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.
Current scientific foci of large tokamak experiments are energy confinement (dependence on plasma pressure, collisionality and aspect ratio); dependence of density peaking on collisionality; control of plasma instabilities (resistive wall modes, neoclassical tearing modes at high beta, edge localized modes, disruptions); identity and similarity studies of the edge plasma; material erosion, migration re-deposition and fuel retention; long-duration sustainment of steady-state high plasma pressure plasma discharges with reduced TF ripple and high bootstrap currents; hybrid and other advanced modes and ITER scenario development; effect of q profile on triggering high confinement and fast particle induced MHD instabilities; and real time control of plasma profiles,.
The objectives of investigation of the issues listed above are to advance the scientific basis for the burning plasmas in tokamaks and contribute to the resolution of ITER Science and Technology Advisory Committee (STAC) issues identified in the ITER design review in 2007 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) and JT-60 (Japan) and the U.S. national devices ( DIII-D, CMOD and NSTX), and many university researchers. The International Tokamak Physics Activity (ITPA), now operating under the ITER auspices, 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 and IEA Plasma Wall Interaction in TEXTOR IAs, the tokamak leaders and the ITPA on "Implementation of the ITPA coordinated research recommendations". The 2007 annual workshop was held at JET in Culham, U.K. on Nov.-29-30, 2007. In this 6th annual workshop, leaders representing 11 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) in JT-60U, tritium and remote handling in JET (including cleaning of plasma facing components using a flash lamp and a small plasma torch), as well as diagnostics improvements. In general, it was considered that the interactions between IEA/ITPA/ITER were working well, with the primary path for the proposal of experiments being the ITPA Topic Groups. The need for improved coordination in joint modeling activities was also recognized.
In the EU, JET has seen the successful completion of the 2007 Shutdown and 2008 Restart of Operations, the elaboration and start of the 2008 Experimental Campaigns, major R&D activities and procurements related to components and buildings for enhancements to be installed on JET during 2009/10, the conduct of Fusion Technology Tasks using the JET Facilities, and a proposal for a longer-term JET programme to end 2014. The 2008 scientific programme has a strong ITER focus (High level commissioning of new systems and issues that could impact on the design of ITER components; preparation of integrated operating scenarios for ITER; and physics issues essential to the efficient exploitation of ITER) and extends over 102 days (including 30% contingency), together with 73 days for commissioning the ITER-like ICRH Antenna.
In Japan, objective of JT-60 experiments is to establish advanced tokamak concept for ITER and DEMO. Toward this challenging goal, various techniques have been introduced and tested, such as improvement of heating system, installation of ferritic steel tiles to reduce toroidal field ripple, development of new diagnostics, various real-time control techniques. As for the research and development in FY2007 to 2008, items emphasized are i) long sustainment of high integrated performance plasma, ii) sustainment of high βN plasma exceeding no-wall ideal limit, iii) sustainment of high radiation fraction with high H factor at high density, and iv) long-pulse and high-power injection of negative-ion based neutral beam and ECRF. The JT-60U facility will be shutdown in August 2008 for its replacement with the new superconducting tokamak JT-60 Super Advanced (SA) that is being designed and constructed as a part of the Broader Approach (BA) Agreement between Japan and EU. JT-60SA will be operational in about eight years.
The U.S. fusion community, through the U.S. ITER Project Office and the U.S. Burning Plasma organization, has been extensively engaged in the ITER Design Review and in finding solutions for technical issues raised by the ITER STAC. The three facilities (DIII-D, C-MOD, and NSTX) are in the midst of experimental operations. Five year program plans for the period of 2009-2014 have been established for these three facilities through community workshops and peer reviews of their proposals. The U.S. fusion community is also in the process of developing long term strategic plan to identify approaches for addressing gaps in physics and technology beyond burning plasma experiments in ITER, towards a fusion demonstration plan. As a first step, the community identified gaps and opportunities, as documented in the October 2007 report of the U.S. Fusion Energy Advisory Committee on "Priorities, Opportunities and Gaps: Towards a Long-Range Strategic Plan for Magnetic Fusion Energy". This plan is to be developed by June 2010.
The physics-related work in the collaboration is conducted under eight topical areas, six of which correspond to those used in the ITPA. These are Transport and ITB Physics, Confinement database and modeling, MHD, Edge and pedestal physics, SOL and divertor physics, and Steady State Operation. In addition, Tritium and Remote Handling Technologies, and Other issues, such as Diagnostics and Power Supplies issues, are conducted in two separate Task Areas. Accomplishments in these Task Areas are described in Attachment A2.
Two Workshops were held during the reporting period. These were:
The joint Ex-Co meeting of the LT and PD IAs on June 3-4, 2008 provided the opportunity to discuss several issues for future strategy of IEA LT and other tokamak related activities. These are briefly summarized below (with further details available in the minutes of the meeting);
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 program into new era, symbolized by the establishment of ITER Organization, 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.
The IEA Large Tokamak Implementing Agreement is 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. The IEA LT ExCo looks forward to the establishment of a new Tokamak Agreement to streamline tokamak related activities in FPCC, and enhancement its productive interactions with ITPA, IO, and the IFRC.
These reports can be found on the IEA LT 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 at Naka, JAEA, JAPAN.
Objective of JT-60 experiments is to establish advanced tokamak concept for ITER and DEMO. Toward this challenging goal, various techniques have been introduced and tested, such as improvement of heating system, installation of ferritic steel tiles to reduce toroidal field ripple, development of new diagnostics, various real-time control techniques, and so on. As for the research and development in FY2007 to 2008, items emphasized are i) long sustainment of high integrated performance plasma, ii) sustainment of high βN plasma exceeding no-wall ideal limit, iii) sustainment of high radiation fraction with high H factor at high density, and iv) long-pulse and high-power injection of negative-ion based neutral beam and ECRF.
As for the item i) mentioned above, integrated performance of βN = 2.6, HH98(y,2) ≥ 1.0, fBS ≥ 0.4 was sustained for 25s. There are two key points for this success. First one is the extension of pulse duration of 3 co-directed near-perp. NBs from 10s to 30s. This is because we found that these central heating beams are effective for sustainment of ITB, resulting in high βN. Second key point is to weaken MHD modes in the central and edge regions which deteriorate confinement. The MHD mode is still alive even in a good discharge, however, careful adjusting of q95 weakened the activity, resulting in high HH factor
As for the development on long pulse and high power injection of negative-ion based neutral beam, field-shaping plate successfully reduced heat load onto grid. By using this technique, we achieved 30-second injection with one ion source this year. Next target is to demonstrate 100MJ injection by using two ion sources, together with increased acceleration voltage (320-340kV).
Long pulse and high power injection of ECRF have been also progressed. As for long pulse injection, we realized the stable oscillation condition by pre-programmed control of both anode voltage and cathode heater. As a result, pulse duration of the gyrotron to the plasma successfully extended up to 30s where gyrotron power is 0.5MW. As for high power injection, 1.5MW for 1s has been achieved at 110GHz. Here an RF shield for the adjustment bellow and a low loss DC break enabled this achievement.
As for diagnostics, Li-beam injector for Zeeman polarimetry has been put into operation, and beam emission signal has been successfully detected.
As for remote collaboration, JAEA provides JT-60 data within the scope of the proposal document sheet (PDS). There are 8 active PDS since 2007 (3 EU, 2 US, 2 AUG, 1 EAST). Recently two more PDS (1 US, 1 EU) were discussed in the PDS committee in JAEA and procedure for approval is in progress. Another topic of the remote collaboration is the demonstration of remote fusion experiment from Europe with high network security. On the last Christmas Day, remote experiment of JT-60U was successfully carried out from Max-Planck Institute for Plasma Physics of Germany. First, IPP scientists set discharge parameters by using their PC, then send them via Internet to JT-60 control system. During the discharge, they can see visible TV image of plasma and reconstructed plasma shape in real-time. Just after the discharge, they can check the data and discuss the parameters to be changed in the next shot with JAEA physicists in Japan. As a result, fruitful results on NTM suppression were obtained. This is an advance towards remote experiments in ITER and the JT-60SA.
After the completion of 2008 campaign (11 weeks + 2 conditioning weeks), JT-60U will shut down toward the next superconducting device JT-60SA. Objective of JT-60SA is to contribute to ITER and DEMO with enhanced performance of shaping control, heat and particle control, stability control, and heating and current drive. According to the Executive Summary of Conceptual Design Report in 2007, its construction takes 7 years and first plasma is planned in 2015. Procurements are going well as follows. Since last October, three contracts are successfully concluded. First one is the supply of the conductors for EF coils and central solenoid. Second one is supply of the Vacuum vessel. And the third one is supply of the raw materials for in-vessel components such as divertor targets.
Status of U.S. Fusion Program: The Office of Fusion Energy Sciences is engaged in several critical issues for guiding the program's future direction. Among these are intense activity for the overall ITER program direction, including its budget issues and establishing its new Baseline Design, extensive reviews and discussions for a decision on the NCSX Project to deal with its large cost increase and schedule delays, working with FESAC and the community to develop a long range strategy for the U.S. Fusion program, community workshops on establishing a Fusion Simulation Project, and establishing a joint program with the National Nuclear Security Agency (NNSA) on High Energy Density Physics (HEDP). The Fusion Energy Sciences Advisory Committee has been given two new charges on Alternate Concepts and High Energy Density Plasma Program. In addition to these, the National Academy of Sciences issued its report on Decadel Survey of Plasma Science and Engineering, and conducted a review of the US Plan for Scientific Participation in ITER, which was prepared by the U.S. Burning Plasma organization as a part of the requirements by the Energy Policy Act of 2005. All of these activities are expected to provide a multitude of input to the development of Long Range Strategy for the U.S. Fusion program.
The U.S. fusion community, through the U.S. ITER Project Office and the U.S. Burning Plasma organization, has been extensively engaged in the ITER Design Review and in finding solutions for technical issues raised by the ITER Science and Technology Advisory Committee. The three facilities (DIII-D, C-MOD, and NSTX) are in the midst of experimental operations. A Workshop was held at MIT in September 2007 to have community discussions on 5-year plans for these three facilities. Peer reviews of their 5-year proposals are in progress. Further details on technical highlights from the three facilities will be provided by Tony Taylor (DIII-D), Earl Marmar (C-MOD), and Rich Hawryluk (NSTX).
Several ITPA Topical Group meetings were held at MIT, GA, and ORNL in April and May. The 7th workshop on ITPA-IEA Joint Experiments is planned for MIT in December 2008. The U.S. is prepared for continuation of ITPA under the ITER auspices, providing Chairs for the Transport & Confinement, Divertor &SOL, and Diagnostics Topical Groups, and Deputy Chair for the Stability TG.
The last year on JET has seen the successful completion of the 2007 Shutdown and 2008 Restart of Operations, the elaboration and start of the 2008 Experimental Campaigns, major R&D activities and procurements related to components and buildings for enhancements to be installed on JET during 2009/10, the conduct of Fusion Technology Tasks using the JET Facilities, and a proposal for a longer-term JET programme to end 2014.
The 2008 scientific programme has a strong ITER focus (High level commissioning of new systems and issues that could impact on the design of ITER components; preparation of integrated operating scenarios for ITER; and physics issues essential to the efficient exploitation of ITER) and extends over 102 days (including 30% contingency), together with 73 days for commissioning the ITER-like ICRH Antenna (600kW coupled during first tests on plasma in May 2008; target capability with Type I ELMs of 7MW for 5s at 42MHz, together with ≈ 6MW from the A2 antennae using 3dB Couplers and External Conjugate T for ELM-resilient coupling). Campaigns C20a and C20b (comprising 38 two-shift days from 7 April to 6 June 2008) are nearing completion. Other systems to be commissioned and exploited during 2008 include an High Frequency Pellet Injector (for ELM control and deep fuelling (with Russian Federation and US involvement)), two TAE antenna arrays (with US involvement), a Disruption Mitigation Valve and an enhanced diagnostic capability (Edge LIDAR, improved ECE, sweeping reflectometer, fast visible camera, divertor infra-red camera, additional QMBs, improved magnetics, halo current sensors, γ-ray cameras) with detector developments (radiation hard Hall probe, CVD diamonds, Si on insulator, Cellular Non-linear Network (CMOS), activation foils) and more routine use of the Error Field Correction Coils (for Resonant Magnetic Perturbations for ELM control and Resonant Field Amplification for RWM studies), TF Ripple variations (0.08-3%) and diagnostics (High Resolution Thomson Scatterring (with US involvement), Scintillator probe, Faraday cups (with US involvement), CXRS (with US involvement)).
A strong European and International participation is foreseen for 2008 from 21 European Countries (337 people; 70 ppy total; 48 working days per person on average) and with Japanese and US collaborations concentrating on 27 ITPA ITER high priority coordinated experiments, requiring more than 30% of overall run-time and conducted through IEA Implementing Agreements, and Russian Federation collaborations concentrating mainly on diagnostic and code developments.
For the period January-June 2009, two Experimental Campaigns separated by an Intervention (for vertical stabilisation system upgrades for higher resilience against ELMs) are foreseen. For the longer-term JET programme, major R&D activities have been conducted and many procurement packages placed over the last year for enhancements to be installed from mid-2009. These enhancements are of high scientific value and strategic importance (ITER-like combination of first wall materials (tungsten divertor and beryllium wall), NB Power Upgrade (35MW for 20s to allow scenario development at high current, β and density); upgraded and new diagnostics, and a programme of machine refurbishments). From late 2010, the experimental programme will focus on preparation of ITER operational scenarios at high power with acceptable plasma/wall interactions and optimisation of ITER auxiliaries. Critical issues will be the minimisation of T-retention, material erosion and migration, mixed materials effects, melt layer behaviour, impurity control, and the development of ITER scenarios fully compatible with a Be/W material mix. A major challenge will be to accommodate up to 45MW of heating power with the ITER-like combination of first wall and divertor materials.
Since 2000, ninety-eight JET Fusion Technology tasks have been launched (total resources ~16.5M€ (~3.7M€ in 2006), concentrating on tritium in tokamaks, tritium process and waste management, plasma facing components, engineering, and neutronics and safety.
Collaborative work on ITPA-IEA joint experiments was performed.
In the experiments on hybrid scenario (TP-2), no confinement improvement has been so far observed in JET hybrids discharges. Also in JET, beta degradation of confinement is observed, with BτE ∝ 1/βN at the same ρ*, ν*, q~1 and Ti/Te in contrast to previous experiments carried out at low triangularity (δ=0.22)
In the high Te/Ti discharges (TP-3.1) with ECRF electron heating, confinement degradation was observed in DIII-D and JT-60U hybrid scenarios. In DIII-D, the confinement degradation correlated with increase in low-k turbulence. On the other hand, increase in turbulence level was not observed during the confinement degradation in JT-60U.
In the study of effects of low momentum input on transport (TP-4.2), renumbered as TP-4 from 2008), low rotation hybrid discharges in DIII-D project to ITER target of Q>10 at low q95, although fusion performance parameter was reduced with low rotation. In JET, investigation of Ti-ITB trigger events with very low levels of injected momentum using ICRH indicated that ExB shear driven by toroidal rotation was not important with a shear-reversed target q-profile.
In QH mode study (TP-5), improved error field correction in reversed Ip allowed QH-mode at lower torque input and QH-mode with dominant co-NB was obtained. Rotation direction of EHO changes with reduced torque input and higher density in QH-mode. Balanced double-null shape allows extension of QH-mode operating space to higher pedestal pressure.
In spontaneous plasma rotation study (TP-6.1), new points from JET plasmas with low momentum input have been added to the spontaneous rotation database. These new entries fall in with the previous observations. In JT-60U, spontaneous toroidal rotation has been determined using beam modulation techniques. The magnitude of the spontaneous rotation scales with increasing pressure gradient. On C-Mod, numerous rotation profiles (with temporal evolution) have been obtained with the new imaging x-ray spectrometer. Similarity experiments have been performed between TCV and C-Mod to study rotation inversion in L-mode plasmas.
In momentum transport study (TP-6.3), beam modulation experiments showed that χi increases with χi radially and locally, and both decrease as Ip is increased in JT-60U. A large database has been formed from a variety of JET discharges, showing a good correlation between MA and Mi, and effective Prandtl numbers typically around 0.4.
In core microinstabilities (ITG/TEM) study (TP-7), recent experiments on DIII-D have achieved plasma conditions in which the existence of different modes is predicted by codes and a full set of fluctuation measurements is obtained for each case, in the relevant wave number spectrum for these instabilities. No obvious shift in frequency due to a turbulence transition could be observed because the mode phase velocity, as predicted by dedicated theoretical modelling, is an order of magnitude smaller than the frequency shift due to the equilibrium radial electric field. These results are being compared with those previously obtained in other tokamaks, in particular AUG, T-10 and TEXTOR.
In study on rational q effects on ITB formation and expansion (TP-8.2), ITB formation experiment was performed with balanced NBI in DIII-D, where reduction of turbulent fluctuations and transient confinement improvement were observed near integer qmin but no ITB formation. Strong TAE modes and drop in core ion temperature, which had been observed in co-injection case, were absent in the balanced NBI case. Both rational-q and ExB shear are of importance for ITB formation. Studies to date have concentrated on discharges with weakly reversed shear. Physics of the ITB triggering mechanism needs to be investigated with both strongly reversed shear, and narrower regions of weak/zero shear. In MAST, the high-resolution CX measurements of Ti have been done with Δr ~ ρi ~ 1 cm. The 'ITB like' region forms in core on appearance of q = 3/2 surface in L-mode plasma. Location of ITB expands following q = 3/2 surface. Ion channel exhibits local maxima in ρs/LTi close to low-order rational surfaces. Lower resolution of YAG TS system is unable to resolve such detail in ρs/LTe.
In JT-60U/JET ITB similarity experiment (TP-8.3), the effect of TF ripple on the formation and performance of ITBs has been studied on JET for both reversed shear and low/weak shear discharges. The TF ripple was varied on a shot-to-shot basis: δ~0.08% - 0.62% - 0.82% - 1.0%. Large effect of TF ripple on toroidal rotation has been observed. The initial formation (trigger) of internal transport barriers was found to be unaffected by TF ripple. However, the performance of ITBs is reduced at larger TF ripples (i.e. lower toroidal rotation).
The LTA report on confinement database and modeling 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 2007 - May 2008.
CDB-2 Beta Scaling
DIII-D, NSTX and JET each performed additional beta scaling experiments in 2007. There are now confinement scaling trends that have been observed in studies on JET, DIII-D, JT-60U, AUG and NSTX ranging from a null beta dependence to a strong beta degradation. The difference in the confinement scaling as a function of β appears to be associated with shape through edge (pedestal) stability, which can affect, for instance, the character and severity of the ELMs in the scaling experiments. Theory and modelling show that core transport should be ~β-independent for experiments, but a strong overall degradation of confinement with β could occur in lower triangularity plasmas, where electromagnetic instabilities (e.g., kinetic ballooning modes) are particularly important near the edge. Future work should focus on increased analysis of pedestal effects in existing experiments and consider 2-term scalings with more realistic pedestal models.
CDB-4 Scaling with Collisionality and Greenwald Fraction
Prior JET and C-Mod experiments showed that collisionality rather than Greenwald fraction was the key scaling parameter. An additional collisionality experiment at high beta_N requiring high ICRF power and effective cryopumping on C-Mod was proposed for 2007 but did not receive high enough priority to be performed during the experimental campaign this year.
CDB-6 Aspect Ratio Scaling
MAST expanded their H-mode operating space up to 1.2 MA and 3.2 MW in the double-null divertor configuration and found confinement scaling similar to that found in NSTX, namely weaker than linear scaling with plasma current and a strong toroidal field dependence compared to higher aspect ratio devices. Further analysis of NSTX rotation data in the H-mode database showed that the confinement times were not well ordered by either rotation in the core or at r/a = 0.5.
CDB-8 Rho* Scan to ITER
No experiments were performed in this area this year. The aim is to perform a rho* scan with collisionality, beta, q and plasma shape matched to the ITER baseline scenario. Both JET and C-Mod are considering experiments for 2008 but they have not been scheduled. C-Mod must obtain good density control through cryopumping to be able to reduce the collisionality to ITER levels.
CDB-9 Density Peaking Dependence on Collisionality
Past experiments on JET, AUG, C-MOD and JT-60U reported very similar behavior of density peaking increasing as collisionality decreases in H-mode discharges. Experiments this year on C-MOD H-modes clearly indicate an increase in density peaking with increasing q95 that exists over the entire collisionality range (0.3<νeff <5).
JET studies performed in 2006-07 concentrated on identifying further dependencies for νeff <0.5 in H-mode. There appear to be weak additional dependencies on the NBI source, li and/or Ti/Te. These parameters are however correlated, precluding a definite statement on what the governing parameter(s) may be. The li (or q95) dependence in JET H-modes (if any) remains clearly below that observed in L-mode in most (if not all) devices, now including MAST. L-modes typically show little or no νeff dependence. A clear q95 dependence in H-mode (in addition to the νeff dependence) has however been reported from CMOD. Further analysis of such discharges exhibiting simultameously li (or q95) and νeff dependencies may help in developing a unified picture for density peaking in both confinement modes.
Experiments on AUG and JET show that density peaking in helium plasmas exhibit the same collisionality dependence as seen in deuterium plasmas.
CDB-10 H-mode Power Threshold Hysteresis and Access to Good Confinement
This topic was identified as a new area of focus this year. The goal is to operate devices close to the H-mode threshold power in an ITER-like plasma shape and characterize the H-mode properties. Then increase density to ~ 0.8 times the Greenwald density while maintaining the heating power near the low density threshold value to assess and characterize any back transition which occurs. Proposals for experiments in this area were submitted late in 2007 by JET, DIII-D, AUG and TCV. Experiments are planned in JET (Nov. 08) and TCV (after Sep. 08).
CDB-11 L-H Threshold Power at Low Density
DIII-D, JET and JT-60U have observed that the minimum H-mode threshold power occurs at a density of about 2.5x1019 m-3 and rapidly increases below this density. The minimum Pth for C-Mod was found at 8-10x1019 m-3 at BT = 5.4T. Recent C-Mod experiments have explored the scaling of the L-H transition at low density and found that the low density limit was independent of plasma current and decreased nearly linearly with decreasing BT so that at 2.2 T the minimum threshold power was at a density of about 4x1019 m-3, closer to but not quite in agreement with the other devices.
CDB-12 H-mode operation in H and He
During ITER’s non-activation phase of experiments it will operate in H and is considering operation in He which may allow an L-H transition at lower threshold power than obtainable in H. Thus threshold power and characterization of H-mode confinement in H and He are of renewed interest for ITER. At the recent joint CDBM and TP meeting at ORNL it was proposed to create another focus area for this work. Joint experiments between AUG, DIII-D and JET were discussed with experiments planned on DIII-D (Jul 08), later in 2008 for AUG and in 2009 for JET.
MHD physics tasks proposed by the ITPA and implemented under the IEA LTA have been conducted in a range of areas.
Resistive Wall Modes
There has been considerable experimental progress on JT-60U and DIII-D using balanced NB injection. Previous measurements of the critical velocity below which the RWM becomes unstable, using n=1 magnetic braking, have been shown to be too high on DIII-D, probably due to the applied error field seeding locked mode growth. Furthermore, magnetic braking experiments suggest that even the lower rotation threshold obtained using balanced NB injection could still be caused by residual error fields. Detailed modelling using the MARS-F code shows strong sensitivity to the shape of the rotation profile - meaning RWM stability cannot be quantified in terms of the rotation at a given radius (e.g. q=2) - in agreement with results previously reported from NSTX. Detailed data on Residual Field Amplification (RFA) with travelling waves previously obtained in JET, allows direct theoretical comparison with equivalent DIII-D data (although presently there are issues in reconciling the JET data with MARS-F calculations).
Non-resonant Magnetic Braking
Most tokamaks with external error field coils have observed non-resonant magnetic braking e.g. on DIII-D and JET, and this appears to agree quantitatively with neoclassical toroidal viscosity theory in NSTX. It is evident that the braking can be quite significant and in view of the effect of plasma rotation on RWMs, NTMs and locked modes it is important to examine this effect. In 2007 a comparison experiment, using n=2 magnetic braking, between JET and C-Mod commenced and continued into 2008. The experiment is based on matched applied error field spectra in the two machines. In JET plasmas, braking of up to ~50% has been observed, but in C-Mod the braking, if any, was small. Also during 2007, MAST data on magnetic braking from n=2 applied fields was obtained; as in C-Mod, the braking was small and analysis to compare with theory for both MAST and C-Mod is ongoing. Experiments have commenced using the DED (Dynamic Ergodic Divertor) on TEXTOR. So far these experiments have been conducted using the m=3/n=1 configuration in both the direction resonant with the plasma field and in the non-resonant direction.
Previously on DIII-D the co/counter-current NB balance was varied and it was found that reduced rotation, or more likely reduced rotation shear, lowers the threshold for 2/1 NTMs. New joint experiments conducted on NSTX, where n=3 braking was applied to vary rotation, confirm the effect in the extreme geometry of the spherical tokamak. Results from joint experiments on rotation effects on NTMs were reported in an invited talk at the 2007 APS DPP and will be reported in an ITER session at the 2008 IAEA Conference. New data has been obtained on JET for the critical β below which the 2/1 NTM is unconditionally stable. This indicates an approximately linear scaling with ρ* and a weak ν* dependence. On JT-60U suppression of growth of m/n=3/2 NTMs by central co-ECCD has been demonstrated. This suppression of growth occurs with IEC/Ip~0.1 and candidates for this suppression, which are still under investigation, include the current profile or sawtooth changes due to the ECCD.
Massive gas injection disruption mitigation experiments continued on ASDEX Upgrade, DIII-D, TEXTOR, C-Mod and Tore Supra. Results on using gas mixtures of a fast low-Z carrier gas (H2 or He) with a small fraction of high-Z radiation (e.g. argon) are promising in that the high-Z gas is entrained with the carrier minimising the overall response time of the gas injection, while achieving high mitigation effectiveness. DIII-D experiments continued on aspects of gas assimilation into the plasma volume. A real-time locked-mode detector was tested on C-Mod. A new concept for a fast valve, located in the vessel close to the plasma and actuated by air pressure has been developed for ASDEX Upgrade. This valve shows an extremely fast response of the plasma and high plasma cooling and fuelling efficiency.
Progress is being made empirically, and with numerical simulations, towards extrapolation of the massive gas injection disruption mitigation technique to ITER. Gas mixtures appear promising to maximize disruption mitigation efficiency. Experiments foreseen for JET should help elucidate mitigation issues with regards to size and energy of the initial plasma. Better quantification of the runaway loss mechanisms appears critical for extrapolation to disruption mitigation optimization on ITER.
With regard to future plans from June 2008 to May 2009, it is expected that joint experiments on Disruption Mitigation, Neoclassical Tearing Modes, Resistive Wall Modes and non-resonant magnetic braking will continue, together with the related exchanges of personnel.
Coordinated experimental activities and exchange of personnel took place for the following ITPA pedestal and edge topics.
PEP 1 & 3: JET/JT-60U pedestal identity experiments and modelling
No new experiments were carried out in this area for the period covered by this report. Preparation for new similarity experiments is ongoing, with new experiments planned for JT-60U in August 2008. In Autumn 2008, JET will also carry out new H-mode experiments with enhanced Toroidal Field ripple, and JT-60U similarity experiments could be part of this campaign.
PEP 6: AUG/MAST/NSTX pedestal structure and ELM comparison in double null
Experiments were carried out on MAST (July 2007) and NSTX (March 2008). Data of relevance to this experiment has also been collected on JET and Alcator C-MOD. Experiments on ASDEX Upgrade and MAST are scheduled for later in 2008. The MAST experiments in 2007 were unsuccessful because of difficulty in accessing H-mode in lower single Null (L-SN), close to double null (DN) in this particular shape. The same has been observed on NSTX, but was resolved by using a shape with an X-point closer to the target plates. Here the ELM behaviour changes between DN and L-SN from Type-I to Type-V. The pedestal data still awaits analysis. On C-MOD a reduction in pedestal density as well as the pedestal density width has been observed between L-SN and U-SN in the EDA H-mode. No improvement in confinement was observed in DN.
PEP 9: Dependence of the H-mode Pedestal Structure on Aspect Ratio (DIII-D/MAST/NSTX)
Data obtained on DIII-D during 2005 and on NSTX and MAST during 2006 and 2007. A good double-null shape match was obtained in all three machines, along with a heating power and density scan to obtain a range in electron collisionality νe* and normalised ion gyroradius ρ*. Specifically, ELMy H-modes were obtained in all three machines with νe*~1 and ρ*~0.015 matched at the outer mid-plane. To accomplish this, DIII-D was run at toroidal field on axis Bt0=0.52T. In DIII-D, the pedestal widths correspond to 6-8% of ΨN, which is ~50-100% larger than typical pedestal widths at the nominal Bt0=2.1T. Edge stability analysis indicates the plasma in DIII-D was at the boundary for peeling/ballooning instabilities just before a Type-I ELM. In MAST, the pedestal widths measured on the inboard side correspond to 3-4% of ΨN, which is typical of other experiments in MAST. The peak electron pressure gradients were comparable in DIII-D and MAST. The stability analysis in MAST indicates proximity to the ballooning stability boundary. Finally, analysis of the edge Thomson scattering data and peeling/ballooning stability is being completed for NSTX. The research was carried out through both on-site and remote participation of contacts at all three facilities.
PEP 10: Collaborative experiments between MAST and ASDEX Upgrade on the effect of pedestal parameters on ELM radial extent
Experiments carried out on AUG in July 2007 and MAST in May/June 2007 allowed the evolution of the filaments observed during Type-I ELMs on ASDEX Upgrade and MAST to be studied. On both devices it is observed that the filaments start rotating toroidally/poloidally with velocities close to that of the pedestal. The velocity then decreases as the filaments propagate radially. On AUG and MAST, the radial efflux due to Type-I ELMs has been measured for discharges at different IP, BT, ne and PNB. On both devices the ion saturation current e-folding lengths of the filaments show a weak, if any, dependence on the size of the ELM (ΔWELM/Wped). On MAST, the measured radial velocities of the filaments also show a weak dependence on ΔWELM/Wped.
PEP 13: Comparison of small ELM regimes in JT-60U, ASDEX Upgrade and JET
No new results to report
PEP 16: C-MOD/MAST/NSTX small ELM regime comparison
Experiments have been performed on all three devices (C-MOD, MAST and NSTX). The data on MAST were taken in double null (DN) rather than lower single null (L-SN) configuration. On NSTX data was taken in both DN and L-SN. On MAST the small ELMs vanish at high βped and low collisionality, and the structure is distinctly different from Type-V ELMs observed on NSTX in L-SN. The small ELMs observed in NSTX in DN seem also to disappear at high βped. There seems to be no lower βped threshold for the small ELMs in DN. In L-SN both on C-MOD and NSTX the small ELMs appear only at high heating power (high βped), although on C-MOD this seems to be rather more a pedestal temperature effect than βped. Analysis is ongoing to determine whether the structure of the ELMs is similar in C-MOD and NSTX. On MAST further experiments are planned for 2008 to try to access type-V ELMs in L-SN
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 2007 - May 2008.
D/T retention, recovery (DSOL-12, 13). In addition to direct ITPA activity, many members were involved in the ITER Design Review, specifically in ~a dozen ad hoc taskforces during the summer and fall of 2007 in response to a request from ITER DR Working Group No. 1 for expert input on their "Task 5", which covered a range of wall and PFC materials questions. Substantial reports were tabled in September by the EU and US ad hoc taskforces. Iteration, refinement and a degree of convergence has occurred since, a process that continues.
With regard to the present mix of PFC material, C+Be+W, it is not clear at this time whether the T retention will be dominated by C- or Be-codeposition. In either case, the 350g limit could be reached after several 100's (1000's) discharges, assuming tritium-rich (-lean) codeposits, e.g. T/Be ~ 0.1 (0.01) or T/C ~ 0.2 (0.02). The T-content of codeposits depends very strongly on their temperature. The non-saturation of hydrogenic retention in carbon (linear in ion fluence) was demonstrated in the new Tore-Supra dedicated long pulses over 2 weeks. The codeposition of D with Be appears to be more complicated, depending on the incident energy of deuterium, the Be deposition rate and the dependence on surface temperature. Recent results on thermo-oxidation ("O-baking") recovery of hydrogen from C codeposits are encouraging: for JET tiles with Be/(Be+C) up to 60%, the D removal was independent of Be content and codeposit thickness, a contrast with findings for B-contaminated C-codeposits from DIII-D, where B was found to suppress the recovery process; it was estimated that if the ITER codeposits have similar structure and impurity content to the JET tiles, O-baking at ~350 C and >160 torr O2 pressure would remove >85% of the inherent D/T content within a day - independent of codeposit thickness and Be content.
Estimates were also made for the impact on T-retention if ITER went to all-W. For W, codeposition is unlikely to be significant because the absolute erosion rates are smaller than for low-Z PFCs. Long-known data for the uptake of hydrogen in W, based largely on laboratory ion beam measurements, also indicate rather modest rates of uptake. The recent Task 5 activity, however, brought focus to two mechanisms: (a) tritium trapping at damage sites created by 14 MeV neutrons, and (b) an as yet unexplained retention process driven by plasma pressure that may be responsible for the high uptake observed in C-Mod (which employs Mo, however, not W). The estimates for an all-W ITER presented by the EU and US taskforces were close, indicating that the 350 gram in-vessel T limit could be reached at between 2000-3000 discharges (based on neutron-induced trap creation estimates of 0.6% to 1%). The lack of data on 14 MeV damage of materials will make it challenging to refine estimates in the near future. The level of hydrogenic retention in (undamaged) tungsten generally is lower than for carbon based on the factor of 10 reduction in D retention (post-campaign analysis) in ASDEX-Upgrade following conversion to fully-W. While He+ impacting surfaces leads to nanostructure growth (up to ~ microns in depth) on W in a narrow range of surface temperatures, their importance for ITER T retention remains to be evaluated.
ELMs (DSOL-1, 19). Wall and divertor loadings during transients continues to be a difficult subject to understand. Characteristics of ELMs incident on the outer limiter in ASDEX-Upgrade were inferred to be in the range 100 eV per electron-ion pair. New measurements of ELM erosion at the outer divertor of ASDEX Upgrade show that as the local temperature drops the between-ELM erosion also drops relative to that
occuring during ELMs (which is concentrated near the strike points). Increasing the impurity level through impurity puffing may increase divertor radiation but also increases
erosion during ELMs. The current model of ICRF-enhanced sheaths, so-called sheath rectification, were contradicted (C-Mod) in that the plasma potential increases proportionally to PRF and potentials are generated even when the current loop is broken by insulating tiles. The ICRF-enhanced plasma potential has been observed to occur on flux tubes connected to the antenna as well as passing in front of the antenna.
Edge diagnsotics in ITER (Avila ITPA session): the ITER plans for diagnostic coverage of the main chamber, SOL and divertor were reviewed. During the meeting itself it was clear that several measurements were either not available or not well-specified (e.g. wall fluxes, injected gas, dust…). It was agreed that those diagnostics will be reviewed by the SOL/divertor group between meetings for spatial/measurement coverage and specifications. The gas injected during, and gas remaining, after a discharge were discussed at length and following the meeting. The group suggested that some method of H retention must be made during the H phase either through a small fraction of D injected, or through lowering the ambient levels of H in the chamber. Diagnostics for surface temperature measurements were reviewed and several laboratory developments of in-situ surface analysis (ion beam analysis and laser beam erosion/desorption) were presented.
The dust issue (Avila ITPA session). Dust is a very important issue for ITER and was discussed as part of the diagnostics session. Based on a lack of experimental data the IO is assuming that dust is generated at a rate equal to the erosion rate of all surfaces in the machine. The SOL/divertor group felt that the specification should be lower as evidenced by JT-60U results, which were more like 10% of the net surface erosion rate. It is hoped to form a working group to obtain a better specification for the dust generation.
ITER H-phase (Avila ITPA session). A draft plan for the ITER H-phase was presented for comment by A. Loarte (for the IO). A number of assumptions of availability of the machine as well as diagnostic and heating systems were made together with a phased plan for bringing all of those systems up to full levels by staging the powers and plasma current levels at reduced pulse length (20-25s). It will be impossible to properly test the PFCs up to full power-handling capability and the ability to handle transients such as ELMs or disruptions at full levels. Nevertheless close attention will need to be paid to monitoring the various surfaces during the startup phase to determine if they fail at those lower power levels. A second aim during the H phase is to monitor the H retention. While ITER currently has no plan for this it is clear that one is needed and the SOL/divertor group plans to work closely with the ITER diagnostics group to agree on the proper diagnostics as well as a plan of measurement.
The following joint experiments are coordinated by the SSO-TG:
SSO-1: Document performance boundaries for steady state target q-profile.
SSO-2.1: Qualifying hybrid scenario at ITER-relevant parameters.
SSO-2.2: MHD effects on q-profile for hybrid scenarios (joint with MHD-TG).
SSO-2.3: ρ* dependence on transport and stability in hybrid scenarios (joint with TP-TG).
SSO-3: Qualify real-time profile control methods for hybrid and steady state scenarios.
SSO-PEP-1: Document the edge pedestal in advanced scenarios (joint with Pedestal-TG)
SSO-5: Simulation and validation of ITER startup to achieve advanced scenarios.
SSO-6: Ability to obtain and predict off-axis NBCD.
Significant progress was made in the period June 2007 to May 2008 in view of steady state operation development and preparation of ITER operation.
For the documentation of the performance boundaries for the proposed steady state target q-profile (SSO-1), DIII-D and JET were the most active in this area in 2007/2008. JET established stationary discharges with the prescribed parameters and carried out scans of the q profile primarily by changing the timing of the NBI onset and the current ramp rate. The maximum stationary beta varied inversely with qmin. New experiments in JET (2008) are continuing the documentation of the beta limits for advanced scenarios with qmin near 2. DIII-D demonstrated stationary operation consistent with Q=5 in ITER for more than a resistive time under conditions that project to full non-inductive operation. DIII-D reports that a key topic is simultaneous optimization of shape with respect to stability, confinement, and density control. In these discharges the optimum location of the ECCD was studied, for these experiments 5 gyrotrons now routinely available (>3 MW delivered). In JT-60U, fully non-inductive CD with relaxed j profile has been demonstrated in weak-shear regime maintaining qmin above 2 using off-axis LHCD in order to avoid NTM (m/n=3/2 and 2/1). High βN=2.8 discharge sustained for 2s and high bootstrap fraction (fBS=0.9) discharge, both exceeding the no-wall limit were demonstrated. Operation at high βN below and above the no-wall limit has been documented.
For the hybrid scenario (SSO-2.1, SSO-2.2 and SSO-2.3), experiments to study the physics of stability and beta limits in hybrid plasmas were performed. DIII-D has performed studies on the q-profile evolution and obtaining Te~Ti in hybrid discharges. Moreover, DIII-D has developed ITER demonstration discharges. New data on the access to hybrid with ITER-like current rise are now available. In JT-60U hybrid scenario experiments have obtained steady discharge at high beta. ASDEX Upgrade has shown a variation on improved H-mode (hybrid) confinement/performance with small changes in the q-profile. For 2008, ASDEX Upgrade has obtained hybrid operation in a full tungsten machine (without boronisation) matching stored energy levels of previous campaigns. In JET, detailed analyses of the hybrid scenario obtained in 2006-2007 were performed, there are no new data for 2008 as the campaign has only just started.
In order to prepare experiments under SSO-3 ("Qualify real-time profile control methods for hybrid and steady state scenarios"), a large database has been compiled detailing the real time control tools available for each experiment.
The documentation of the edge pedestal in hybrid scenarios (SSO-PEP-1) has made good progress. Dedicated power scans were performed at DIII-D to study the role of pedestal on global confinement in hybrid discharges. The plasma shape was also varied, by matching a low δ ASDEX Upgrade shape in DIII-D at constant q95 and plasma density, so as to differentiate between shape and power variation of the pedestal. Experiments in ASDEX Upgrade in 2008 will complete these studies. The analysis of these experiments is in progress and publications including comparison with results from power scans in ASDEX Upgrade hybrids are in preparation.
On simulation and validation of ITER startup to achieve advanced scenarios (SSO-5): DIII-D, ASDEX Upgrade and JET carried out simulations of the ITER reference startup scenario. The main conclusions were very similar. The small bore startup facilitates rapid current penetration, as designed. However, this leads to current profiles that are marginal with respect to the ability of ITER poloidal coil set to control the vertical position. An alternative startup for ITER was proposed (larger plasma, early X-point formation) and tested successfully on DIII-D and JET, including feedback control of the internal inductance during the diverted phase of the current ramp up. For ITER relevant breakdown studies, Tore Supra, DIII-D and JET have performed new experiments at low loop voltage (matching ITER) and documenting the use of ECRH assist. The results obtained in this area play a significant role in the ITER Design Review discussions. In future, the focus should be on developing requirements for the ITER startup for access to advanced scenario operation and documenting the requirements for the ITER ramp-down phase.
On the ability to obtain and predict off-axis NBCD (SSO-6): The aim of these joint experiments is establish a good target discharge with no observable MHD activity (H-mode at relatively high q95). The neutral beam current drive is to be matched to the ITER case (vertical shift to achieve off-axis current drive). Experiments have been performed at DIII-D and ASDEX Upgrade in April 2008, and in May in JT-60U.
Concerning personnel exchange, in the area of SSO-1 significant remote participation to experiments was done. S. Ide (JA) and C. Challis (EU) visited DIII-D in July 2007. S. Ide, T. Suzuki (JA) and E. Joffrin (EU) participated remotely in DIII-D experiments, while J. Ferron, T. Luce, and M. Murakami (US) participated remotely in JET experiments. For the ITER current rise studies (SSO-5), G. Sips (EU) visited JET in April/May 2008, a Korean team visited IPP in May 2008, S. Ide and C. Challis visited DIII-D in July 2007. For the NBCD studies under SSO-6, C. Challis and J. Hobirk (EU) visited DIII-D (April 2008), J. Hobirk visited JT-60U in May 2008, and T. Suzuki participated remotely in DIII-D experiment.
De-tritiation of plasma facing components at JET
Several activities have been launched in the EU related to in-situ de-tritiation. Apart from the flash lamp technique, laser de-tritiation is the most highly developed technique and in 2006 a laser de-tritiation system was used in the JET Beryllium Handling Facility (BeHF) to clean several tiles. All co-deposited layers were removed in one laser pass and almost all products obtained after treatment were dust. The efficiency of the treatment was confirmed using IBA analysis which showed the complete removal of the film with the fuel trapped in the co-deposited layer. No fuel diffusion was observed in the bulk material after laser de-tritiation. On a mock-up of the inner divertor tiles (3 and 4) the laser accessed and treated remote surfaces. In 2008, this activity continues with optimization of the laser for de-tritiation.
A small plasma torch was developed during 2006 and tested on laboratory samples. The torch operated with active gases such as Nitrogen, and the plasma plume allowed the sample surface temperature to be raised to more than 800C. From the first trials, it appears that the removal rate is much lower than that for laser de-tritiation. However, the chemical processes induced by the torch could allow treatment within castellations.
In the past year, a new system, called Inside Gap Plasma Generator (IGPG), has been developed and could be mounted on a robot to clean castellations using an RF Argon plasma. First trials have shown that the plasma penetrates the void and gaps and a layer of graphite coated on Stainless Steel is completely removed to a depth of 5mm in the castellation after 5 minutes of IGPG operation (Argon, P=27mbar, Pref=50W). It is proposed to use this IGPG and the plasma torch on real samples in the JET BeHF in 2009. The efficiency of the IGPG and the plasma torch will need to be checked via IBA measurements, as has been done for the laser and flash lamp techniques
In 2006-2007, the first results on JET of in-situ Laser Induced Breakdown Spectroscopy (LIBS) were obtained. With this technique it seems possible to obtain the chemical composition of co-deposited layers, including the concentration of T/D/H. However, in the first trials using the JET LIDAR laser characteristic spectroscopic lines were not observed. In 2008, the experiment is being improved to increase the laser fluence on the inner divertor tiles. Ex-situ calibrations are also needed. The concentration of the species in the layer can be extracted from optical spectra obtained during LIBS. This requires modelling activities which are supported by JET, together an with improvement of the LIBS system. In 2009, a tool needs to be designed and developed to allow the installation of a LIBS system on the JET Remote Handling boom for subsequent tests of the hardware in the JET Vacuum Vessel.
In the frame of the ITER-like wall project on JET, activities devoted to the preparation of IR surface temperature measurements in a metal machine have been launched. Based on active IR measurements techniques, they have been assessed in 2007 in laboratory with trial samples. The implementation of the technique was also part of this assessment. In 2009, activity will be launched in order to select the most suitable technique for JET and ITER. An assessment of the implementation in JET including design drawings of such diagnostic will be included in this task.
Surface analysis of plasma facing components from JET
Modelling 13C transport resulting from the puffing of 13CH4 into the outer divertor on the last day prior to the 2004 shutdown has been completed and complemented by further experiments prior to the 2007 shutdown. Puffing was from the outer mid-plane into H-mode discharges. Analysis of a poloidal set of divertor tiles and selected tiles removed from the main chamber during the shutdown has already shown that some 13C has migrated to the inner divertor (Tile 1) whereas negligible 13C has been found in the outer divertor (Tile 8). The analysis will continue in 2008-9.
A set of W-coated tiles were exposed in JET during the 2005-7 Experimental Campaigns to check predicted lifetimes for coatings during the planned ITER-like wall experiment and removed for analysis during the 2007 shutdown. One of the load-bearing base divertor tiles, on which the outer strike point for high triangularity plasmas is situated, was coated with regions of differing W film thickness and the erosion during the 2005-7 campaigns was estimated. Coated tiles, placed where the ILW will use W-coated CFC tiles to protect from NB shine-through and re-ionisation, show no evidence of erosion due to NB effects. However, the 3μm W coating was removed completely to a distance of 10mm into the scrape-off layer from the leading edge of an Inner Wall Guard Limiter (IWGL) tile placed near the mid-plane; this probably occurs when the plasma is in contact with the IWGL during plasma ramp-up.
The analysis of mixed deposited materials is a major topic for ITER. JET provides a supply of Be-C mixed deposits, and in future will provide Be-W and Be-W-C films. During 2008 the Ion Micro-probe and XPS/AES have been tested using JET tile samples to see if they can provide data to augment that obtained from the existing analysis techniques such as IBA and SIMS. Sufficient promise has been shown to let the work continue in 2008-9.
Marker layers have proven to be very useful in determining the extent of erosion/deposition. The pattern of erosion/deposition is likely to be changed considerably with the introduction of the ITER-like wall, which will have W coatings in the divertor and solid beryllium tiles in the main chamber. Each of these regions requires the development of new marker systems in order to measure the degree of erosion and the new locations for deposition (and tritium trapping). The development of marker coatings for Be substrates (using a Be coating on a nickel interlayer) has developed to the final testing stage, and a contract has been placed to develop a marker coating for the W-coated divertor tiles.
Surface analysis of plasma facing components from JT-60
In JT-60, erosion/deposition analyses of the plasma-facing wall have shown that in addition to long-range transport, the local carbon transport to the in-board divertor is also substantial. The total deposition and erosion rates in the divertor region were ~10x1020 C atoms/s and ~-6x1020 C atoms/s, respectively, whereas about 40% of the deposition in the divertor region should have originated from the main chamber wall. The highest hydrogen concentration in (H+D)/C ratio and the retention rate were found to be ~0.13 and 6x1019atoms/s, respectively. In the plasma-shadowed area beneath the divertor region at around 420 K, re-deposited layers of ~2μm thick were found with high hydrogen concentration of ~0.8 in (H+D)/C, which was nearly the same level as that observed in JET. Large deuterium retention was also observed at the main chamber walls which were covered with boron layers. Their (H+D)/C and H+D retention were ~0.16 and ~10x1022atoms/m2, respectively, for the vacuum vessel temperature of 570 K. Integrating the retention over the whole main chamber wall, results in a significant hydrogen inventory.
The transport of carbon generated on the outer divertor region has been investigated using the 13CH4 gas puffing in JT-60. On the surface of the inner and outer dome tiles, the 13C areal density rapidly decreased towards the dome-top. This result indicates the 13C flux to the dome-wing tiles surface decreases toward the dome top and/or the 13C deposited near the dome top is re-eroded. The poloidal distribution of the 13C areal density on the inner divertor tiles peaked a little outboard of the inner strike point. It might be caused by transport through the private flux region. In the outer divertor region, 13C was deposited only in the down-stream direction indicating 13C ion particles are transported down-stream by a plasma flow. To investigate carbon transport more closely, 13C deposition in the gap of the divertor and dome-wing tiles will be analysed.
Tungsten is considered as a suitable plasma facing material for fusion reactor components such as divertor baffles. Plasma purity should be acceptable with tungsten provided the energy of impinging particles is kept below the sputtering threshold of tungsten. However, blistering can occur at the tungsten surface, even if the ion energy is too low to create displacement damage such as vacancies. Tungsten blistering could lead to an instability in the plasma due to impurity release into the plasma core. For this reason, deuterium blistering in the surface region of tungsten has been studied by exposure to high flux (1022D+/m2s) and low energy (38eV/D) deuterium plasmas. For the tungsten samples exposed to the plasma up to 1027D/m2, the blistering and blister bursting were clearly observed. Preliminary position annihilation measurements indicated that the vacancy concentration in the near-surface region of tungsten increased after the deuterium plasma exposure.
As part of the continuing PPPL/JAEA LTA collaboration to understand the physical processes which have limited the performance of the first generation of large negative ion sources, and to use this knowledge to improve the next generation for ITER and other large devices such as JT-60SA, Dr. Grisham made three trips of about two weeks to work with the JT-60U negative ion group, and also devoted a few days of another trip to Japan to this purpose. This LTA collaboration was synergistic with tasks for the Heating and Current Drive ITER design review working group.
During the past year of activity, a novel type of negative ion neutralizer was proposed using a supersonic lithium vapor jet which, if implemented on the ITER neutral beams or on beams for fusion power plants, would reduce the total gas load into the beamline by 75 -80%, resulting in an increase in run time between cryopanel regeneration of a factor of 4 - 5, and making steady state beam operation feasible with advanced pumping techniques that would not be practical with the current gas loads. The lithium jet would also permit higher neutralization efficiencies, reduce the heat load on the accelerator, ion source, and beam duct, and would increase the electrical efficiency and total beam power by 15 - 20% or more.
As part of the ITER design review, study in Japan focused on why the beamlet steering using aperture-offset steering failed to work as it was implemented in the JT-60U accelerator, and whether it could be made to work in the ITER accelerator using a different distribution of offsets. This study is still underway, but appears to be yielding plausible results.
Studies also started this year on an idea that using a hollow beam would make it much easier to compress. This would have the added major advantage that it would make it practical to use a supercusp magnetic filter (more recently called a tent filter) of the sort that was used on positive ion sources in the 1970's and 1980's, and which resulted in much more uniform plasmas than the filters used on the JAEA negative ion sources. The disadvantage of the tent filter or supercusp filter is that the magnetic field reaches a null on the centerline of the source. While not a problem in positive ion sources, this would result in large co-extracted electron currents in a negative ion accelerator.
In parallel with the JAEA collaboration and ITER design activity under the LTA agreement, collaboration also continued on the NIFS NNBI system.
IEA Large Tokamak Cooperation
Workshop Number: W67
SUBJECT: "Control of ELMs and RWM"
Date: Feb. 25 - Feb. 26, 2008
Place: JAEA, Naka, Japan
Name (s) of attendees: (All names of attendees are listed in the attachment.)
Brief description of the activities in the Workshop W67
The workshop W67 “Control of ELMs and RWM”was held as a joint meeting with the ITPA MHD TG meeting and US-Japan workshop on MHD Behavior and Control of Burning Plasma at Japan Atomic Energy Agency Naka Fusion Institute on Feb.25 and 26th 2008 focusing on the Edge Localized Modes (ELMs) and Resistive Wall Mode (RWM) physics, with 65 participants and 28 presentations.
On Feb.25th, experimental and theoretical progress for RWM physics and control was reported. In particular, effects of plasma rotation on the RWM stability were discussed in detail as the central issue: The critical toroidal rotation required for RWM stability decreases with decreasing error field. The decrease in rotation speed inside the q=2 surface prior to the onset of RWM seems to the key factor. Effects oft he 3D structure of the stabilizing wall, plasma response to the stabilizing control, progress of the numerical codes were reported.
On Feb.26th, experimental and theoretical progress for ELM physics and control was reported with an emphasis on ELM mitigation issues in ITER: The design study of RMP (Resonant Magnetic Perturbation) cols for ITER was introduced from the ITER organization. The experimental studies in various devices concluded that n=1, 2, 3 RMP can stabilize ELMs. At the same time, however, avoidance of the locked modes is the critical issue. These experimental evidence and numerical evaluation / prediction of ergodized magnetic field by island overlapping suggested effectiveness of RMP in ITER. Needs for further understanding of the RMP physics and clarification of applicability to a wide range of operational area ( in particular q95 ) were suggested. Effects of plasma rotation on ELM mitigation was also emphasized as the central issue for ELM control.
February 25 (Monday)
- Welcome address T Tsunematsu
- Opening address T Hender
(also T Ozeki, T Evans, Y Kamada , M Okabayashi, K Yamazaki)
- Information of schedule and logistics local organizer
9:40 High beta operation (Chair: N Ivanov)
9:40 - Operation above the no-wall limit in DIII-D E Strait
10:05 - Recent results of RWM study in JT-60U high-beta plasmas G Matsunaga
(Coffee Break) 10:30-11:00
11:00 - 3D effects on RWMs in ITER and ASDEX Upgrade S Günter
11:25 - JET high beta operation T Hender
11:50 Lunch (need ticket)
13:00 RWM control (Chair: M Okabayashi)
13:00 - Error field correction in DIII-D E Strait
13:25 - Challenges of the RWM Feedback at low rotation M Okabayashi
13:50 - Determination of Error Field Resonant Harmonics from Dynamic Magnetic Measurements V Pustovitov
14:15 - Critical beta analyses of JT-60SA plasma G Kurita
14:40 - Wall-unlocking of tearing modes by feedback control P Zanca
15:05 - Suppression of error-field-induced magnetic islands by Alfvén resonance effect M Furukawa
(Coffee Break) 15:30-15:50
15:50 RWM and related code development update
15:50 - CarMa RWM control calculations Y Liu
16:15 - RWM modeling with the inclusion of self consistent kinetic terms Y Liu
16:40 - Extension of MARG2D to RWM S Tokuda
17:05 - Summary of CEMM SciDAC center activity on the modeling of ELMs, RWMs, NTMs, pellet injection, and error field S Jardin
18:30 Reception at Akogi-Club
February 26 (Tuesday)
8:30 Status of ELM issues (Chair: J Menard)
8:30 - Introduction: Present issues of ELM study Y Kamada
8:50 - Present status of ELM control in ITER M Shimada
(Coffee Break) 9:50-10:10
10:10 ELM mitigation and control
10:10 - ELM control experiments on JET Y Liang
10:35 - ELM control and physics on JT-60U K Kamiya
11:00 - MAST ELM control experiments and future plans E Nardon
11:25 - Magnetic modeling of ELM mitigation with the EFCCs on JET E Nardon
13:00 ELM mitigation and control (cont.) (Chair: M Shimada)
13:00 - ELM suppression using RMP in DIII-D plasmas
with ITER similar shapes T Evans
13:25 - Perturbed equilibrium modeling of RMP for ITER J Menard
13:50 - Plasma response on RMPs M Becoulet
14:15 Type I ELM behavior and stability
14:15 - Overview of ELM stability and control results from NSTX J Menard
14:40 - Correlation between ELMing pedestal and ITB in JT-60U Y Kamada
(Coffee Break) 15:05-15:30
15:30 - Effect of equilibrium properties on the structure of the edge MHD modes in tokamaks N Aiba
15:55 - ELM-like activities induced by pressure driven modes at the edge of LHD plasmas F Watanabe
16:20 Discussion on ELM mitigation & control All (Chair: T Evans)
17:20 Summary of IEA Large Tokamak Workshop T Evans & Y Kamada
IEA Large Tokamak Cooperation
Workshop Number: W66
SUBJECT: Sixth Joint Workshop on Large Tokamak, Poloidal Divertor and TEXTOR IA's "Implementation of the ITPA Coordinated Research Recommendations"
Date: 29 - 30 November 2007
Place: EFDA-JET, Culham Science Centre, Abingdon, Oxon., UK
Name (s) of attendees: (All names of attendees are listed in the attachment.)
Brief description of the activities in the Workshop W66
The Workshop was the sixth in the series and was held jointly by the three tokamak-related IEA IA's and ITPA, with the participation of members of the ITER IO. While recognising 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 26 participants, including the 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 six ITPA TG's, representatives of the ITER IO, the Programme Leaders representing 11 major world tokamaks (JET, JT-60U, DIII-D, AUG, C-MOD, Tore Supra, TEXTOR, FTU, NSTX, MAST and KSTAR. Representatives of TCV and the Russian, Chinese and Indian tokamaks were unable to attend the Workshop.
Specifically, the Workshop:
An oral report from the last Workshop (Y. Kamada) was followed by a Report from the ITPA Coordinating Committee Chair (R. Stambaugh) on the status of the implementation of the IEA/ITPA Joint Experiments between the various tokamaks for 2007, a Proposal from the ITPA TG's for new Joint Experiments between the various tokamaks for 2008 (R. Stambaugh) and a Report on the ITER Research Needs (D. Campbell). The Programme Leaders indicated their level of commitment to the new Joint Experiments. In addition, the ITER Principal Deputy Director General (N. Holtkamp) presented the Status of the ITER Project, the Operation plans of the various Facilities were collated by J. Schweinzer on behalf of the Programme Leaders and members of the ITPA Coordinating Committee met with members of the ITER FS&T Department.
The coordinated effort which resulted from this Workshop adds great value to the experiments on the individual Facilities, and will result in personnel and, possibly, some hardware exchanges. The commitments made by the various tokamak Programme Leaders were confirmed, and consolidated by R. Stambaugh, after the Workshop.
This report received by Secretariat on day mo y
|29 November 2007|
|- 08.30||- Welcome||- F Romanelli|
|- 08:40||- Logistics||- M Watkins|
|- 08:45||- Opening Remarks||- Y Kamada|
|- 08:50||- Report from last Workshop||- Y Kamada|
|- 09:15||- Status of IEA/ITPA Joint Experiments (2007)||- R Stambaugh|
|- 09:45||- ITER Research Need||- D Campbell|
|- 10:15||- Coffee Break|
|- 10:30||- Machine Status (5 minutes & 3 viewgraphs each)
- VG#1 – New machine capabilities in CY08 - 09
- VG#2 – Operating Schedule in CY08 - 09
- VG#3 – Research Focus Areas in CY08 - 09
|- (Ch. E Oktay)|
|1. JET||- M Watkins|
|2. JT-60U||- S Ide|
|3. DIII-D||- T Taylor|
|4. AUG||- J Schweinzer|
|5. C-Mod||- B Lipshultz|
|6. Tore Supra||- G Giruzzi|
|7. TEXTOR||- U Samm|
|8. FTU||- A Tuccillo|
|9. NSTX||- S Kaye|
|10. MAST||- B Lloyd|
|11. KSTAR||- M Kwon/J Y Kim|
|- 12:00||- Meeting of the members of the ITPA Coordinating Committee and representatives of the ITER FS&T Department|
|- 13:00||- Lunch|
|- 14:00||- Proposals for IEA/ITPA Joint Experiments (2008)||- (Ch: F Romanelli)|
|1. Transport Physics||- R Stambaugh/P Gohil|
|2. Confinement Database and Modeling||- E Doyle|
|3. Steady State Operation||- W Houlberg|
|4. Edge Physics and Pedestal||- G Sips|
|5. MHD||- Y Kamada|
|6. Divertor and SOL||- T Hender|
|7. Diagnostics||- B Lipshultz (T Donne)|
|- 15:30||- Coffee Break|
|- 16:30||- Break-out session for Programme Leaders|
|- 18:00||- Workshop Adjourns|
|- 18:00||- Continuation of this morning's meeting, of ITPA CC Members and ITER FS&T Department representatives|
|- 19:30||- Dinner at the Coach and Horses, Chiselhampton|
|30 November 2007|
|- 09:00||- Chair : D Campbell|
|- Plenary Discussion of the Proposed Roster of Joint Experiments for 2008||- R Stambaugh|
|- 12:00||- Status of the ITER Project||- N Holtkamp|
|- 12:00||- Lunch|
|- 13:00||- Concluding Discussions||- E Oktay|
|- 14:20||- Concluding Remarks and Close of Meeting||- F Romanelli|
|- 14:30||- END (nominal time)|
|Jin Yong||Kim||National Fusion Research Institute||Korea|
|Myeun||Kwon||National Fusion Research Institute||Korea|
|Erol||Oktay||U.S. Dept.of Energy, Office of Fusion Energy Sciences||USA|
Minutes of the 23rd Executive Committee Meeting for
the IEA Large Tokamak Cooperation Programme
3 - 4 June 2008, General Atomics, San Diego
Building 7, Room 217
|Attendees:||M. Kikuchi (JA)||: Member|
|Y. Koide (JA)||: Alternate|
|K. Kamiya (JA)||: Secretary|
|E. Oktay (US)||: Member|
|R. Hawryluk (US)||: Member|
|E. Marmar (US)||: Alternate|
|T. Taylor (US)||: Alternate|
|P. Gohil (US)||: Expert|
|M. Foster (US)||: Expert|
|R. Nazikian (US)||: Expert|
|M. Wade (US)||: Expert|
|R. Giannella (EU)||: Member|
|M. Watkins (EU)||: Alternate|
|J. Schweinzer (EU)||: Expert|
|M. Kwon (KO)||: Expert|
|C. Pottinger (IEA)||: Expert|
The Twenty Third Executive Committee Meeting for the IEA Implementing Agreement on Cooperation among Large Tokamak Facilities was held at General Atomics, San Diego Building 7, Room 217 in 3 - 4 June 2008. This meeting was held jointly with the ExCo meeting of IEA Implementing Agreement on Tokamaks with Poloidal Divertors; (PD). The participants by the IEA PD ExCo are shown as experts above.
The Committee elected Dr. E. Oktay as the chairman until the next meeting (Dr. M. Kikuchi had been replaced by Dr. M. Mori after 23rd ExCo meeting. Dr. S. Clement-Lorenzo had been replaced by Dr. R. Giannella). The present members of the Executive Committee are shown in Appendix A.
The Committee adopted the agenda, which is attached as Appendix B.
The status and plans of the fusion programs of EU (EFDA-JET and AUG), U.S. (DIII-D, C-MOD, and NSTX), and JT-60U were presented by Drs. M. Watkins, J. Schweinzer, E. Oktay, (T. Taylor, E. Marmar, and R. Hawryluk), and Y. Koide. The status reports are attached as Appendix C. Also, the status and plans of KOREA (KSTAR) was presented by Dr. M. Kwon.
The list of Task Coordinators are appended in Appendix D1. The activities of the Tasks (submitted reports) are attached in Appendix D2. The presentations from each device will be uploaded on the LT web page.
The detailed presentations will be uploaded on the following Web-site; http://www-jt60.naka.jaea.go.jp/lt/
The ITER issues in terms of programs on PFC materials and experiments in each device were reported; C-Mod results (E. Marmar), AUG results (J. Schweinzer), JET plan and preparations for ITER-like wall project. Also issues in terms of international plan for ELM-Control studies were reported by M. Wade (for A. Leonard). Furthermore, Experiments for the last operational campaign on JT-60U, including present plan and schedule, were reported by M. Kikuchi.
The latest experimental results from C-Mod and ASDEX Upgrade (AUG) were presented by E. Marmar and J. Schweinzer, respectively, followed by a presentation of M. Watkins dealing with the JET project for an ITER-like wall.
A major part of the C-Mod report dealt with the effect of RF-enhanced sheaths which result in ICRF impurity production. This ICRF sheath enhancement appears on field lines that are connected to the active antenna. Probe measurements point to 100V/1MW ICRF which accelerate D+ above the sputtering threshold for Mo. Even more dramatic is the energy gain of low Z impurities which increase sputtering of Mo and B layers. The attempt to reduce Mo influx and thus to improve plasma performance by replacing Mo antenna limiters with insulating BN tiles was not successful. By mapping of open field lines possible source locations have been identified.
The experience on ASDEX Upgrade with ICRF heating in an all-tungsten device is very similar to the C-Mod results. A strong increase of W erosion on the antenna limiters as well as of the central W concentration goes along with ICRF heating. In the discussion in became clear that a high-Z wall together with the present day design of ICRH antennas is not a solution for ITER or DEMO. Plans exist on both C-Mod and AUG to improve the antenna design.
The second important topic in this session concerned the issue of hydrogenic retention in metal walls and its comparison with carbon. After the 2007 operation of the all-W AUG, surface analysis of divertor tiles showed that 0.3% of the injected D2 gas was retained in W tiles. On carbon PFCs typically 4% of the injected D2 gas was found in previous campaigns. Thus, long term D2 retention is in the all-W AUG reduced by one order of magnitude. The D-inventory in the all-W AUG is dominated by D2 trapping in the outer divertor rather than by co-deposition in carbon/boron layers on the inner target tiles. The avoidance of boronisations during 2007 provided a clear test bed for these post campaign investigations, because co-deposition with B would increase the D inventory significantly. Gas balance measurements (rather high error bars) on a shot by shot basis showed a similar reduction of D retention. In addition, a clear wall saturation at very high D2 levels was observed.
The C-Mod presentation dealt with gas balance measurements on a shot by shot basis. No saturation effects have been observed. High fractions of retained D are found. The long term retention was reported to be close to zero which is explained by an effective removal of D caused by regular disruptions. Due to this effect the net long-term retention on C-Mod is close to zero. In 2011 C-MOD will start operation with a heated (600 C) W divertor. Comparing the retention results from AUG and C-MOD the most striking difference is the effect of disruptions on C-MOD's long-term D inventory.
M. Watkins reported on the status of the ITER-like wall project of JET. In June 2009 a shutdown will start with the aim to install a tungsten divertor together with a Be main chamber wall. In the second half of 2010 JET operation will resume. This project together with the AUG tungsten programme demonstrates the strategy of the EU to prepare the use of W as a first wall material in ITER since a tungsten first wall is at present the sole candidate for DEMO though there remain unsolved issues.
In the discussion the US (T. Taylor, GA) seemed to be disappointed by 'only' a factor of 10 less D retention measured in the all-W AUG compared to a carbon dominated inner wall. There was a discussion how this result extrapolates to ITER. The amount of D trapped by co-deposition with carbon increases linearly with particle fluence (or time), while the amount of D trapped in W increases only with the square root of fluence. This behaviour results from laboratory experiments and will lead to an even more pronounced reduction of the D inventory in a long-pulsed W machine compared to one with carbon tiles. In addition, higher wall temperature in ITER will lead to less D retention.
Japanese side (M. Kikuchi) questioned whether JET will continue with their W divertor plans, keeping in mind that the outcome of the AUG programme might be negative. The group was quite surprised by this question and did not interpret the present AUG results as negative.
Both carbon and tungsten have issues as a material for the divertor/first wall in long pulse burning plasma experiments and further research is needed to resolve these issues.
M. Wade presented for A. Leonard and the ITPA Pedestal Physics Working Group the research and development plan for developing ELM control capabilities for ITER as developed and proposed by the ITPA Pedestal Working Group. This plan included activities aimed at improving the physics understanding of established ELM control techniques and their impact on other aspects of ITER scenarios as well as research activities that could lead to new means of ELM mitigation/suppression in ITER. The primary ELM control techniques that are presently being pursued for use on ITER include ELM pacing using pellets, the use of resonant magnetic perturbations (RMPs), pellet pacing, and the exploration of small or ELM-free H-mode regimes.
M. Watkins inquired as to the process through which this plan was requested and developed. M. Wade indicated that the formal request was made by the ITER IO physics team based on a recommendation from the ITER Science and Technology Advisory Committee (STAC). T. Taylor indicated that the STAC recommendation was made in response to the ITER IO's determination that ELM control was a "mission-critical" issue for ITER and the associated recommendation for the implementation of non-axisymmetric RMP coils in the ITER design. M. Wade commented that this is new method of operation for the ITPA working groups. Previously, physics topics for the ITPA had primarily determined their physics topics in a bottoms-up grass root means. The definition of a physics topic to be focused on by the ITPA by ITER management is a first. It was agreed by the group that this was an appropriate means of defining the ITPA activities and the associated joint experiments given the importance of ITER in international fusion development.
M. Kikuchi suggested that joint research be considered between DIII-D and JT-60U to develop a common physics understanding of the role of rotation on ELMs. Experiments on JT-60U have shown that the ELM character changes significantly as the rotation is varied with small ELMs observed in counter-rotating plasmas. This suggests that the ideal MHD picture of ELMs may not be adequate in describing all aspects of the ELM onset and character. T. Taylor indicated that DIII-D does not see a significant difference in the type 1 ELM character based on rotation alone. Therefore, more experiments are needed to reconcile the DIII-D and JT-60U observations.
E. Marmar suggested that the scope of the plan not be solely focused on H-mode plasmas as C-mod had recently obtained good confinement in plasmas that appear to have an L-mode edge with respect to particle transport, while having a substantial temperature pedestal. These experiments are performed at high input power and low q95 (~3), with the grad-B ion drift direction away from the X-point. The committee was open to this suggestion.
Workshops and personnel assignments completed in the period of June 2007 - May 2008 are listed in Appendix E1. Two workshops on "6th joint WS of LT (W66) PD and Textor IA's on implementation of the ITPA Coordinated Research Recommendations", and "Control of ELMs and RWM linked with US-Japan WS" (W67) were carried out. The total number of personnel assignments completed in the period was 25. One PA was participation (more than 4 weeks), and the others were for review tours (less than 4 weeks) (see Appendix E2). Subjects are summarized as follows (see Appendix E3): Task 1 (Transport and ITB Physics) was 4 (16%); Task 2 (Confinement database and modeling) was 1 (4%); Task 3 (MHD, disruptions and control) was 5 (20%); Task 4 (Edge and pedestal physics) was 4 (16%); Task 5 (SOL and divertor physics) was 0 (0%); Task 6 (Steady State Operation) was 2 (8%); Task 7 (Tritium and RH Technologies) was 0 (0%); and Task 8 (Other) was 9 (36%). 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 5 and 7.
Proposed Workshops and Personnel Assignments for June 2008 - May 2009 are listed in Appendix F. This includes one postponed Workshop as a new one (W68: "Development of high βN scenarios for ITER"). The Committee discussed these proposals and authorized their implementation.
~Restructuring of the tokamak related IAs (LT, PD, and PWI in TEXTOR).~
A brief background; this topic has been discussed by the ExCos of the three IAs and by the IEA Fusion Power Coordinating Committee (FPCC) for the past ~ 5 years. The key motivations for these discussions at different times have been:
The consensus of these FPCC discussions was:
Consensus of the June 3-4 meeting:
Following the FPCC recommendations, the participants at the LT and PD ExCo meetings discussed these issues with an aim to develop a consensus on the future direction of the tokamak IAs, for further discussion with the following absent members of the LT and PD ExCos, and with Uli Samm, the Chair of the IEA PWIT IA, within a month:
The consensus of the June 3-4 meeting is as follows:
"The LT and PD ExCos wish to create a new IEA Tokamak Implementing Agreement, and invite the PWIT to join this IA. While these two ExCos respect the wish of the PWIT to extend its activities on PWI on tokamaks and stellarators, they note that this topic is an integral part of activities in both the LT and PD collaborations and these would be duplicative if the PWIT did not join the new Tokamak IA. The LT and PD ExCos note that the special area of plasma-surface interactions studied in linear devices such as Magnum PSI and Pisces, in particle beam-based high heat flux test facilities such as GLADIS and in the material sample exposure program planned for TEXTOR would benefit from a coordinated collaborative activity in the IEA FPCC. These activities might be better carried out under the PWI annex of the IEA NFT IA or, possibly, a newly formulated version of it'.
If the PWIT ExCo agrees to join a single tokamak IA, a small committee of the three IAs would be asked to carry out the following action items to implement the consensus:
1) The process to achieve a single tokamak IA, either through modification of one of the agreements or by starting a new agreement;
2) Identify a list of Parties that would participate in the IA, and the tokamak facilities in those countries
3) Establish a drafting committee to
a) develop a list of scientific tasks and procedures for operations in the IA. (This work was started at the June 3-4 meeting by adapting tasks from the existing tokamak agreements. This draft includes a statement on including implementation of ITPA joint experiments as being one of the roles for the new Tokamak IA).
b) Describe operational and management procedures for the Executive Committee
If the PWIT does not wish to join a single tokamak IA, the LT and PD ExCos would proceed with the establishment of a single Tokamak IA to include all tokamaks except TEXTOR.
A new issue arose in these discussions on interactions of the proposed Tokamak IA with the Spherical Torus IA. It was noted that a new ST agreement became operational in 2007 and that it fosters collaboration and community building among a variety of experimental groups of different sizes, including some experimenting on very small size devices. The two major STs in the ST IA, MAST in EU and NSTX in the U.S., participate in ITPA-IEA joint experiments extensively. Thus it would be appropriate to enable MAST and NSTX to participate in joint experiments within the proposed Tokamak IA.
The aim is to provide a final consensus of the three tokamak related IAs on their future direction to the FPCC in September 2008 for inclusion in its new Mandate, for submission to CERT in October 2008. The preparations for a new tokamak IA would then start after CERT action on the proposed FPCC Mandate, in close coordination with the IEA secretariat of the FPCC.
A further update of this consensus will be provided to the ExCos of the three tokamak IAs following the informal discussions around the ITPA CC meeting at Aix en provence on June 30-July 1, 2008.
2) Remote Televideo Participation: A few of the IEA LT exchanges from DIII-D to JET include several remote participants from DIII-D to accompany a lead DIII-D participant on site at JET. A written process should be developed to guide such remote televideo participation, and the names of the remote participants should be included in the 'Form A' used to approve the travel of the lead collaborator for the exchange.
In general, remote televideo participation should be encouraged to minimize travel and gain experience in the 'sociology' of remote participation. With the increased scientific interactions among the fusion communities and the need to prepare for remote participation in ITER, the current collaborations provide the opportunity to develop this tool.
3) Sponsorship of IAEA TCMs: The IAEA recently announced that it would reduce its fusion activities and discontinue sponsorship of two TCMs on H-Mode and Energetic Particles. These TCM's, held every two years, have been ongoing for over ten years. The last H-Mode TCM was held in Japan in February 2008 and it attracted about 200 participants. The ITPA Topical Groups on Transport Physics and Confinement Database hold their workshop immediately after the TCM, to benefit from the broader expose of H-mode issues at the TCM. A benefit of these TCMs is the publication of the TCM conclusions in refereed journals that contributes to the dissemination of fusion knowledge.
The ExCos discussed the pros and cons of the three IAs taking over the sponsorship of these two TCMs. The consensus is that IAs should sponsor these TCMs, which provide the basis for travel of scientists to the meetings. Discussions will be held with the IAEA to explore joint sponsorship under the IEA procedures.
The IAEA TCMs in the past helped the Nuclear Fusion journal to reduce periodic (two year cycle due to IAEAFEC) fluctuations of number of paper submissions through the cluster papers and special issues on various IAEA TCMs. The chair of Nuclear Fusion Board of Editors, M. Kikuchi expressed his wish that IEA IA's and ITPA's proposals on cluster papers/special issues to compensate reduction of IAEA activity on TCM.
4) Exchange of information on selected critical issues for ITER: Considering that the programs participating in the IEA LT/PD IAs have much to contribute to ITER physics issues, two topics were selected for some information exchange at this meeting. These were on 'PFC materials and experiments' in C-MOD, ASDEX-U, and JET, and on 'ELM Control issues'. These discussions were useful and feed into the planning of future scientific exchanges.
5. End of JT-60U and start of KSTAR: It was noted that one of the three original members of the IEA Large Tokamak Facilities, JT-60U, will be shutdown at the end of this August, to be replaced by the JT-60SA, which is expected to start operations in about seven years. The participants discussed the plans for the last set of experimental campaigns on JT-60U before its permanent shutdown. It was also noted that the newest tokamak in the international community, KSTAR is in its commissioning phase for first plasma in the next few weeks. KSTAR is a participant in the IEA PD IA. Myeun Kwon shared information on the successful commissioning of KSTAR.
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 2008.
The next Executive Committee Meeting will be held in May, 2009 in JET (EU). It will be a joint meeting with PD IA.