaks under this IA provided essential data and operating experience for ITER and the advancement of the tokamak concept.

2. Current foci and objectives of the LT IA

Current scientific foci of large tokamak experiments are: ITER baseline discharge simulation (start-up, flat-top, landing); candidate steady state scenarios for ITER and DEMO (long-duration sustainment of high plasma pressure, high bootstrap current discharges); qualification of hybrid scenarios for ITER; cross-machine experiments on plasma edge, L-H power threshold in He plasmas, Internal Transport Barriers; control of disruptions using massive gas injection; control of edge localised modes by perturbing magnetic configuration, using pellet injection and so on; characterisation of plasma instabilities (resistive wall modes, neoclassical tearing modes); effect of ITER Test Blanket Modules on plasma behavior, material erosion, migration re-deposition and fuel retention; effect of wall conditioning; effect of plasma rotation on confinement and MHD.
The objective of these investigations is to advance the scientific basis for the burning plasmas in tokamaks and contribute to the resolution of the issues identified in the ITER Research Plan and to prepare for ITER scientific exploitation. ITER will be the first burning plasma experiment to demonstrate the scientific and technical basis of fusion energy. The IEA LT scientific exchanges to carry out these investigations are accomplished through coordinated experiments and supporting data analysis and computational modeling using JET (EU), JT-60 (Japan), KSTAR (Korea) and the U.S. national devices (DIII-D, CMOD and NSTX), and many university researchers. The International Tokamak Physics Activity (ITPA), operating under the auspices of ITER, identifies high priority research tasks for ITER in close coordination with the ITER Organization, and proposes experiments and modeling activities to resolve them. The IEA LT IA holds annual workshops, in close cooperation with the IEA Poloidal Divertor (PD) and IEA Plasma Wall Interaction in TEXTOR (PWIT) IA’s, the tokamak leaders and the ITPA on "Implementation of the ITPA coordinated research recommendations". The 2009 annual workshop was held at NFRI, Daejon, Korea on 15-16 December 2009. In this 8th annual workshop, leaders representing 12 major world tokamak programmes were among the participants.
Current foci of large tokamak technology are the development of negative-ion-source-based neutral beam injector (N-NBI) for JT-60SA, tritium and remote handling in JET (including the installation and tests of the ITER-like Wall materials in JET), as well as diagnostics improvements. In general, it is considered that the interactions between IEA/ITPA/ITER work well, with the primary path for the proposal of experiments being the ITPA Topical Groups.

3. Highlights and accomplishments during the reporting period June 2009-June 2010

In the EU, JET experiments ended on 23 October 2009 (118 two shift days; strong EU involvement (294 researchers, 22 countries, 42 working days/person average); strong International collaborations (US, RF, Japan and Korea; 34 researchers); 39 ITPA ITER high priority coordinated experiments, requiring 60% run-time). Cooperation Agreement between EURATOM and Brazil signed on 27 November 2009. Contributions to ITER ICRH system design and non-active phase heating, IC wall cleaning; ITER coil set and divertor design; qualification in D₂, ⁴He and H for ITER baseline scenario (to 4.5MA/3.4T, H₉₈~0.9; independent of heating mix (NB/ICRF) despite strong rotation reduction (5-10) and dominant electron heating with ICRF; ELM frequency increased by 3-5 (RMP, kicks, pellets) but no complete suppression, ELM-associated peak heat loads reduced on outer divertor by ~30-40%, pellets penetrate to pedestal top to trigger ELMs; TF ripple <0.5% (preferably 0.2%-0.3%) for adequate τ_E in ITER) and advanced scenario (normalised conditions achieved for ITER steady-state (β_N=2.7, H₉₈=1.2-1.3, T_e=T_i, f_bs~0.4, f_GW~0.7, but not ρ^*); hybrid matched to AUG/DIII-D, extended to H₉₈~1.4 at ρ^*=0.005, 2.4T and high δ, rotational shear linked to reduced ion stiffness; C wall characterised for ILW (H retention, C source strength and migration; extrinsic seeding (Ne or N) reduces inter-ELM loading to ~1MWm^-2 for ~10% loss in τ_E; MGI reduces heat loads by ~50% during disruption thermal quench. Subsequent shutdown for installation of ITER-like W divertor and Be wall, NB power upgrade to 30MW long pulse, new/upgraded diagnostics and machine refurbishments will be followed by exploitation in 2011/12 (minimisation of T-retention, material erosion and migration, mixed material effects, melt-layer behaviour and impurity control; development of ITER scenarios fully compatible with ITER-like first wall and divertor materials and ~40MW input power). DT experiments foreseen for subsequent period up to end-2015. Feasibility studies quantified resources for ECRH (~10MW; collaboration with RF) and ELM control coils (collaboration with US). This Alternative Scenario would require significant increase in level of international contributions; international partners would be more actively involved in decision-making. Since 2000, 152 JET FT tasks launched (~23M€ total; ~2.6M€ in 2009), concentrating on tritium in tokamaks, tritium process and waste management, plasma facing components, engineering, and neutronics and safety.

After the shutdown of JT-60U in August 2008, the activities and the structure of JT-60 team were substantially shifted towards modification to the superconducting tokamak JT-60 Super Advanced (SA) while the team also continued physics studies and plasma evaluations for JT-60SA, ITER and Demo based on existing JT-60U data. Objective of JT-60SA being promoted as a joint programme of the Satellite Tokamak programme under Broader Approach (BA) agreement between Japan and EU and a domestic core programme of tokamak development in Japan, is to contribute to ITER Project and also to technical preparations for the decision of DEMO construction, with enhanced performance in plasma duration, plasma shaping control, heat exhaust and particle control, stability control, and heating & current drive capabilities. First plasma is planned for the end of FY2015. Procurements of JT-60SA components are shared by Japan and EU. By March 2010, the first two NbTi superconducting conductors of 450 m for PF coils were manufactured successfully in JAEA Naka Institute. Manufacturing of the first 20 degree sector for the vacuum vessel was on-going as scheduled. As for the European contributions to the procurements, the procurement arrangements for Quench Protection Circuit, High Temperature Superconducting (HTS) current leads and Cryostat-base were signed in FY2009. Removing components in the JT-60 assembly hall and disassembling components in the JT-60 machine hall started in November in 2009 and will be completed in 2012 as scheduled. As for the development of gyrotron aiming at output power of 1MW/100s required for JT-60SA ECRF system, the pulse length was extended up to 17 s (1 MW) with a new improved mode converter in December 2009. As for the NB system, hydrogen negative ion beams of 490keV/3A and 510 keV/1A have been successfully produced in the JT-60 negative ion source by tuning electrode gaps based on newly acquired breakdown voltage data for ITER-like wide electrodes.

The deliberations in the U.S. Fusion Energy Sciences (FES) program continue to evolve, with primary focus on developing pathways for establishing the credibility of fusion energy. The complexity of and challenges in fusion science and technology are great and requires break out from scientific and political isolation. The FES mission continues to be to expand the fundamental understanding of matter at very high temperatures and densities and to develop the scientific foundations needed to develop a fusion energy source.
In his recent presentations to a variety of different forums and stakeholder groups, Ed Synakowski has identified at least the following "three major scientific needs for establishing credibility for fusion energy:

1. We must generate, study, optimize, and learn to predict the properties of the burning plasma state
2. We must develop the scientific basis for robust control strategies for the burning plasma state
3. We must develop the understanding of the material/plasma interface, and the fusion nuclear science needed to endure the fusion environment and to harness fusion power".

Korea joined the LT IA on February 5, 2010 as a step to combine the IEA LT IA and the IEA PD IA into a single IA for streamlined management. The cool-down for the 2nd operation campaign of KSTAR was started from September, 2009 and the plasma experiments were carried out from October to November same year. During this second campaign, the machine was operated with the toroidal magnetic field of up to 3 T. Cicular plasmas with current of 300 kA and pulse length of 2 seconds have been achieved with limited capacity of PF magnet power supplies. To test the operational limit, the toroidal field coils were operated up to 36 kA corresponding to the toroidal field of 3.6 T at the axis. The second harmonic pre-ionization with 110 GHz, 250 kW gyrotron at 2 T has been studied. Various parameters such as injection angle, position and pressure have been scanned to optimize the effect. The ICRF wall conditioning (ICWC) was routinely applied during the shot to shot interval and the effect of ICWC has been quantitatively assessed. After finishing the 2nd campaign, a significant upgrade of the KSTAR device is made. To achieve D-shape, diverted plasmas, all of the plasma facing components (PFC) including divertor has been installed. The sixteen-segmented in-vessel control coils (IVCC) has also been installed prior to the installation of the PFCs. The IVCCs will be externally connected to form two sets of circular coils for vertical and radial position control, and the IVCCs will have additional capability to be utilized as the RWM/FEC and ELM control coils. The first NBI system, designed to deliver 8 MW deuterium beams into the KSTAR plasmas with three ion sources, is now under commissioning. The 3rd campaign is scheduled to begin July, 2010 from the cool-down of the super-conducting magnet systems.

The physics-related work in the collaboration is conducted under eight Task areas, seven of which cover the Topic areas used in the ITPA. These are Transport Physics and ITB Physics, Confinement Database and Modelling, MHD, Disruption and Control, Edge and Pedestal Physics, SOL and Divertor Physics, Steady State Operation, and Others (including Diagnostics, and also Power Supplies). In addition, Tritium and Remote Handling Technologies are conducted in Task Area 7. Accomplishments in these Task Areas are described in Attachment A2.
Two Workshops were held during the reporting period. These were:

• W70: Key ITER Disruption Issues at Culham Science Centre, Abingdon, UK; 7-9 October 2009;
• W71: Eighth Joint Workshop of Large Tokamak, Poloidal Divertor and Textor IAs on "Implementation of the ITPA Coordinated Research Recommendations" at KSTAR Research Center, NFRI, Daejeon, KOREA; 15-16 December 2009.

4. Future strategy

The Ex-Co meeting of the LT on 9 May 2010 provided the opportunity to discuss the future strategy of IEA IA as briefly summarized below (with further details available in the minutes of the meeting):

1. Development of a single tokamak IA: The discussions initiated some 6-7 years ago in the IEA Fusion Power Coordinating Committee (FPCC) and the Executive Committees of the IEA LT, PD, and PWIT IA’s on "re-structuring" the tokamak related IA’s, have come to conclusion. With the amendments discussed extensively at the meeting of 21-22 May 2009 and since, the IEA LT IA will become the single “Implementing Agreement for Co-operation on Tokamak Programmes (CTP)” that would be open to participation by the major tokamaks of the ITER Members. The Executive committee provided an amended text for the CTP IA in February 2010. US, Japan and Korea gave their assent to the amendments before the Executive Committee meeting in May 2010, and EU’s assent was expected to be given soon in the meeting (It was done and the CTP IA came into effect in June 2010. ).
   
2. Extension of the IEA CTP IA: The LT Executive Committee had resolved unanimously to request a five year extension of the CTP-IA from 15 January 2011 to 14 January 2016. It was agreed that the Request for Extension (RfE) would be drafted and distributed to ExCo members for comment before end-June 2010. The final documents would be sent to the Secretariat end-July 2010 for review by the FPCC by written procedure as soon as possible following that date. However, if the CERT approve the request for one-time extension for all IAs by written procedure in its next meeting, the Committee agreed that it would present the RfE to the FPCC in February 2011 and then to the CERT at the meeting June 2011 (The one-time extension until 30 June 2012 was approved by the CERT in December 2010.).
   
3. Invitation for participations of China, India and Russia: The LT Executive Committee agreed to send invitation letters to China, India and the Russian Federation under the CTP IA.

5. Collaborations inside/outside IEA

The close coupling between the ITPA, the ITER organization, the IEA FPCC, and the IAEA IFRC provide the opportunity to streamline international collaborations in fusion, with its priority for the success of ITER in achieving its key scientific and technological objectives. In recognition of the change of the world fusion programme into this new era, symbolized by the establishment of the ITER Organization, collaborations inside/outside IEA have to be strengthened in view of support and supplement ITER towards DEMO. As for the collaboration inside IEA, Korea joined the LT IA in February 2010 and the CTP IA as an amendment of the LT IA to include the PD IA has come in effect in June 2010. As for the collaboration outside IEA, invitation letters for participation in the CTP IA will be sent to China, India and the Russian Federation. The IEA LT homepage (http://www-jt60.naka.jaea.go.jp/lt/) is open to all IEA IA’s and the public.

The close coupling between the ITPA, the ITER organisation, the IEA FPCC, and the IAEA IFRC provide the opportunity to streamline international collaborations in fusion, with its priority for the success of ITER in achieving its key scientific and technological objectives. In recognition of the change of the world fusion programme into this new era, symbolised by the establishment of the ITER Organisation, collaborations inside/outside IEA have to be strengthened in view of support and supplement ITER towards DEMO. The IEA LT homepage (http://www-jt60.naka.jaea.go.jp/lt/) is open to all IEA IA's and the public.

6. Message to policy makers

The IEA Large Tokamak Implementing Agreement remains one of strongest fusion IA’s and has been effective in developing tokamak research to reach break-even conditions and in developing the necessary databases for the next step device ITER and a steady-state tokamak reactor. This Agreement provides leadership in coordinating ITPA joint experiments with other tokamak related IEA IA’s. With the Korea’s participation as a Contracting Party and revising the LT IA to the CTP IA, tokamak-related activities in FPCC are streamlined. Productive interactions with ITPA, IO and the IFRC will be further enhanced if other fusion countries such as China, the Russian Federation, and India join the CTP Agreement in the future in order to facilitate science and technology exchanges among the domestic programmes of all ITER Members.

7. List of attachments

These reports can be found on the IEA CTP IA web-site, http://www-jt60.naka.jaea.go.jp/lt/index.html, in the ‘Internal Use’ sub-area. Please contact Kensaku Kamiya (secretary) for password to access this part of the website.
A1 : Status and Plans of Three Parties
A2 : Accomplishments in Task Areas
A3 : Summary Reports on Workshops
A4 : List of Personnel Exchanges
A5 : Minutes of Executive Committee meeting in Aix-en-Provence, France.

The Status and Plans of Four Parties

<JT-60>

After the shutdown of JT-60U in 2008, the activities and the structure of JT-60 team were substantially shifted towards modification to the superconducting device, JT-60SA, while the team also continued physics studies and plasma evaluations for JT-60SA, ITER and Demo based on existing JT-60U data. Objective of JT-60SA programme being promoted as one of joint programmes under Broader Approach (BA) agreement between Japan and EU and also as a domestic core programme of tokamak development in Japan, is to contribute to ITER Project and also to technical preparations for the decision of DEMO construction, with enhanced performance in plasma duration, plasma shaping control, heat exhaust and particle control, stability control, and heating & current drive capabilities. The first plasma in the end of FY2015 is planned. Procurements of JT-60SA components are shared by Japan and EU. Existing infra-structure and equipments for JT-60U, such as heating & current drive systems, cooling facilities, power supplies, diagnostics etc., will be utilized as many as possible.
Seven procurement arrangements for Japanese contributions were launched between the Implementing Agencies, JAEA and F4E, for the supply of PF (poloidal field) magnet conductor, PF coil manufacture buildings, PF coil manufacturing, vacuum vessel, buildings for vacuum vessel sector-assembly, materials for in-vessel components, and Divertor Components by 2009. The coil manufacture building and PF conductor manufacturing building were completed in Naka Fusion Institute in March 2009, and the superconductor manufacturing machines were installed in the building in September 2009. Two conductors using the copper dummy cables were manufactured in Naka Institute as a trial, and these met the specifications and successfully passed the helium leak test. In March 2010, the first two EF-H (equilibrium field at the high field side) superconducting conductors of 450 m were manufactured successfully. Manufacturing of the first 20 degree sector for the vacuum vessel is going on as scheduled. As for the European contributions to the procurements, the procurement arrangements for Quench Protection Circuit, High Temperature Superconducting (HTS) current leads and Cryostat-base were signed in FY2009.
Removing components in the JT-60 assembly hall, such as the high-voltage bushing of the Negative Neutral Beam Injector (N-NBI), support structures for maintaining the Negative Ion Source and the high-voltage table (HVT) of N-NBI, and the shielding wall (15m width, 16m height, 0.35m thickness) between the JT-60 machine hall and the JT-60 assembly hall were started from November in 2009, and was completed in March 2010. Disassembling of components in the JT-60 machine hall started in this April and will be completed in 2012.
The development of gyrotron with a new improved mode converter was started aiming at 1MW, 100s which is the long-pulse requirement of JT-60SA ECRF system. It was confirmed that the RF diffraction loss was remarkably decreased and the cooling water temperature for the DC break, which had limited the pulse length, was saturated at about half (ΔT~30C) of that before the improvement. And the pulse length was extended up to 17 s (1 MW) in December 2009. Further conditioning is continued for extending pulse length further. Using another gyrotron, a new operation technique of Active-Anode-Voltage-Control was developed and the pulse length at 1.5 MW was extended from 1 s (achieved in 2007 as a world record) to 4 s. A new linear motion (LM) antenna was designed which has an advantage of feeding cooling water reliably to the linearly driven mirrors, compared to the conventional antenna having rotatable mirrors. The mock-up antenna showed satisfactorily wide steering range (~80) and sufficiently small beam radius (~100 mm) at EC resonance layer for JT-60SA.
Hydrogen negative ion beams of 490keV, 3A and 510 keV, 1A have been successfully produced in the JT-60 negative ion source with three acceleration stages. These successful productions, the first acceleration of the H- ions up to 500 keV at high-current of > 1 A, have been achieved by overcoming the most critical issue, i.e., a poor voltage holding of the large-sized negative ion sources with the grids of ~2 m², which were required for JT-60SA and ITER. To improve voltage holding capability, the breakdown voltages for the large grids was examined for the first time, and it was found that a required vacuum insulation distance for the large grids was 6-7 times longer than that for small-area grids (0.02 m²). Based on the result, the gap lengths between the grids were tuned in the JT-60 negative ion source. The modification of the ion source realized significant stabilization of voltage holding and also reducing the conditioning time.

<KSTAR>

After the successful first plasma generation in the middle of 2008, several upgrades had been made in the power supply, heating, wall-conditioning, and diagnostic systems of KSTAR device. The cool-down for the 2nd operation campaign was then started from September in 2009 and plasma experiments carried out from October to November. During this 2nd campaign, the machine was operated with the toroidal magnetic field of up to 3T. Circular plasmas with current of 300 kA and pulse length of 2 seconds have been achieved with limited capacity of PF magnet power supplies. To test the operational limit, the toroidal field coils were operated up to 36 kA as the designed operation current was 35 kA, corresponding to the toroidal field of 3.5 Tesla at the axis. The second harmonic pre-ionization with 110 GHz, 250 kW gyrotron at 2 T has been studied. Various parameters such as injection angle, position and pressure have been scanned to optimize the pre-ionization. The ICRF Wall Conditioning (ICWC) was routinely applied during the shot to shot interval. The effect of ICWC has been quantitatively assessed by dedicated diagnostic systems. For the study of in-vessel dust characterization, duct collectors have been installed and coupons have been installed to study the campaign-integrated deposition characteristics.

There are several engineering and physics issues studied during the 2nd campaign. One of main emphases was put on the understanding of magnetics since the jacket of the cable-in-conduit cable (CICC) of the superconducting magnets are made of incoloy 908, which is a slightly magnetic material with the permeability of about 10. A numerical model including the effect of magnetic materials has been developed and the result has been verified by measuring the effect of residual magnetic field by using electron beam and the Hall probe array system. From the accurate measurement of the stray field, it was also found that the KSTAR PF-coil systems are quite up-down symmetric with a negligible installation error. This in turn indicated that the downward shift of plasma column observed during the start-up phase is not from the static field, but from the dynamic field, probably coming from the up-down asymmetric eddy current induced in KSTAR cryostat system. Initial experiments have been carried out of ohmic start-up, MHD phenomena which include sawtooth, locked-mode, disruption etc, and plasma heating by ECH and ICRF in circular ohmic plasmas. Theoretical studies have been also carried out for the 1st and 2nd harmonic ECH pre-ionization mechanisms, and the toroidal current observed during the ECH pre-ionization with a finite vertical stray field.

After finishing the 2nd operation campaign, a significant upgrade of the KSTAR device is being made. To achieve D-shape, diverted plasmas, all of the plasma facing components (PFCs) including divertor will be installed inside the vacuum vessel. The PFCs will be covered with actively-cooled graphite tile until 2012, and then upgraded to be covered with carbon-fiber-composite (CFC) for long-pulse operation. The sixteen-segmented in-vessel control coils (IVCCs) will be installed prior to the installation of the PFCs. The IVCCs will be externally connected to form two sets of circular coils for vertical and radial position control, and then the IVCCs will be additionally connected to form twelve picture-framed coils for RWM/FEC control later. The first NBI system (NBI-1), designed to deliver 8 MW D0 neutral beams into the KSTAR plasmas with three ion sources, is now under fabrication and will be commissioned for the 1 MW beam power with the first ion source during the 3rd campaign. It is now scheduled that the cool-down for the 3rd campaign will start from July, so plasma experiments expected from September.

<U.S.>

DIII-D Highlights and Plans

The 2009/10 research program consisted of 16 weeks of physics operations in 2009 and 17 weeks in 2010. Research in support of ITER included experiments simulating the effect of Test Blanket Module (TBM) magnetic field perturbations, disruption mitigation, performance extrapolation and scenario development (startup, rampdown, and helium operation), ELM control and ELM-free operation, and hydrogenic (tritium) retention. An International Team of ITER scientists used a scaled mockup of a single ITER TBM pair inserted in a DIII-D port to assess the effects of TBM ripple on H-mode threshold, plasma performance, ELM suppression, mode locking and fast ion transport: the effect on plasma performance is relatively small with the largest effect being ~20% reduction in plasma rotation. Progress on disruption mitigation included demonstration of active runaway electron control (position and current), diagnosis of their energy distribution, and evaluation of mitigation new mitigation techniques using large shattered D₂ pellets, resonant magnetic perturbations (RMP) and large shell pellets. Shattered pellet injection (SPI) has achieved the highest local electron density for achieving collisional suppression of runaway electrons. Joint experiments in JET and DIII-D extended ρ^* transport scaling to the hybrid plasma regime, while an H-mode pedestal similarity experiment showed very little, if any, ρ^* scaling of pedestal width, consistent with predictions from the EPED1 code. A broad suite of new multi-field, multi-scale-length fluctuation diagnostics has provided detailed measurements which are being compared with gyrokinetic transport simulations and models for fast-ion stability and transport. An experimental campaign using high purity (>95%) helium plasmas revealed that the L-H power threshold is between 30-50% greater than that for deuterium plasmas. Electron cyclotron pre-ionization for low voltage startup was demonstrated in both deuterium and helium plasmas. ELM-free QH-mode plasma operation was sustained with zero net NBI torque using an n=3 non-resonant magnetic field perturbation to maintain the required edge rotational shear. Hydrogenic retention experiments showed that wall uptake occurs during startup/rampup and is very low during H-mode, a promising result for ITER. DIII-D also tested an oxygen bake as a method for removing tritium-containing carbon co-deposits. DIII-D is now beginning a one year shutdown to modify one neutral beam line to allow for variable off-axis injection; install additional n=3 magnetic perturbation coils on the centerpost; and complete a number of diagnostic and other system improvements. These upgrades will increase capabilities for current profile control experiments, energy and momentum transport studies in the core and pedestal regions, research on ELM-control physics, and exploration of 3D field effects on plasma stability and transport. Improved physics understanding and validation of predictive simulations remain a major emphasis of the DIII-D research program.

Alcator C-Mod

Alcator C-Mod has completed nearly 15 out of our planned 18 weeks of research operation in FY2010 (October 2009 - September 2010). We have significantly extended the I-mode regime to high power and plasma performance. I-mode yields strong edge ion and electron temperature barriers, excellent energy confinement (H_ITER-98 up to 1.2), and low collisionality. The I-mode regime has no need for ELMs to maintain density and impurity control. Experiments to simulate ITER-like plasma evolution during startup and rampdown have been carried out on C-Mod using the ITER shape and magnetic field, with comparable safety factor, normalized pressure and energy confinement scaling. During ramp-up, with early divert times and transitions to H-mode, significant loop-voltage savings are realized, as predicted for ITER from TSC simulations. Detailed studies of ICRF-induced flow drive on C-Mod reveal that the efficiency depends strongly on He³ concentration in the D(He³) mode conversion regime, with driven core toroidal rotation up to 110 km/s (M~0.3). Experimental and theoretical studies of intrinsic rotation show that central toroidal rotation, observed in the absence of external momentum input, scales with edge temperature gradient, and the relationship to fluctuation-induced residual stress is under investigation. For line average electron densities above 1x10²⁰m^-3 LHCD efficiency drops off more rapidly than expected theoretically, and mechanisms of anomalous absorption in the edge plasma are under investigation. A new, advanced Lower Hybrid launcher, aimed at low-loss and high power density (~100 MW/m²) is currently being installed, and first results are expected this summer. Lower Hybrid waves have been used to produce a seed population of non-thermal electrons (E>100 keV), which can be accelerated during the thermal quench (TQ) phase of disruptions up to ~20 MeV. Modeling using the 3-D NIMROD code shows that, in these conditions, when massive gas puffing is applied for disruption mitigation, the strong MHD activity which grows during the TQ causes a nearly complete stochasticization of the magnetic field, in turn causing loss of the runaway electrons during the TQ. As part of the joint research milestone to characterize power flows in the scrape-off-layer, we have installed several new advanced diagnostics, including IR imaging, probes, and fast thermocouples. Experiments in this area are being closely coordinated with the DIII-D and NSTX facilities, as well as with the US theory/modeling community.

NSTX

NSTX has completed approximately 3 weeks of operation in its FY10 run campaign after finishing the FY09 campaign in July 2009. During the FY09 campaign, there was great progress in integrating elements of plasma control and wall conditioning to produce long-pulse, wall stabilized plasmas. In these experiments, lithium coating of the plasma facing components was used to reduce the density early in the discharge and to minimize ELMs. High-poloidal beta, κ~2.6 discharges produced a high-fraction of non-inductive current (65%) that was sustained for 800 to 900 ms, corresponding to 2.5 to 3 current redistribution times. Discharges with normalized betas of up to 6 %-m-T/MA and toroidal betas up to 25% were sustained for three energy confinement times. Both high-beta scenarios could enable a CTF to achieve high neutron wall loading to fulfill its mission more effectively. Research pertaining to the Joint Research Target focused on comparing deuterium retention in Ohmic and auxiliary-heated discharges with and without lithium, indicating between 87-94% retention immediately after a shot in all these cases. The deuterium was released on time scales ranging from seconds to weeks or longer. Surface sample analysis indicated that the increasing retention with increasing lithium coverage was due to changes in surface chemistry, and that the prompt recovery even with lithium coatings applied indicated that the deuterium was only loosely bound to the lithium coated graphite. The application of 3D magnetic fields to the plasma was used to trigger ELMs in a controlled fashion in order to reduce impurity accumulation while still maintaining high core confinement and beta in ELM-free discharges. Extensive conditioning of the divertor plates prior to Co-Axial Helicity Injection experiments, along with field-nulling coils in the upper divertor to delay or suppress arcs, led to increases in CHI-generated current to nearly 200 kA at the time of handoff from CHI to induction. This translated into a flux savings of nearly 180 kA of current, which is 25% of the current flattop of 700 kA in a typical long-pulse scenario in NSTX. Recent research on Resistive Wall Modes have predicted the importance of kinetic effects in determining the plasma rotation required to stabilize the mode, and showed that the mode stability is not a simple function of rotation. Experiments indeed showed that higher fast ion content was stabilizing, consistent with the theoretical predictions. "Snowflake" divertor configurations were developed and exhibited a reduction in peak heat flux to the divertor while sustaining high confinement and beta in the plasma core. Experiments to understand the L-H transition better were performed in support of ITER high priorities. It was found that helium plasmas had threshold powers approximately 20 to 30 % greater than those for comparable deuterium plasmas, and application of the 3D fields prior to the transition caused a significant increase in threshold power. Dependences with plasma current and plasma triangularity were also found, and these could be understood within the framework of the neoclassical behavior of thermal ions, and resulting E_r wells due to their losses. A Liquid Lithium Divertor for increased particle control and production of discharges at lower collisionality, was installed between the FY09 and FY10 runs, and a focus of the FY10 campaign, which has just started, will be to assess the LLD and related plasma performance. Further, a BES diagnostic was installed during this period, which, along with the high-k scattering diagnostic, will give extended turbulence k-range coverage. This will enable more comprehensive investigations of mechanisms governing energy, particle and momentum transport.

<EFDA-JET>

The last 12 months on JET have been a period of intense operations (Campaigns C26-C27 from 12 January 2009 to 23rd October 2009; 118 S/T days in two shift operation) with very good machine performance, especially the reliability of the NB injection systems at high power. Other new hardware systems tested and exploited included a high frequency pellet injector, a disruption mitigation valve (DMV), two TAE antenna arrays and several enhanced diagnostic systems. During an intervention (8 April 2009 to 21 June 2009), systems for plasma control and Resonant Magnetic Perturbations were upgraded, respectively, for higher resilience against, and extended control capabilities for, ELMs. On 26 October 2009, the Shutdown started for the installation of enhancements of high scientific value and strategic importance (ITER-like combination of first wall materials (ILW with tungsten divertor and beryllium wall), NB Power Upgrade (30MW long pulse rather than 20MW short pulse to facilitate scenario development at high current, high β and high density); upgraded and new diagnostics; and a series of machine refurbishments). In addition, significant Fusion Technology Tasks were carried out.

The scientific programme contributed to the design of the ITER ICRH system, demonstrating ELM tolerance for all ICRH systems (ITER-like antenna ILA), External Conjugate-T, 3dB couplers), operation at high voltage on antenna strap (42kV), high power density (6.2MW/m²) and novel arc detection systems with the ILA, input for modelling (TOPICA), verifying the suitability of fundamental hydrogen heating in hydrogen plasmas and second harmonic ³He in hydrogen plasmas for the non-active phase of ITER, and assessing Ion Cyclotron Wall Cleaning under ITER-relevant conditions. Simulations of the ITER baseline scenario from start to finish (D₂ and ⁴He) contributed to design modifications to the ITER coil set and divertor. The direct effect of fast ions on the sawtooth stability was demonstrated, indicating encouraging control methods for ITER. Detrimental TF ripple effects in H-mode at low edge ν^* were quantified, showing adequate confinement in ITER requires <0.5%, preferably 0.2%-0.3%.

A wide range of experiments were carried out to qualify the ITER baseline and advanced scenarios in deuterium, helium and hydrogen and provide an improved basis for extrapolation to ITER. Stationary ELMy H-modes were developed up to 4.5MA at 3.4T, showing H₉₈~0.9 for input powers up to 2.2 times the L-H threshold power. Plasma performance was found to be independent of heating mix (NB/ICRF) and temperature profiles were similar, despite a strong reduction (5-10) of toroidal rotation and dominant electron heating with ICRH.

Qualification of the ITER Advanced Scenario showed all ITER normalised conditions relevant for steady state operation (β_N=2.7, H₉₈=1.2-1.3, T_e=T_i, f_bs~0.4, f_GW~0.7, but not ρ^*) could be achieved. Furthermore, Hybrid scenario were matched to AUG/DIII-D and extended to H₉₈~1.4 and ρ^*=0.005 at high triangularity and a toroidal field of 2.4T.Detailed analysis suggests a link between rotational shear and reduced ion stiffness.

A dedicated helium campaign, including current ramp-up and ramp-down scenarios for the start-up of ITER, concluded that helium operation will provide a robust test of the experimental space available in ITER. The H-mode power threshold is similar to that in deuterium, thus providing the possibility for H-mode studies in ITER in helium.

ELM mitigation techniques (RMP, kicks, pellets) were shown to increase the ELM frequency by a factor of 3-5, but not yet to the level required for ITER. Complete ELM suppression was not observed. Detailed IR observations show a modest reduction of 30-40% in ELM associated peak heat loads on the outer divertor. ELMs are triggeredonly when the pellet is significantly large (and fast) to penetrate to the top of the pedestal.

An important part of the programme was devoted to preparatory experiments for the ILW, including the completion of the characterisation of the carbon wall. hydrogen retention in carbon plasma facing components, carbon source strength and carbon migration towards the divertor were determined. Experiments using extrinsic impurity seeding (neon or nitrogen) show a reduction of the inter-ELM power loading to ~1MWm^-2 for only ~10% loss in energy confinement. Disruption studies using the DMV concentrated on reducing transient heat loads (50% heat load reduction during thermal quench) and techniques for the suppression run-away electron production.

The Experimental Campaigns of 2009 were characterised by a strong involvement from the European Associations (294 people from 22 countries; 50 ppy total; 42 working days per person on average) which was complemented by strong International (non-EU) collaborations (with the US, the Russian Federation, Japan and Korea; 34 scientists; 3ppy or 6 percent of staffing in 2009). 39 ITPA ITER high priority coordinated experiments were covered, requiring 60 percent of overall run-time. An Agreement for Cooperation between EURATOM and Brazil was signed on 27 November 2009; a visiting researcher from Brazil is already evaluating at JET the TAE system.

The forward plan for JET ("Reference Scenario") covers the exploitation of the ILW up to 2014, followed by DT experiments up to the end of 2015. Two Task Forces have been established for the Experimental Campaigns of 2011 and a General Planning Meeting on 1-5 March 2010 started the process of experiment elaboration which will culminate in a Second planning meeting on 15-19 November 2010. From late April/early May 2011, the experimental programme will begin to focus on the characterisation of the ILW, together with exploration of ITER operating scenarios with the ILW and physics issues essential to the efficient exploitation of the ILW and ITER. Critical issues will be the minimisation of T-retention, material erosion and migration, mixed material effects, melt-layer behavior, impurity control, and the development of ITER scenarios fully compatible with a Be/W material mix. A major challenge will be to accommodate ~40MW of heating power with the ITER-like combination of first wall and divertor materials.

For the longer-term JET programme ("Alternate Scenario"), two feasibility studies, launched in 2009, have quantified the resources needed for an ECRH system (~10MW) for heating and current profile control (collaboration with Russian Federation) and a system of Resonant Magnetic Perturbation (RMP) coils for ELM control (collaboration with US). These systems would allow the full exploitation of the ITER-like Wall and provide JET with a significantly increased capability for preparing ITER operations. This Alternative Scenario requires a significant increase in the level of contributions by international partners who would then become involved more actively in the decision-making, including the joint revision of the JET plan of exploitation.

Since 2000, 152 JET FT tasks have been launched (allocated resources ~23M€ (~2.6M€ in 2009)), concentrating on tritium in tokamaks, tritium process and waste management, plasma facing components, engineering, and neutronics and safety.

Task 1: Transport and ITB Physics

Collaborative work on ITPA-IEA joint experiments was performed for the following topics.

TC-5: Determine transport dependence on Ti/Te ratio in hybrid and steady-state scenario plasmas.
Enhanced transport was also observed during ECH in AUG as well as JT-60U and DIII-D. In AUG, R/LTi variation in NBI and NBI+ECH phases agrees with ITG threshold, provided both ExB shearing and Te/Ti stabilizing effects are taken into account. Experimentally, causality of increase in Te/Ti and enhancement of transport is not clear from responses of Te and Ti after ECH switch on. Summary of work presented at H-mode workshop.

TC-7: ITG/TEM transport dependence on Ti/Te, q profile and rotation in L-mode plasmas.
Experiments on JET shows importance of stiffness at low rotation, profiles less stiff at high rotation. R/LTi threshold shows modest dependence on Te/Ti at low rotation. Discussions to be held with DIII-D to see what experiments can be done before shutdown. Do not restrict analysis to stiffness paradigm.

TC-9: Scaling of intrinsic plasma rotation with no external momentum input.
JET experiment on L-mode plasmas at higher β_N shows strong current dependence of core rotation with plasma current observed in C-Mod. Rotation inversion experiments performed in TCV using C-Mod similar plasmas, results agree with previous TCV experiments (sense of rotation inversion opposite in limited vs diverted, rotation directions relative to C-Mod opposite). Continue to populate H-mode database, expanding to include profiles, and start L-mode database with C-Mod, JET and TCV data.

TC-13: ITG critical gradient and profile stiffness
The JET result that ion stiffness is high at low rotation shear and is reduced for increasing rotation shear (a mechanism not theoretically foreseen) calls for confirmation by performing similar experiments in other machines.

TC-14: RF Rotation Drive.
LHCD causes increase in counter rotation in C-Mod. Rotation direction changes across ECH deposition layer in JT-60U. Rotation directions opposite for MCFD (Mode Conversion Flow Drive) in C-Mod and JET (grad-B direction different). Rotation data of ECH experiments will collect from DIII-D, AUG, TCV, LHD.

TC-15: Dependence of momentum and particle pinch on collisionality.
Experiments performed on JET, DIII-D and NSTX to quantify effect on collisionality and relation between momentum and particle pinch. Pellets used on DIII-D to decouple R/Ln effect from collisionality. Preliminary analysis from NSTX shows transition of momentum pinch from inward to outward as collisionality decreases. Results also show possible outward pinch for impurities as n=3 fields applied in NSTX. Good exchange of physicists on these experiments was performed.

Task 2: Transport and Confinement

Task 2: Transport and Confinement (S. Kaye)

TC-1: Beta Dependence of Confinement. Experiments on NSTX showed a weak dependence on beta and little dependence on shape. In JET, hybrid discharges were found to have a weaker power dependence, P^-0.34, than normal H-mode scaling, P^-0.65. This actually yields a stronger beta scaling than standard H_98(y,2).

TC-2: Hysteresis and access to H-mode with H~1. In JET experiments no hysteresis was reported. In NSTX, it was found that hysteresis was calculated to be present if one did not include the dW/dt term in the power, but was absent if one did. It was concluded that ITER should not rely on hysteresis to access high density H-mode regimes.

Access to H~1 regimes was found to depend on the ELM type and power: Near P_th type III ELM’s led to H~0.8. At P>>P_th H~1 but accompanied with type I ELMs. At intermediate power levels H~1 can be obtained in ELM free regimes, but these are not stationary.

TC-3: Scaling of low density limit of P_{th, min}. JET performed experiments at 1.8 T and 3T but did not access densities low enough for a definitive conclusion. NSTX results indicated that total B may be the governing parameter.

TC-4: Species dependence of L-H power threshold. A significant amount of work was performed on a number of devices to assess this very important ITER issue.

TC-5: Transport dependence on Ti/Te in hybrid and steady state scenarios
TC-6: Transport dependence on ExB shear and momentum input

Enhanced transport has been observed at low values of T_i/T_e during ECH heating. Difficult to separate this effect from rotation changes. The experimental r/L_Ti is found to follow ITG threshold scalings. Need experiments that can decouple Ti/Te, v_phi and grad v_phi.

TC-7: ITG/TEM transport dependence on Ti/Te, q and rotation in L-mode plasmas.

JET performed experiments scanning q at high and low rotation values. Opposite behaviors of the R/LTi threshold with s/q were observed at high and low rotation.

TC-8: QH/QDB plasmas.

DIII-D was able to produce QH modes with co-NBI and extend the duration of QH plasmas at lower target densities. Mast was unable to obtain QH mode discharges with counter-injection.

TC-9: Scaling of intrinsic rotation with no external momentum input.

Additional data added to database from JT-60U, DIII-D, NSTX, JET and AUG. Trend remains the same, M ~ &beta_N. L-mode data base exhibits a different scaling, M~ I_p/n_e. Rotation inversion in L-modes was studied. Inversion observed when q~3. Inversion was found to change sign between limited an diverted discharges.

TC-10: Experimental identification of ITG/TEM/ETG turbulence and comparison with codes.

DIII-D experiments were performed in TEM dominated discharges. Also, a kappa scan was performed and n/T fluctuation cross phases were measured. Transport fluxes were underpredicted in the first two study. In the later study, cross phases consistent with the experimental measurements were found for the outer half of the plasma. On NSTX ETG predictions match the transport but not the observed k-spectra. On Tore-Supra, reasonable agreemnet between GYRO predictions and Doppler reflectometry and derived χ_eff were found in a standard case discharge.

TC-11: He profiles and transport coefficients.

New task which has begun with input from JET where He profile evolution with He gas puffing experiments were performed.

TC-12: H-mode transport at low aspect ratio.

New data from MAST are consistent with NSTX for I_p, B_T and ν^* scaling. Compared to conventional aspect ratio devices, stronger B weaker I and inverse ν scalinds are found. On NSTX χ_e was found to strongly deoend on j(r).

TC-13: ITG critical gradient and profile stiffness.

JET experiments used ripple to change and decouple Ω, grad Ω during ICRF heating. Indications of low stiffness in the behavior of R/L_Ti and q_i were found.

TC-14: RF rotation drive with ICRF, LH and ECH.

In C-Mod LHCD was found to drive a counter current rotation proportional to teh driven current. Also in C-Mod, mode conversion flow drive via IBW waves led to peaked co going v_φ and higher core T_i. On JET the same heating regime was found to drive a counter current rotation ~P_rf. On JT-60U, the rotation direction was seen to change across the ECH absorption layer.

TC-15: Dependence of momentum and particle pinch on ν^*

NSTX and D-IID studied the dependence of particle pinch on collisionality with a stronger dependence observed on NSTX although both sets of data were in agreement on the diffusivity.

Task 3: MHD, Disruptions and Control (Now ITPA TG on MHD Stability)

Resistive Wall Modes
NSTX observations of RWM onset at intermediate plasma rotation, and the measured rotation dependence of the n=1 plasma response in DIII-D, are both consistent with MISK code predictions of weaker RWM damping in the gap between precession and bounce frequencies. A DIII-D experiment (joint activity with JAEA) addressed the possible nonlinear relation between energetic particle-driven modes and RWM onset.

Non-resonant Magnetic Braking
Non-resonant magnetic braking results are becoming available from a wide range of machines, and validation of Neoclassical Toroidal Viscosity (NTV) models for rotation damping is in progress at DIII-D, JET, MAST and NSTX. In DIII-D, the NTV torque from n=3 perturbations can maintain a QH-mode even with zero NB injection torque. The n=3 torque shows a strong peak at low plasma rotation, in good qualitative agreement with NTM theory. On NSTX, significant variations were made to the ExB frequency to examine the effect on NTV braking, in plasmas with lithium wall preparation. The required criterion for 1/ν-regime torque scaling was examined with data over a wide range of the ratio ν_i/nqω_E. With sufficient lithium, rotation damping at rational surfaces does not appear to cause surfaces to lock, suggesting an NTV torque effect. In MAST NTV torque from a naturally occurring mode removes the shear in the core rotation - a result that is quantitatively consistent with NTV theory.

Neoclassical Tearing Modes
Data is now available on the aspect ratio dependence of NTM behaviour from a joint IEA/ITPA experiment on AUG, DIII-D, MAST and NSTX. On AUG-MAST (3/2) and (2/1) NTM onset data have been collected, while on DIII-D-NSTX mainly (2/1) mode data has been collected. Onset β and marginal β-values are consistent with present theories relating the marginal island width and the ion banana width. On DIII-D, pre-emptive ECCD at q=2 was used to maintain stability of ITER demonstration discharges at ITER collisionalities and low rotation. Experiments in AUG and HL-2A have demonstrated ECCD destabilisation of fast ion-stabilised sawteeth (that could otherwise trigger NTMs); the AUG experiments used steerable ECCD. Experiments and modelling show ICRH destabilisation of NB-stabilised sawteeth in JET is a fast ion effect, not due to ICCD. Cross-machine scalings (DIII-D, NSTX and JT-60U) show a dependence of the 2/1 β-limit on rotation - the theory is not yet resolved.

Disruptions
Massive Gas Injection (MGI) experiments, for disruption mitigation, have been undertaken on AUG, DIII-D, MAST, JET and TEXTOR. The experiments on MAST show a 70% reduction of the power at the outer divertor strike point, using 10% Ar in He. TEXTOR experiments show a clear reduction in energy deposited on the limite with Ne and Ar MGI, compared to He MGI and natural disruptions. A common feature is that the fuelling efficiency is found to be low for heavy gases, such as Ar. AUG achieved 24% of the critical density for runaway avalanche suppression. First results on injection of shattered cryogenic pellets were reported from DIII-D, with densities in the current quench comparable to the best MGI cases. The radiated power during MGI in C-Mod, JET and DIII-D shows strong poloidal/toroidal asymmetries at the start of gas injection, but the power soon symmetrises. Analysis of the plasma current decay time during JT-60U disruptions shows the importance of the time derivative of l_i for positive and reversed shear plasmas.

With regard to future plans from June 2010 to May 2011, it is expected that joint experiments on Disruptions, Neoclassical Tearing Modes, Resistive Wall Modes and non-resonant magnetic braking will continue, together with the related exchanges of personnel.

Task 4: Edge and Pedestal Physics (Now ITPA TG on Pedestal)

Collaborative work on ITPA-IEA joint experiments was performed for the following topics.

PEP-1 + 3: Dimensionless identity experiments in JT-60U and JET: studies of ripple effects and rotation No new experiments have been carried out, but analysis has continued. In particular, under the IEA/LT framework (JE152), H. Urano (JAEA) visited JET for the analysis and discussion of results inter-machine experiment between JET and JT-60U. The comparison of JET and JT-60 results highlights that pedestal conditions may be crucial for ripple effects, consistently with JET results showing that low collisionality plasmas are more sensitive to ripple. Tests of models for the interpretation of the particle pump-out are ongoing. The results of the analysis will be presented at the coming IAEA conference under a joint paper Comparison of pedestal characteristics in JET & JT-60U similarity experiments by H. Urano, G. Saibene, N. Oyama, V. Parail, P. de Vries, R. Sartori, Y. Kamada, K. Kamiya, A. Loarte, J. Lönnroth, Y. Sakamoto, A. Salmi, K. Shinohara, H. Takenaga, M. Yoshida, the JT-60 Team and JET EFDA Contributors.

PEP-2: Pedestal gradients in dimensionally similar discharges and their dimensionless scaling
In 2009 PEP-2 experiments were carried out in all three devices. Both JET and DIII-D have finished a full ρ^* scan at low triangularity (~0.25) and high triangularity (~0.4). AUG has carried out initial experiments to match the JET and DIII-D plasma shapes. Main results are that (a) T_e and n pedestal widths do not depend (or only very weakly) on ρ^*, misalignment of the pedestal density with respect to pedestal T observed, but only in DIII-D, (b) DIII-D experiments show a strong positive ρ^* scaling of ELM size; this is not observed in JET. The effect of input power (relative to the L-H threshold power) is being investigated. In 2010, experiments are planned in AUG (ρ^* scan) and in, DIII-D. DIII-D will study scaling of ELM size with ρ^* by investigating the effect on ELM size of proximity to the L-H threshold power, as well as further experiments (fine β scans) to clarify the dependence of pedestal width on β_p and other variables.

PEP-6: Pedestal Structure and ELM stability in DN No new results were reported.

PEP-19: Basic mechanisms of edge transport with resonant magnetic perturbations in toroidal plasma confinement devices
Experiments on DIII-D, MAST, NSTX and TEXTOR have been performed to address the basic transport physics issues associated with resonant magnetic perturbation fields. These experiments and the experimental comparison focused on three topics:
(a) Pedestal structure and transport with RMPs and relation to ELM suppression: A reduction of the pedestal electron pressure was measured in ITER-shaped H-modes in DIII-D, consistent with a field line pitch resonance. Based on vacuum magnetic field modelling it was shown (consistent with TEXTOR results that the resonant feature of this pressure reduction is driven by a near field pitch angle alignment of the magnetic field lines with the RMP coil leading to a resonant change in the edge magnetic topology and a strongly q₉₅ resonant pedestal temperature reduction while the density pump out does not show a strong q₉₅ sensitivity within the 3.2<q₉₅<4.1 range investigated. This provides further support for the applicability of this paradigm as an effective modelling approach for extrapolating RMP effect to ITER. However, the ELM suppression window does not match the widest extension of the stochastic layer and the resultant strongest pedestal pressure reduction. On the contrary, ELMs can be suppressed for rather moderate pressure reduction and considerably smaller stochastic layer extent. MAST attempts to suppress ELMs in double null H-mode plasmas with a RMP spectrum causing island overlap just below the DIII-D engineering design Chirikov criterion did not show ELM suppression and furthermore, no noticeable reaction of the plasma in terms of density pump out, strike point splitting or rotation was observed Further experiments comparing DIII-D and MAST experiments are planed with the completed internal coil set at MAST and increased NB heating capabilities.
(b) Density pump out with RMP: At MAST density pump out was found in L-mode plasmas accompanied or caused by enhanced edge turbulence and an increase in the radial electric field. Modelling of these experimental observations qualitatively reproduced the density pump out in L-mode and shows a clear dependence of the pump-out magnitude on density. At DIII-D a series of L-mode experiments in high field side limited, weakly shaped and lower single null, high triangularity plasmas with the RMP spectrum typically applied for ELM suppression have been performed. Density pump-out was measured in both shapes, but only if τ_p^* is used to quantify particle confinement. This was suggested by TEXTOR studies on density pump-out and the dependence of τ_p on the detailed RMP spectral properties and related field line trajectories. Here it was shown that at high local resonant field amplitudes on the dominant resonant surfaces a density pump-out is obtained which grows above a threshold linearly with the perturbation field. Below this threshold an increase in the density and τ_p was measured connected to an increase in the radial electric field shear and a decrease of the turbulence on this resonance. Plasma fluid and neutral transport modelling with the EMC3-Eirene code based on the vacuum paradigm was accomplished after 2 years of development and revealed in detail the reflection of the three-dimensional magnetic topology in the plasma structure and the non-turbulent particle flows. The impact of RMP driven transport changes on heat and particle exhaust have also been modelled with this code. Detailed experimental investigation of RMP effects on particle transport and changes in the global particle balance were carried out in high purity helium (He) plasmas at TEXTOR and subsequently at DIII-D. Initial analysis indicates control of the exhaust properties, i.e. the external pumping efficiency without wall pumping by the perturbed magnetic boundary. Evidence for mitigation of ELMs in helium H-mode plasmas at DIII-D was found as an important input for ITER, planning on using helium as operating gas during the non-active phase.
(c) Particle and heat flux measurements on the divertor surfaces in relation to the magnetic topology: The striation of the divertor heat and particle fluxes was investigated at all devices involved. In L-mode plasmas at TEXTOR, DIII-D and MAST, striation of heat and particle fluxes was found and the relation of the location of the striated fluxes and the width of the splitting was compared to vacuum magnetic field line tracing. At DIII-D the measured fluxes are in fair agreement with vacuum field line tracing for low collisionality H-mode plasmas and in lower single null L-mode discharges. However, high collisionality H-mode plasmas show a significant wider experimentally measured splitting in the divertor heat flux. This apparent contradiction is still under analysis. At MAST, striated heat flux patterns were observed in L-mode and the width and location matches the vacuum predicted location. At NSTX, heat flux striation was also observed. Detailed comparison between machines is planned for future to reveal the exact information which can be extracted from the separatrix lobes occurrence in terms of magnetic and plasma transport response.

PEP-21: The spatial and temporal structure of Type II ELMs
The operational space for access to small ELMs on MAST has been established. This parameter space is very similar to the region in which Type II ELMs are observed on ASDEX Upgrade so there is a strong likelihood that the two ELM types are linked. Following repairs to the EZ4 flywheel generator on AUG the Type II ELM regime has been re-established. The radial propagation of the ELMs on both devices has been measured and analysis is ongoing.

PEP-22: Controllability of pedestal and ELM characteristics by edge ECH/ECCD/LHCD
No report received; no new experiments carried out.

PEP-23: Quantification of the requirements for ELM suppression by magnetic perturbations from internal off mid-plane coils
ELM suppression has not been established on MAST, but it has been possible to increase the frequency of Type I ELMs by a factor of 5 and mitigate the ELM energy loss by carefully tuning q₉₅. Vacuum modelling shows that the q₉₅ scan performed has little effect on the Chirikov parameter profile but rather that it maximises the normalised component of the resonant field suggesting that on MAST this may be the more critical parameter. ELM suppression experiments on DIII-D in April 2010 (see also PEP-19 report) used balanced double null (DN) plasmas (similar to the MAST CDN configuration). Previous attempts to obtain ELM suppression in DIII-D DN plasmas were unsuccessful although ELM mitigation (increased frequency low amplitude ELMs) was obtained. The goal of this experiment was to maintain very well balanced discharges during the RMP pulse since previous experiments suffered from difficulties with precision control of the shape in the DN configuration. Although very good shape control was successfully maintained, no ELM suppression or density pump-out was observed. Although it is possible that some indications of an ELM suppression window were seen at high q₉₅, well outside the usual ELM suppression window in ITER. Similar Shaped plasmas. In general, when RMPs are applied to these DN plasmas there is a large density pump-out and small high frequency ELMs (possibly Type IV ELMs) appear in place of Type I ELMs. Comparisons of the DIII-D and MAST RMP experiments in DN plasmas will be performed to see if the density pump-out and ELM behaviour is similar in the two machines.

PEP-24: Minimum pellet size for ELM pacing
At AUG, penetration to at least the pedestal top was found necessary for small LFS injection to trigger ELMs. The HFS centrifuge launch system is under re-commissioning. A study on room temperature solid pellets showed C and B are good materials to mimic Be pellets in ITER. JET confirmed pellet pacing in a large tokamak. Smallest achievable pellets were found already below a size threshold for ELM triggering. This threshold was reached in the baseline scenario for 1-3 10¹⁹ D pellet particle content, correlating with pellet penetration to about pedestal top. The first ELM filament was found to develop from the ablation plasmoid. This filament can cause an additional impact zone in the divertor and the first wall. DIII-D confirms this finding, higher spatial resolution of the camera system showed filament formation in front of the LFS pellet. This is interpreted by the steeper pressure gradient causes the trigger. Small pellets were found to increase the intrinsic ELM frequency even beyond the pellet rate. Pellet ELMs show less energy losses than spontaneous ELMs during this phase. Penetration of only about 1cm was required to release the ELM.
AUG plans until the end of 2010 to bring the centrifuge system for pacing using modest pellet sizes but high repetition rates back into full operation. As well, tests for launching single room temperature pellets in order to investigate their ELM trigger potential is planned. JET has requested re-commissioning of the HFPI in order to reach design criteria (50Hz repetition rate for pacing size pellets) before March 2011. Experiments for the campaigns following Restart with the new ITER-like Wall expected from summer 2011 are under discussion. DIII-D intends to modify the existing injector by spring 2011 to obtain smaller pellets at higher rates. An additional injection line will be installed to mimic the ITER approach. MAST envisages installation of a launcher capable to inject rather fast (>250m/s) LFS pellets but limited to ~3 pellets per pulse.

PEP-25: Inter-machine comparison of ELM control by magnetic field perturbations from midplane RMP coils
ELM control experiments have been performed on JET aiming at a better understanding of the plasma response to the magnetic perturbation: (a) splitting of the outer strike point in L-mode plasma has been observed with n=2 fields on JET when field penetration occurs; (b) compensation of density pump-out has been achieved with either gas fuelling or pellet injection. However, no recovery of energy confinement has been observed; (c) complete ELM suppression was not obtained by application of n=1 or n=2 fields with a Chirikov parameter larger than 1 at Ψ_pol^1/2 > 0.925 which is one of the important criteria for the design of ITER ELM suppression coils; (d) multi-resonance effect on q₉₅ in ELM control with a low n (1 and 2) magnetic perturbation field has been observed. The mechanism of edge ergodisation, which is used to explain the results of ELM suppression with n=3 field on DIII-D, may explain the global effect of the n=1 field on ELM frequency, but it cannot explain the multi-resonance effect observed with the low n fields on JET; (e) the amplitudes of the observed torque are between the NTV torque in the 1/ν and ν regimes. The collisionality dependence of the observed RMP torque is similar to the NTV torque in the 1/ν regime; (f) a new in-vessel ELM control coil system (8+24 coils) has been designed. The NTV torque calculated from both MAST (n=3 field) and TEXTOR (m/n=6/2 field) are small, which is consistent with the experimental observations. A large plasma rotation braking has been observed on MAST with n=2 field induced by EFCCs. A quantitative measure of spectral quality (Figure of merit) has been used for comparison with different devices.
Additional joint experiments for completing the database are planned. These include (a) the investigation of the q₉₅ dependence (the multi-resonance effect) on ELM control with RMP fields on other devices (MAST, AUG, NSTX, TEXTOR); comparison of the plasma braking by the low n perturbation fields with NTV theory in the super-banana plateau regime for JET, MAST, AUG, TEXTOR; and (c) investigation of the magnetic spectrum dependence of ELM control for JET, DIII-D, MAST, NSTX, AUG, TEXTOR. Combining the experimental results with modelling and making a more detailed comparison of the experimental results on the devices will be carried out in parallel with experiments over the next years.

PEP-26: Critical edge parameters for achieving L-H transition
Experiments have been performed on several devices in 2009/2010. On ASDEX Upgrade the influence of decoupling electrons and ion has been studied in ECR heated low density discharges. On Alcator C-MOD the pedestal saturation mechanisms in EDA H-modes has been investigated during L-H transitions in discharges with high plasma and neutral density. On MAST the evolution of profiles before and during the L-H transition has been measured with high time resolution on the order of 200μs for Te, ne and Er. There is evidence that the ne profile evolves on a faster time scale than the Te profile, and fluctuations seem to be suppressed on even faster time scales. On NSTX L-H transitions at low power have been studied showing that the differences of edge parameters before the L/H transition are subtle at best. Analysis of these experiments is on going with more experiments on all devices planned or proposed.

PEP-27: Pedestal profile evolution following L-H transition
The aim of this PEP is to characterise the build-up of the density and temperature pedestals, their saturation and associated evolution of core plasma parameters with P_in/P_thr~1. In these experiments, special emphasis is required on fluctuation measurements. Currently existing data are analysed (not from dedicated experiments).
AUG: Experiments in the Oct-Dec 2010 campaign. MAST: Experiments with increased spatial resolution in 2011. JET, C-MOD, DIII-D: Proposals for 2011 will be submitted.

PEP-28: Physics of H-mode access with different X-point heights
The effect of X-point height and divertor structure has been observed on many devices. In 2009/2010 dedicated experiments were done on MAST and circumstantial evidence was collected on Alcator C-MOD, DIII-D and NSTX. On MAST an increase of the power threshold by about a factor 2 to 3 higher injected power is needed to access H-mode if the X-point height is increased by about 10cm in 0.75MA SN discharges. Data analysis remains to be completed and experimental proposals will be put forward for the experimental planning of ASDEX Upgrade, C-MOD, JET, MAST, and NSTX for their forthcoming campaigns.

Task 5: SOL and Divertor Physics

The LTA report on confinement SOL and divertor physics consists of the description of reports on ITPA led activities in this area. Below are the major ITPA elements associated with this task agreement, spanning the period June 2009 - April 2010.

Recovery of tritium from co-deposits. For H/Be co-deposits with low C fractions (below a few %) the co-deposits act like pure Be in that ~90% of the H can be desorbed at bakes of 350C. Oxygen baking at 350C results in C and H removal from co-deposits but no removal of Be. Heating of surfaces to high temperatures (1000C) may be feasible by moving the plasma wetted surface around the chamber, strike point sweeping and heating during disruptions (planned and unplanned). Oxygen-radicals produced by ECR discharges appear to be efficient at removing C from gaps (as opposed to surfaces where O bake is more efficient).

Dust behaviour. Dust injected on a number of tokamaks (e.g. MAST, DIII-D, TEXTOR) indicate a dependence of the trajectory on the mass/Z of the material. Initial modelling studies reproduce a number of the dust trajectory characteristics but apparently there are still too many unknowns. For proper comparison with modeling more effort is needed to develop injection at a known velocity/direction and to make sure a 3D trajectory can be followed by appropriate stereoscopic views. The study of the mechanisms for dust generation is in much poorer state due to lack of diagnostics. Examination is needed of how the dust collected correlates with events in the plasma.

Be sputtering yield database. The ITPA DIVSOL TG reviewed all available information on Be erosion derived from studies in tokamaks and in dedicated laboratory experiments. The results are in poor agreement which means predicting the lifetime of the Be wall in ITER and the contribution of wall erosion to tritium co-deposition has very large uncertainties. Further laboratory studies are required, together with the upcoming work within the JET ILW programme. An informal collaboration was initiated between UCSD, Sandia-Livermore and FOM to investigate Be erosion mechanisms. In addition, the wall migration experiments planned for EAST (driven by the IO) should help bring new information to benchmark code material migration calculations being performed for ITER.

Tungsten melting. Regarding use of W at the ITER strike point region, where melting is very likely, the question of melt layer dynamics and effect of eroded W on the core plasma has received more emphasis recently. Modelling of melted W surfaces observed in TEXTOR are consistent with thermal emission current moving the melted W up the tile (i.e. against gravity) as seen experimentally. The poloidal movement of the W is consistent with JxB motion, with no bridging of gaps observed. The re-solidified W exhibits poor structure (e.g. holes inside). Initial experiments on the effect of a divertor-localized W source (ASDEX-Upgrade) show minimal effects on the core plasma but there is no way at present to scale this small source to a melted tile in ITER. More experiments are required both in melt layer dynamics and effect on the core. A number of new W materials (alloys, or specially prepared W) are available on which more research is needed.

Disruptions. Even though several tokamaks have recently expanded their IR coverage of the main chamber, the results are sparse on power flows during disruptions. In addition, the variation of power and energy deposition can be very large, even within one machine; DIII-D reported a range of 3 orders of magnitude in the divertor power loads during the thermal quench for different kinds of disruptions with beta limit disruptions being the worst. It is now becoming clearer that for many major disruptions, (but not including high-β cases), a significant amount of the disruptive energy is not deposited in the divertor. This is encouraging for ITER since the thermal load specifications for the divertor can be somewhat relaxed, but worrying for the main chamber, where the ITER Be surfaces have lower damage thresholds under transient loads. An example of such observations is provided by the new IR system on JET, which indicates that first-wall energy deposition during a disruption can approach that of the divertor as the power flow profile in the SOL broadens.

Disruption mitigation. There has been a significant enhancement of diagnostic capability over the last year, resulting in studies of toroidal and poloidal asymmetries in the induced radiation following massive gas injection (MGI). There is overall agreement that MGI can significantly reduce localized, conducted heat loads to the first wall and divertor. AUG, JET, and DIII-D all report that the injected impurities are swept by poloidal drifts over the top (crown) of the plasma toward the inner wall. This raises the concern that a repeatable "hot spot" may form. ITER currently envisages the most intense radiative activity to be concentrated around the gas injection location, which these new results appear to contradict to some extent. Nevertheless, results presented from C-Mod showed clearly that the toroidal asymmetry of the thermal quench radiation flash is quite variable, suggesting that repeated radiation flash heating of a single wall location during the TQ is unlikely. More fast bolometry of MGI shots should be performed. Simultaneous MGI at two different toroidal locations should be attempted if possible - this has not yet been done and would help ITER verify that radiation flash heat loads can be distributed by going to more than one MGI port.

Limiter plasma SOL profiles. On the basis of SOL profile measurements (Tore Supra and DIII-D dedicated experiments) the important ITER assumption of significantly broadened profiles for inboard versus outboard limiter configurations is clearly verified. However, the scaling of power widths with key plasma parameters assumed by ITER is not found in these experiments. The Tore-Supra results, which included excellent edge T_i data, demonstrate how important this quantity is in determining the power flow channel.

Unmitigated ELM loads. More information is needed on ELM power flows to main chamber surfaces, including the scaling of this power flow with relative ELM size. Dedicated experiments have been performed upon ITER request in the last year to look at secondary divertor power deposition. Results are still under analysis (from DIII-D and TCV), but do show that ELM filaments can deposit energy far from the secondary strike line, in accord with the ITER assumption. New main chamber ELM energy deposition data from JET (higher temporal resolution) are consistent with the previous findings of a rough (ΔW_ELM/W)^1/2 dependence of wall energy deposition. Impressive new JET divertor IR measurements also indicate a broadening of the heat flux profile with ΔW_ELM consistent with the main chamber observations, though this is seen to occur only for much higher relative ELM sizes than ITER can tolerate. The question of how these observations scale to ITER (high density, but low collisionality) remains open.

Task 6: Steady State Operation

The following joint experiments are coordinated by the IOS-TG:
IOS-1.1: ITER baseline, at q₉₅=3, β_N=1.8, n_e≤0.85n_GW
IOS-1.2: Study seeding effects on ITER baseline discharges
IOS-2.1: ECRH breakdown assist at 20 toroidal angle
IOS-2.2: Ramp-down from q₉₅=3
IOS-3.1: Beta limit for AT with ITER recommended q-profile.
IOS-3.2: Define access conditions to get to SS scenario
IOS-4.1: Access conditions for advanced inductive (hybrid) scenario with ITER-relevant restrictions
IOS-4.2: ρ^* dependence on transport and stability in hybrid scenarios
IOS-5.1: Ability to obtain and predict off-axis NBCD
IOS-5.2: Maintaining ICRH coupling in expected ITER Regime
IOS-6: Modulation of actuators to qualify real-time profile control methods for hybrid and steady state scenarios
Significant progress was made in the period June 2009 to May 2010 in view of development of integrated operation scenarios for ITER.
For the ITER baseline scenario (IOS-1.1 and IOS-1.2), JET operation in stationary type-I ELMy H-mode at q₉₅=2.7-3 was extended to 4.5MA with H₉₈=0.9-0.95, where 1.3-2 times the L-H threshold power P_L-H was required. DIII-D long pulse operation at q₉₅=3 showed n=1 modes appearance after 2τ_E and stability boundary for the modes can be measured by neither β_N nor l_i. P_L-H in helium was ~1.3-1.5 times larger that in deuterium in DIII-D and 1-1.4 in JET. Confinement was significantly lower in helium by 40-50% than in deuterium in JET and DIII-D. Ar seeded radiative divertor was studied in combination with ELM control by RMP in DIII-D. Re-emergence of ELMs at higher gas-puff rates may account for the apparent similarities in core argon accumulation between RMP and comparable non-RMP discharges. With nitrogen seeding in JET, ITER-acceptable type III ELMs were realized.
On ECRF assisted startup (IOS-2.1), ECRH assisted breakdown in X2-mode was found to be possible with toroidal injection angle of 20 in DIII-D, AUG, KSTAR and TJ-II. The required power was similar or slightly higher as compared to perpendicular injection.
On ramp-down from q₉₅=3 (IOS-2.2), JET performed a series of ramp-down experiments in helium, since experiments in deuterium was already done in 2008. Plasmas with moderate heating power kept l_i(3) below 1.3 and in some examples ramp down with H-mode phase was obtained. ITER-proposed ramp-down scenario starting with a slow ramp-down in H-mode was performed in DIII-D. The DIII-D experiments show that the proposed ramp-down rate must be increased to avoid flux consumption in CS. DIII-D results also show the importance of an aperture reduction to the reduction in density to avoid density limits. DIII-D did a 17 MA simulation discharge for ITER with no additional problems to ramp-down.
For the ITER steady-state operation scenarios (IOS-3.1 and IOS-3.2), existing data from 2008 JT-60U experiments (H₉₈~1.7, β_N~2.7, f_BS~0.92, f_CD~0.94, n_e/n_Greenwald~0.87, q₉₅~5.3) was analyzed. The q profile is similar to the ITER scenario except higher central q~10 due to transient phase. Stability analysis using MARG2D indicates ideal wall beta limit was 2.9. On DIII-D, systematic scans on qmin and q₉₅ were performed, investigating dependence of the temperature and density profiles on the q profile. Systematic variation of temperature profile with q was found while density profile seemed to be determined primarily by the H-mode pedestal. The maximum achieved β_N decreased from 3.8 to 3.1 as q_min was increased from 1 to 1.7 at q₉₅=6.8 and decreased from 3.6 to 3.4 as q₉₅ decreased from 6.8 to 4.5 at q_min=1.5. Maximum achieved β_N was mostly 10-20% below the n=1 ideal wall stability limit. In JET, both q_min~1 and q_min~2 scenarios using I_p overshoot technique was investigated. Stability and confinement was reproducibly good in q_min~2 (H₉₈~1.25 and β_N up to 3), and very good at q_min~1 (H₉₈~1.4 and β_N up to 3). The beta limits in both high and low q_min scenarios do not appear to be due to RWM, but NTM at m/n=2/1 and 3/2, respectively.
For the ITER hybrid (or advanced inductive) operation scenarios (IOS-4.1 and IOS-4.2), access condition (new for 2009) in terms of q profile is independently investigated in various devices. JET used I_p overshoot technique to broaden the q profile, while DIII-D used a change of shape. DIII-D is going to test the I_p overshoot in 2010. On transport and stability (IOS-4.2), specific joint experiments were performed on DIII-D (May 2009) and JET (October 2009). The combined JET and DIII-D range in ρ^* was about 3 with these experiments. The on-going analysis shows that a similar plasma regime (type I ELMs, no ITB etc.) has been achieved on both DIII-D and JET despite differences in operation techniques and comparable H₉₈ was observed. The preliminary indication from the 0-D analysis is that the ρ^* dependence is closer to the Bohm scaling. Local transport analysis using the database is progressing.
On off-axis NBCD (IOS-5.1), analysis of experiment in 2008 in AUG showed anomalous fast ion transport. Analysis of 2008 JT-60U experiment showed peak in the off-axis NBCD roughly agreed with calculation. DIII-D finished analysis in 2008.
On ICRF coupling (IOS-5.2), it was found in JET that coupling perturbation was minimized during ELMs at larger antenna-plasma gap and that coupling was improved by injecting D₂ gas. In DIII-D ELMy H-mode plasma, no evidence that ionization due to antenna near fields is significant and difference between results obtained by puffing from near antenna and plasma top were subtle.
On modulation of actuators for real-time profile control (IOS-6), data analyses were completed in JET and JT-60U, and there was no new experiment. DIII-D experiment was performed using co-/ctr-/balanced-NBI, ECCD and possibly loop voltage as independent actuators, and it appears that sufficient data for system identification was obtained.

Concerning personnel exchange, C. Challis and E. Joffrin (EU) visited DIII-D for IOS-3&4 in March and May 2009, P. Politzer and T. Luce (US) remotely participated in JET for IOS-4.2 in October 2009, D. Moreau (EU) visited DIII-D for IOS-6 in November 2009 and J. Hobirk (EU) visited JT-60U for IOS-5.1 in January 2010.

Task 7: Tritium and Remote Handling Technologies

The practical development of methods for de-tritiation of first wall components has concentrated on the nature of the by-products of the process. Removal of films by laser ablation results in the generation of dust from film break-up containing part of the fuel originally in the film; this dust will have to be collected, for example with a vacuum cleaner. Developments are being co-ordinated between EFDA-JET and the EFDA Emerging Technology and System Integration Task on dust and tritium control. In June 2009 laser ablation was used in the JET Beryllium Handling Facility to remove films from mirrors exposed to JET discharges during the period 2004-2007. The laser power thresholds for removing deposited films and substrate material had been explored previously, so that power levels were set nominally to remove deposits but not damage the substrate. Up to 90% of the reflectivity was restored in the near infra-red, but only 20-30% at 400nm. Some substrate damage was observed, so more optimisation of laser parameters is required.

Surface analysis of plasma facing components from JET
Tiles removed during the 2007 Shutdown have been analysed by Ion Beam Analysis and SIMS; amounts of deposition (including ¹³C) have been assessed. A successful ¹³CH₄ puffing experiment was carried out on the last day before the 2009/10 Shutdown for the installation of the ITER-like Wall (ILW). It has already been determined that 33% of the injected ¹³CH₄ did not react inside the vacuum vessel but was pumped by the divertor cryopumps. During the Shutdown, all the plasma-facing surfaces will be removed, giving the opportunity for a more comprehensive analysis of the retained ¹³C. Tiles removed from the vessel will also be analysed for erosion, deposition and H-isotope retention.

A programme of dust collection from the JET divertor was completed at the start of the ILW Shutdown. The dust samples collected from various regions of the divertor will be weighed separately; the samples will then be available for further characterisation. This provides an opportunity for a detailed assessment of deposition in a carbon-based tokamak, for a future comparison of deposition with a metal-walled tokamak, and to provide data for the ITER Safety Case.

Marker layers for the ILW W-coated divertor tiles (a 4μm Mo layer on the W coating with a 4 μm top-coat of W) have been produced and characterised, and the marker tiles for Be substrates (using a Be coating on a nickel interlayer) are being progressed. Be tiles with a ¹⁰Be tracer have also been developed. All the markers are designed to assist in measuring erosion/deposition and transport in the ILW to support modelling for ITER.

Task 8 Other: Negative Ion Neutral Beam Technology

Dr. Grisham made four trips to collaborate with the negative ion beam groups at JAEA during the course of the past year. As in previous years, the principal thrust of this collaboration is to understand the physical mechanisms which have limited the performance of negative ion neutral beam systems, and to use this knowledge to improve future systems.

Most of the work this year was focused on finding ways to improve the voltage holding in accelerators of the sort being developed for the ITER and JT-60SA negative ion neutral beam systems, and on the design and validation of a beamlet steering concept for the JT-60SA and ITER accelerators. Improved modeling of accelerator geometries planned for these devices is now showing that, with the large acceleration gaps required for voltage holding in high voltage negative ion accelerators, each beamlet is influenced by the electric fields from many other beamlets as well as from the accelerator support structure. This means that any change to any part of the accelerator structure can easily affect the overall focusing of the total beam envelope. This requires either very careful control and optimization of the accelerator and the accelerator support structure, or alternatively, finding a way to reduce the size of accelerator gaps through improved voltage holding. A concept, magnetic insulation, was developed for doing this, as well as improving the operability of accelerators such as those needed for ITER, but it requires testing on a high voltage component test facility before it can be incorporated into accelerator designs.

IEA Large Tokamak Cooperation

Brief description of the activities in the Workshop W70

The W70 workshop on "Key ITER Disruption Issues" was jointly co-hosted by UKAEA Culham and the EFDA-JET organisation, and was held in conjunction with the 3^rd meeting of the ITPA MHD Topical Group. Consideration of disruption related issues is driving several key design issues for ITER and so it was highly timely to hold a workshop on these issues. The workshop was attended by just over 40 participants, and 2 participants from JAEA joined the meeting by teleconference.

The disruption issues identified at the workshop by the ITER IO as requiring most urgent attention include (i) radiation asymmetries during disruption mitigation; (ii) sideways forces from halo current asymmetries; (iii) improvements to the disruption database to include halo current data; (iv) development of improved numerical models for halo currents; (v) limiter heat loads during vertical displacement events (VDEs); (vi) associated with massive gas injection (MGI) mitigation the quantities and species needed to suppress runaways, and the resulting current quench rates; and (vi) other possible control measures for runaway electrons (including resonant magnetic perturbations and killer pellets). Many of the presentations at the workshop were related to these key issues and working groups under the auspices of the ITPA MHD Topical Group were established to tackle several of the most urgent issues.

The workshop was structured around sessions on disruption mitigation, halo currents, disruption statistics and databases, runaways and heat loads, and disruption modelling (the detailed agenda is attached). In the disruption mitigation session data was presented on MGI radiation asymmetries in AUG and C-MOD, also very promising first results from the shotgun pellet injector on DIII-D were presented, as were surveys of MGI mitigation results in JET and DIII-D, and the avoidance of disruptions by ECRH application in FTU and AUG. The halo current session included updates on results in NSTX and JET, modelling of halo widths in DIII-D, a detailed description of the basis for the wall touching kink mode model including comparisons with JET and projections for ITER, comparisons of halo modelling with the TSC and DINA codes, and a description of halo modelling plans (by the US, Japan and India) for ITER. The session on disruption statistics and databases included a novel analysis of root disruption causes in JET, an update on the international disruption database status, and presentations on historical beryllium disruption rates and disruption position excursions in JET. In the session on runaways and heat loads there were presentations on runaway control in Tore Supra and in DIII-D (using resonant magnetic perturbations), on modelling the runaway current plateaus in JET, and on the runaway electron wall heat loads in JET. The final session on disruption modelling included a presentation on the status of TSC modelling, JT-60SA disruption modelling, M3D halo current modelling, DINA modelling including radiation opacity, and NIMROD modelling of mitigated DIII-D disruptions. The workshop concluded with summary presentations from each session chair.

Overall the workshop discussed the latest results on key disruption issues for ITER and concluded forward plans in the most urgent areas under the auspices of the ITPA MHD Topical Group.

Agenda for IEA Workshop (W70) on Key ITER Disruption Issues

List of participants for the IEA W70 workshop (held in conjunction with the 3rd ITPA MHD TG meeting)

IEA Large Tokamak Cooperation

Brief description of the activities in the Workshop W71

The Workshop was the eighth in the series and was held jointly by the three tokamak-related IEA Implementing Agreements (IAs) and the International Tokamak Physics Activity (ITPA). The ITPA has been operating under the ITER auspices since February 27, 2008. The ITER Science and Technology (S&T) Department was well represented at the meeting with the attendance of the Deputy Director General (DDG) and the Assistant DDG at the meeting, and most of ITPA TG chairs or deputy-chairs attended and reported TG activities with the participation of two TG spoke-persons by televideo from San Diego and Naka. The scope of the workshop was to include discussions of a broad range of ITER physics R&D needs in addition to the planning of Joint Experiments. While recognizing that the ITPA is the most effective international body in place for generating coordinated experiment plans across a wide range of fusion research topics, the Workshop aimed to stimulate and facilitate increased multi-machine Joint Experiments amongst the various tokamak programmes.

The Workshop was attended by ~ 29 participants on site and 2 by televideo from San Diego and Naka, including the Chairs (or Deputy Chairs) and additional ExCom members of the three tokamak-related IEA IA’s, the Chair and additional members of the ITPA Coordinating Committee, the Chairs (or their representatives) of the seven ITPA TG’s, representatives of the ITER IO, the Programme Leaders representing 12 major world tokamaks (JET, JT-60U, DIII-D, AUG, C-MOD, TCV, Tore Supra, TEXTOR, NSTX, MAST, EAST and KSTAR). Representatives of FTU and the Russian, HL-2A tokamaks were unable to attend the Workshop. Specifically, the Workshop:

• reviewed the status of implementation of the ITPA/IEA coordinated experiments among the major world tokamaks;
• discussed new proposals made by the ITPA which would benefit from coordination of joint experiments among the major world tokamaks; and
• considered the implementation of these proposals on the major world tokamaks and through agreements such as the three tokamak-related IEA IA’s.

Extra ExCom of the LTA was met to discuss on the new membership to LTA and on the status of the revision of the new draft agreement.

IEA Cooperation Among Large Tokamak Facilities

Minutes of the 25^th Executive Committee Meeting for
the IEA Large Tokamak Cooperation Programme
9^th May 2010, Aix-en-Provence, France

The Twenty-fifth Executive Committee Meeting for the IEA Implementing Agreement on Cooperation among Large Tokamak Facilities was held at 9^th May 2010 in Aix-en-Provence, France. This meeting was held jointly with the ExCo meeting of IEA Implementing Agreement on Tokamaks with Poloidal Divertors (PD).

A. MEMBERSHIP AND CHAIR

The Committee elected Dr. M. Mori as the chairman until the next meeting. Dr. Myeun KWON, Dr. Yeong-Kook OH, Dr. Jin-Yong KIM, and Dr. Woong-Chae KIM are new two executive committee members and two alternate members, respectively, for the Republic of Korea. It is noted that the Government of the Republic of Korea announced that it had accepted the invitation from the Executive Committee of the Implementing Agreement on Co-operation on the Large Tokamak Facilities at its meeting on 22 May 2009 in Culham, United Kingdom, to join this Agreement. The Government of the Republic of Korea accepted the terms and conditions of this Implementing Agreement.
The present members of the Executive Committee are shown in Appendix A.

B. ADOPTION OF AGENDA

The Committee adopted the agenda, which is attached as Appendix B.

C. REPORTS ON THE STATUS AND PLANS OF EACH PARTY

Due to limited time, presentations for status reports from tokamak devices and task reports were skipped, while these reports were distributed to ExCo members. The status reports are attached as Appendix C.
The list of Task Coordinators are appended in Appendix D1. The activities of the Tasks (submitted reports) are attached in Appendix D2. The detailed presentations will be uploaded on the following Web-site;

D. REPORTS ON THE COMPLETED WORKSHOPS AND PERSONNEL ASSIGNMENTS FOR JUNE 2009 - MAY 2010

Workshops and personnel assignments completed in the period of June 2009 - May 2010 are listed in Appendix E1. Two workshops on "Key ITER disruption issues", in parallel to an ITPA MHD Topical Group meeting, (W70), and "8th joint WS of LT, PD and TEXTOR IA's on "Implementation of the ITPA Coordinated Research Recommendations"" (W71) were carried out. The total number of personnel assignments completed in the period was 25. All personal exchanges were for review tours (less than 4 weeks) without any participation of more than 4 weeks (see Appendix E2). Subjects are summarized as follows (see Appendix E3): Task 1 (Transport and ITB Physics) was 6 (22%); Task 2 (Confinement database and modeling) was 1 (4%); Task 3 (MHD, disruptions and control) was 2 (4%); Task 4 (Edge and pedestal physics) was 6 (22%); Task 5 (SOL and divertor physics) was 1 (4%); Task 6 (Steady State Operation) was 3 (11%); Task 7 (Tritium and RH Technologies) was 1 (4%); and Task 8 (Other) was 7 (26%). The reports on the workshops (FORM C) and the short reports for review tours are attached as Appendices E4 and E5, respectively.
We should enhance the task activities, especially for Task 2, 5 and 7.

E. PROPOSALS OF WORKSHOPS, PERSONNEL ASSIGNMENTS AND REMOTE PARTICIPATION FOR JUNE 2010 - May 2011

Proposed Workshops and Personnel Assignments for June 2010 - May 2011 are listed in Appendix F. This includes one Workshop (W72: "The First IEA Coordination of Tokamak Programs Meeting on "implementation of the ITPA Coordinated Research Recommendations"). The Committee discussed these proposals and authorized their implementation.

F. GENERAL DISCUSSION ON COOPERATION

Status and remaining issues for a new IA CTP were discussed. Following a clear presentation (See. Appendix G) of what is required from the LT Executive Committee on this matter, The IEA Secretariat (C. Pottinger) presented the status of the Agreement 4 and the process for requesting an extension. The IEA proposed that the IA request an interim extension of the CERT in November 2010. The ExCo preferred to present the "Requests for Extensions (RfE)" to the FPCC by written procedure so that the RfE can be made at the 3-4 November 2010 CERT meeting. As there was still time before November 2010 the ExCo decided to submit the RfE documents to the FPCC by written procedure and then for a final decision by the CERT at the November 2010 meeting.
Members then discussed the Criteria Table point by point and carried out a self-evaluation of each of the sections. Dr. R. Hawryluk took notes on members comments on the items for discussion for each section that will be incorporated when drafting the end-of-term and strategy plans.
It was agreed that the RfE documents would be drafted and distributed to ExCo members for comment before end-June 2010. The final documents would be sent to the Secretariat end-July 2010 for review by the FPCC by written procedure as soon as possible following that date.
Should the CERT approve the request for extension for all IAs by written procedure, the IA agreed it would present the RfE to the FPCC 15-16 February 2011 and then to the CERT at the meeting June 2011.

G. SCHEDULE FOR ANNUAL REPORT FOR FPCC

The schedule and responsible persons for the production of the annual report for FPCC were discussed. As usual, the Executive Summary will be prepared by the Chairman. He will distribute a draft in the early autumn. The deadline for submission to the FPCC will be the end of November 2010.
In addition, the schedule and responsible persons for the production of the end of term (EoT) report for the period of 2006 to 2011 were discussed. Dr. M. Mori (chair for LT IA) will distribute a draft as soon as possible.

H. NEXT MEETING

The next Executive Committee Meeting will be held in May 2011 in PPPL (U.S.).