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 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; effect of q profile on triggering high confinement and fast particle induced MHD instabilities; real time control of plasma profiles. Also ITER scenario development is a major element of JET in EU, DIII-D in the U.S. and JT-60U in Japan.
Through this IA, experiments, theory and modelling in these topical areas, especially joint experiments requested by ITPA were performed using JET (EU) and JT-60 (Japan) with contributions from the U.S. national devices such as DIII-D, CMOD and NSTX. A workshop on "Implementation of the ITPA coordinated research recommendations" was held at JAEA, Naka, Japan on Nov.-30 to Dec.-1, 2006. This activity is maturing and making substantial contributions to the advancement of tokamak research for ITER. In this 5th in the series of such workshops, leaders representing 17 major world tokamak programmes from the seven ITER partners (European Union, Japan, Korea, the People's Republic of China, the Russia Federation, the United States and the India), members of the Executive Committees of the IEA LT, Pumped Divertor (PD) and TEXTOR IA's, the ITPA and the ITER IT were among the participants. It was noted that remote participation in experiments can be effective. 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 modelling activities was also recognized.
In the EU, the last year on JET has seen the completion of four Experimental Campaigns; the procurement, assembly and testing of enhancements for installation in 2007 and exploitation during 2008; R&D activities and placement of procurement packages for enhancements for installation in 2009 and exploitation from 2010; and a continuation of Fusion Technology activities using the JET Facilities. The Experimental Campaigns (103 S/T days; Autumn 2006 - Spring 2007) had 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), a strong European (18 European Countries) and International participation (US, Japan and Russian Federation, concentrating on joint experiments, diagnostics and codes), and were very demanding on all systems (only a very limited number of programmes required less than 18MW NB power and, often, high ICRH/LH power). Many new enhancements were exploited (new divertor target for high power, high triangularity operation; 15 new/upgraded diagnostics/systems, including 4 with significant US involvement). The experimental programme included 32 ITPA ITER High Priority Coordinated Experiments conducted through the IEA Large Tokamak Implementing Agreement and requiring 30% of the overall run-time; 80% of JET's commitments were completed during 2006/7.
In Japan, the president of JAEA has changed from Mr. Tonozuka to Mr. Okazaki since January 1st, 2007. A change in organization of fusion directorate is new appointment of DDG Y. Okumura in charge of Rokkasho BA project and formation of division of Rokkasho BA project. Three new group were formed in JT-60 divisions, JT-60SA design integration group, JT-60SA torus development group and safety assessment group for the JT-60SA program. As for the promotion of national program, remote experiment from Kyoto University was demonstrated successfully. And we are planning to have remote experiments from abroad. Experimental collaborators from Japanese universities are ~160 for FY2007. As for the international collaboration, remote access to JT-60 data was performed for 9 PDS (proposal document sheet; 5 EU, 3 US, 1 AUG, 1 EAST) in 2006. Six PDS (3 EU, 1 US, 2 AUG) are approved for 2007 campaign. New data control rule has been set up for provision of experimental data. New hardwares prepared during the shut-down phase (Oct. 2006- Oct. 2007) of JT-60Uare H&CD system (3 more perp. NBIs are ready for 30s injection, N-NBI and ECRF will address 25second injection), improved alignment and insert of C sheet in divertor dome tiles, improvement of CXRS system (17.5ms to 2ms time resolution) and installation of Li beam diagnostics.
In the U.S., the Fusion Energy Sciences (FES) Program is entering a new era, fueled by several promising factors including a new leadership in the Office of Fusion Energy Sciences (OFES), the signing of the ITER Agreement, potential for increased funding through the American Competitiveness Initiative (ACI). The USBPO is coordinating the scientific activity in the U.S. FES program in a broad area of burning plasmas and to prepare for the ITER scientific program. The U.S. government has initiated the ACI, which aims at doubling the funding for the physical sciences in the next ten years, which is conducted by the Department of Energy - Office of Science, the National Science Foundation, and the National Institute of Technology. The FES program has already benefited from the ACI in FY 2007 and FY 2008 by obtaining new funds for most of the large increase in the U.S. budget for ITER, which is $ 60 M and $ 160 M, respectively. Thus, the non-ITER part of the U.S. program remains adequately funded to prepare for the ITER scientific program and maintain a strong science focus. The three major facilities (DIII-D, C-MOD, and NSTX) are operating 12-15 weeks each, and the construction of NCSX continues. The community is preparing for establishing a new long term Fusion Simulation project (FSP) with project management discipline.
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 implementation of the ITPA coordinated research recommendations was successfully started in 2002 under the IEA LT IA. This joint experiment arrangement among JT-60U-JET-USDOE has been strengthened to prepare for the successful start-up of ITER operation with wider participation from the other IEA/IA's and Bilateral Agreements. The ITER organization decided in November to invite ITPA to operate under ITER auspices, while legal framework of its implementation continues to be IEA/IA's and Bilateral Agreements. Our IA (IEA/LT IA) will increase interaction with other IA's and Bilateral Agreements to find more consolidated legal structure.
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. Please visit the homepage of the LT IA to understand the activities carried out and to send comments for improvement.
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.
The Status and Plans of Three Parties
The last year on JET has seen the completion of four Experimental Campaigns; the procurement, assembly and testing of enhancements for installation in 2007 and exploitation during 2008; R&D activities and placement of procurement packages for enhancements for installation in 2009 and exploitation from 2010; and a continuation of Fusion Technology activities using the JET Facilities. The Experimental Campaigns (103 S/T days; Autumn 2006 - Spring 2007) had 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), a strong European (18 European Countries) and International participation (US, Japan and Russian Federation, concentrating on joint experiments, diagnostics and codes), and were very demanding on all systems (only a very limited number of programmes required less than 18MW NB power and, often, high ICRH/LH power). Many new enhancements were exploited (new divertor target for high power, high triangularity operation; 15 new/upgraded diagnostics/systems, including 4 with significant US involvement). The experimental programme included 32 ITPA ITER High Priority Coordinated Experiments conducted through the IEA Large Tokamak Implementing Agreement and requiring 30% of the overall run-time; 80% of JET's commitments were completed during 2006/7. The achievements include varying the Toroidal Magnetic Field Ripple (from 0.08-1%) to study its effect on ELMs, H-mode pedestal, plasma and fast ion confinement; a JET record coupled power in H-mode in excess of 30MW; extension of AT scenario performance in βN and Ptot; extension of ITB discharges to lower q95~5 and higher power (32MW) with Ti~Te, and high core and edge densities with ITER-like shape; active ELM control with more than 30MW with ITB and extrinsic impurity seeding (neon) to reduce divertor heat loads; first full edge characterisation of an ITER-AT plasma; sustainment of hybrid regime extended to low q95~3 and high βN~3 with confinement comparable or slightly better than H-mode; extension of hybrid discharges to 20s, with a record NB injected energy of 186MJ; long distance (~0.15m) LHCD and ICRH coupling at high δ, as required for ITER; documentation of gas balance in high d discharges with Type I and III ELMs and in L-modes; use of Error Field Correction Coils for ELM control. For installation during 2007, enhancements have been procured, assembled and tested (ITER-like ICRH Antenna, 7MW power, ELM-resilient coupling; High Frequency Pellet Injector for ELM-pacing and fuelling; additional diagnostics). For the longer-term JET programme (2007-2010), major R&D activities have been conducted and many procurement packages placed for enhancements to be installed in 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, 34MW for 20s; vertical stabilisation system upgrades for higher resilience against ELMs; upgraded and new diagnostics; machine refurbishments). From 2010, the experimental programme will focus on preparation of ITER operational scenarios at high power which are fully compatible with the ITER-like combination of first wall and divertor materials. A major challenge will be to accommodate up to 45MW of heating power. Critical wall-related issues will be the minimisation of T-retention, material erosion and migration, mixed materials effects, melt layer behaviour and impurity control. Ninety-eight JET FT tasks have been launched since 2000, concentrating recently on issues relevant to ITER licensing.
The president of JAEA has changed from Mr. Tonozuka to Mr. Okazaki since January 1st, 2007. A change in organization of fusion directorate is new appointment of DDG Y. Okumura in charge of Rokkasho BA project and formation of division of Rokkasho BA project. Three new group were formed in JT-60 divisions, JT-60SA design integration group, JT-60SA torus development group and safety assessment group for the JT-60SA program.
Main foci of JT-60U experiments are high βN exceeding no-wall limit, integrated performance in ITER relevant regime, and real time control of high bootstrap current fraction plasma. We have good collaborations on ripple effect and SSO with JET, divertor related collaborations with AUG, and RWM physics with DIII-D. Especially, RWM collaboration with DIII-D is quite effect in identifying critical rotation speed (~15km/s) of RWM.
As for the promotion of national program, remote experiment from Kyoto University was demonstrated successfully. And we are planning to have remote experiments from abroad. Experimental collaborators from Japanese universities are ~160 for FY2007. As for the international collaboration, remote access to JT-60 data was performed for 9 PDS (proposal document sheet; 5 EU, 3 US, 1 AUG, 1 EAST) in 2006. Six PDS (3 EU, 1 US, 2 AUG) are approved for 2007 campaign. New data control rule has been set up for provision of experimental data.
New hardwares prepared during the shut-down phase (Oct. 2006- Oct. 2007) of JT-60Uare H&CD system (3 more perp. NBIs are ready for 30s injection, N-NBI and ECRF will address 25second injection), improved alignment and insert of C sheet in divertor dome tiles, improvement of CXRS system (17.5ms to 2ms time resolution) and installation of Li beam diagnostics. Main goals of 2007 experimental campaign ( 8 weeks +4 conditioning weeks) are sustainment of high beta plasma in a range of βN =2.5-3.0 close to ITER steady-state operation for 25-30s, and real time profile control in high βN and high Ibs/Ip plasma.
The construction of JT-60SA will start this year. JT-60SA is capable to produce ITER shaped plasma and plasma shape not foreseen in ITER. JT-60SA will also contribute to optimization of ITER divertor with V shaped divertor target.
New cooperation arrangement between IPP and JAEA on advance tokamak research becomes in operation. Also JAEA - China (SWIP and ASIPP) cooperation agreement has been concluded.
The U.S. Fusion Energy Sciences (FES)Program is entering a new era, fueled by several promising factors including a new leadership in the Office of Fusion Energy Sciences (OFES), the signing of the ITER Agreement, potential for increased funding through the American Competitiveness Initiative (ACI), and the progress that is being made in fusion research and development worldwide. Prof. Ray Fonck of University of Wisconsin ha commenced his position as the Associate Director for Fusion Energy Sciences in the Office of Science as of the beginning of March 2007. He brings a new perspective to the leadership of the FES program through his technical background, membership in several high-level panels that reviewed and provided direction to the U.S. program during the past five years, and through his leadership in establishing the U.S. Burning Plasma Organization (USBPO) during the past three years. The signing of the ITER agreement plants a major burning plasma experiment in the international fusion program as a real project and provides a focus to the direction of the U.S. program. The U.S. ITER Project Office (USIPO) at Oak Ridge National laboratory is the designated U.S. Domestic Agency for managing the U.S. contributions to the ITER Agreement. The USIPO is working closely with the international ITER Organization (IO) and the ITER Members in developing cost, schedule, and R&D plans for timely implementation of the U.S. contributions to ITER. The USBPO is coordinating the scientific activity in the U.S. FES program in a broad area of burning plasmas and to prepare for the ITER scientific program. As a part of the requirement of the 2005 Energy Policy Act, which authorized the U.S. participation in ITER, the USBPO developed a Plan for U.S. Scientific Participation in ITER. This plan will be reviewed by the National Academy of Sciences this year. The Director of USBPO is also the Chief Scientist for USIPO, thus providing a close coupling of the USIPO tasks within the ITER Agreement, and the 'voluntary' scientific contributions the U.S. makes to ITER on technical issues. The U.S. members of the International Tokamak Physics Activity (ITPA) serve as the international component of the USBPO. The USBPO is making major contributions to the ITER Design review activities. The U.S. participation in all aspects of the ITER program is well integrated through USIPO, USBPO (including US part of ITPA), and the Virtual Laboratory for Technology (VLT). The U.S. government has initiated the ACI which aims at doubling the funding for the physical sciences in the next ten years, which is conducted by the Department of Energy - Office of Science, the National Science Foundation, and the National Institute of Technology. These three agencies will compete to maximize their share of the doubling of the overall physical sciences budget. The FES program has already benefited from the ACI in FY 2007 and FY 2008 by obtaining new funds for most of the large increase in the U.S. budget for ITER, which is $ 60 M and $ 160 M, respectively. Thus, the non-ITER part of the U.S. program remains adequately funded to prepare for the ITER scientific program and maintain a strong science focus. The three major facilities (DIII-D, C-MOD, and NSTX) are operating 12-15 weeks each, and the construction of NCSX continues. The community is preparing for establishing a new long term Fusion Simulation project (FSP) with project management discipline. As an outcome of Dr. Ray Orbach being designated as the newly established Undersecretary for Science in DOE, opportunities are increasing for coordination of research among the different groups of the DOE. A joint program is being established between the OFES and the National Nuclear Security Agency (NNSA) on High Energy Density Laboratory Physics (HEDLP).
Collaborative work on ITPA-IEA joint experiments was performed. In study on steady-state scenario (TP-1), high βN ITBs scenario has been developed in JET at 2.3T/1.5MA with ELM control using neon/deuterium injection, where ion and electron ITBs have been formed at βN~2.9 close to the ideal no-wall limits.
In the experiments on hybrid scenario (TP-2), operation regions have been extended to lower q95 (down to 2.7), higher beta (up to 3.6, above no-wall limit), higher density (up to Greenwald density), and longer duration (up to 20 s) in JET.
ITB degradation with ECRF electron heating (TP-3) was studied with newly installed fast sampling (5ms) CXRS in JT-60U weak positive shear plasmas, where plasma current and deposition location of ECRF was changed.
On low momentum effects on transport (TP-4.2), degradation of electron ITB was observed by changing the toroidal rotation from co- to counter-direction in JT-60U high βp H-mode plasmas. In DIII-D hybrid scenario plasmas, the confinement has been reduced by 10-30% due to reduction of ExB shear when the toroidal torque was reduced. Experiments were carried out on JET to create ITBs using dominant ICRH in a 3He minority ion heating scheme with a shear-reversed q-profile. The toroidal torque applied to the plasma was varied, while keeping the ion heating power constant by varying the mixture of heating powers, ICRH (6-0MW) and NBI (1.5-7 MW). The toroidal rotation varied from 10-60 kRad/s with no degradation of the ITB at low rotation speeds. MSE measurements of the q-profile indicate that ITBs are triggered when the q-minimum point crosses an integer (qmin=3 and qmin=2).
In QH mode study (TP-5), the transition from QH-mode to ELMing and vice versa was found at the same pedestal toroidal rotation value in JT-60U, which appears to indicate a lack of hysterisis in this quantity.
In spontaneous plasma rotation study (TP-6.1), toroidal rotation to the co-direction was measured in JT-60U EC heated ELMy H-mode plasmas without NB injection except beam blip. The observed rotation speed was similar to the scaling obtained from the results in other devices including Alcator C-MOD and DIII-D.
In JT-60U/JET ITB similarity experiment (TP-8.3), a series of experiments has been carried out analysing the effect of TF ripple on ITBs in JET. At larger TF ripple amplitude the outer part of the plasma was found to rotate in counter direction, while in the core rotation in co-direction remained. The initial formation of an ITB was observed in all discharges with reversed shear, though in discharges with larger TF ripple (>0.5%) only weak ITBs formed (ρ*T<0.02). The formation of ITBs in low positive shear discharges was found to be more difficult for higher TF ripple values.
In study on rational q effects on ITB formation and expansion (TP-8.2), reduction of turbulent fluctuation levels was observed near integer qmin in balanced NBI case as well as in co NBI case in DIII-D. However, strong TAE modes just before integer qmin, observed in co NBI case, were absent in balanced NBI case.
There was one US to EU and one US to JA personnel exchange in this task. W.A. Houlberg of ORNL visited JET for the transport modeling of JET plasmas. He completed development of a set of equations for the bulk toroidal rotation and radial electric field in a multiple species axisymmetric toroidal plasma. These were summarized in a brief report, and implemented in the JETTO code for analysis of rotation in JET. P. Gohil of GA visited JT-60U to participate in QH-mode experiments (TP-5).
The main areas of activity have remained the same as last year with the proposed addition of a new area of focus. The key areas of focus are (i) confinement scaling with beta CDB-2, (ii) confinement scaling with collisionality in ELMy H-mode plasmas CDB-4, (iii) extension of global database studies to low aspect ratio CDB-6, (iv) ρ* scaling of confinement at ITER relevant dimensionless parameters at low and high beta CDB-8, (v) density peaking at low and high collisionality CDB-9. At the recent IPTA CDBM meeting held in Lausanne, Switzerland a new focus area was proposed to study the L-H power threshold at low density, CDB-10.
CDB-2 - Beta Scaling
Previous results have shown a range of global H-mode confinement scaling with beta from negligible on DIII-D and JET H-mode discharges to a strong degradation on JT-60U and AUG. Several candidate parameters which may play a key role in these differences are upper triangularity, fueling rates and toroidal rotation. This year AUG, DIII-D and JET performed additional experiments. The AUG experiments continued to show a degradation in confinement with beta. The DIII-D experiments were performed in two regimes, standard DIII-D shape with low toroidal rotation in hybrid discharges and in the AUG shape with higher rotation. The results are very preliminary but transport showed a stronger beta degradation in hybrid discharges than previously seen on DIII-D while the AUG results showed beta degradation as previously observed. Since the transport in hybrid plasmas is sensitive to rotational shear, imperfections in the Mach number match will complicate the detailed analysis to be done. JET also performed a beta scan in hybrid-like plasmas. Again analysis is at a preliminary stage. However, the confinement clearly shows a strong beta degradation in contrast to the previous JET studies.
CDB-4 - Scaling with Collisionality and Greenwald Fraction
No additional experiments were performed in this area this year. Prior JET and C-Mod experiments showed that collisionality rather than Greenwald fraction was the key scaling parameter. The NF paper describing these results is being updated following strong referee comments. An additional collisionality experiment on C-Mod awaits a cryopump to allow access to low collisionality.
CDB-6 - Aspect Ratio Scaling
An additional 100 NSTX observations in ELMy and ELM-free H-mode conditions were submitted to the confinement database. The confinement scaling with BT and Ip, τ ~ BT0.6-0.9 Ip0.4, is considerably different than in other higher aspect ratio devices. Trapped electron modes may be playing a more dominant role in the lower aspect ratio NSTX discharges. Further experiments on MAST are expected in 2007.
CDB-8 - Rho* Scaling at High and Low Beta
JET and C-Mod performed test shots for these experiments in late 2006. JET could match the ITER collisionality but the density in C-Mod was too high. Further experiments on C-Mod await a cryopump to allow lower, ITER-like collisionalities.
CDB-9 - Density Peaking Dependence on Collisionality
A database of AUG and JET discharges was assembled and used to determine that the density peaking factor could be fit with a dependence on the log of collisionality and linearly on the particle flux with the collisionality term playing the most important role. Projections to ITER suggest that the peaking factor, ratio of density at r/a=0.2 to volume average density, on ITER could be larger than 1.4. C-Mod data was added to the database and did not change the scaling when included in the regression analysis. Future contributions from TCV and JT-60U are expected which will add discharges with strong electron heating.
CDB-10 - L-H Threshold Power at Low Density
The threshold power has been seen to have a minimum value at low density in many tokamaks including AUG, C-Mod, DIII-D, JET and JT-60U but the physics and scaling of the minimum threshold and the density at which it occurs is not clear and thus projections to ITER are uncertain. The minimum power threshold has been identified as a potential issue for ITER and thus it was proposed to create another focus area for this work. J. Snipes will act as the spokesperson for this group. Future joint experiments are expected to focus on density scans to study the physics and parametric dependence of the minimum threshold and density at which it occurs.
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 significant progress in this area using balanced neutral beam experiments on DIII-D and JT-60U, to clarify the critical rotation required for RWM stability. The plasma rotation needed (~0.3% of Alfvén velocity) is observed to be generally lower than in previous joint experiments using magnetic braking. It is now under investigation whether the applied error field used in magnetic braking influences the RWM critical velocity. Joint experiments on JET with DIII-D participation have used resonant field amplification (RFA) as a routine tool to demonstrate advanced scenarios operating above the no-wall limit and have made an initial start on studying RFA from applied n=2 fields. Results on joint RWM experiments were reported at the 2006 IAEA Fusion Energy Conference.
Low β error fields; The one outstanding issue from the 2004/5 C-Mod, DIII-D and JET identity experiments was the apparent difference in the Bt scaling on C-Mod. Further experiments were conducted on C-Mod at 7.8T to try and resolve this issue. It was found that the higher field C-Mod data are consistent with the JET and DIII-D scaling (δberror/Bt~1/Bt) but the lowest field data (4.1T) still show a stronger inverse Bt scaling. These results are now being written-up as a journal article.
Sawtooth, NTM physics and error fields at high β; Joint experiments have been conducted on DIII-D to examine the effects of rotation on NTMs - these follow-on from experiments on JET substituting ICRH for NB power to reduce rotation. On DIII-D the co/counter-NB balance was varied. It was found that reduced rotation lowers the threshold for 2/1 NTMs (as observed in JET for 3/2 NTMs) and in addition it was found that error fields can combine with 'neoclassical drives' to brake plasmas and further lower NTM thresholds on both JET and DIII-D. These results will be presented at the 2007 EPS conference and written-up as a journal article. In addition, joint experiments (AUG, DIII-D and JET) have continued on the critical β below which NTMs are unconditionally stable, with new 2/1 NTM data obtained on JET. A new cross-machine analysis (AUG, DIII-D, JT-60U and JET) of NTM thresholds in the Hybrid Scenario is progressing. Cross-machine comparisons during β ramp-down (AUG, DIII-D, JET) and with ECCD stabilization (AUG, DIII-D, JT-60U) showed that the marginal island width of the 3/2 NTM is about twice the ion banana width (these results were presented at the 2006 IAEA Fusion Energy Conference). Analysis of experiments performed in AUG, demonstrating the enhancement of the stabilization efficiency of a neoclassical magnetic island by modulated electron cyclotron current drive, has been made. There has been a successful demonstration of destabilisation of fast particle stabilised sawteeth, above the marginal beta for 3/2 NTMs, in JET.
Disruptions; Experiments in DIII-D and C-Mod on disruption mitigation with noble gas injection have shown similar features: the neutral gas is stopped at the plasma edge, where the onset of low-order MHD instabilities followed by a global disruption-like internal reconnection then mixes the edge-deposited ions with the hot plasma core. This MHD mixing results in a radiative thermal collapse with a much slower time scale than the thermal quench encountered in a natural disruption. The slow thermal collapse and MHD mixing is consistent with NIMROD MHD code modelling. In both experiments, the amount of impurity gas assimilated into the plasma prior to the onset of the current quench is much less than required for collisional suppression of runaway electron avalanching. However, the ergodic magnetic field produced by the MHD modes may account for the low level of runaway production observed. Experiments in both machines also confirm that a mixture of hydrogen/helium with a small amount of higher-Z noble gas can improve the response time for onset of the radiative thermal collapse. These and other gas injection results were reported in a joint paper at the 2006 IAEA Fusion Energy Conference. More recent DIII-D results have shown the benefit of a fast gas delivery rise time for obtaining the most efficient impurity and/or hydrogen gas delivery prior to onset of the current quench. A new fast valve close to the plasma is being commissioned on AUG.
With regard to future plans from June 2007 to May 2008, it is expected that joint experiments on Disruption Mitigation, Neoclassical Tearing Modes, Resistive Wall Modes and Error Fields will continue, together with the related personnel exchanges.
Coordinated experimental activities/exchange of personnel took place for the following ITPA pedestal and edge topics.
PEP 1 & 3: JET/JT-60U pedestal identity experiments and modelling
New experiments in JT-60U (June 2006) with the JET identity shape have been carried out in June 2006. The scope was to study ELMy H-modes with JET-similar shape at reduced toroidal magnetic field ripple (compared to the ~1% at the separatrix of the previous experiments), at a reduced toroidal field of 2.2T, q95~3.6 (instead of 3.1T), for best compensation of the ripple. V. Parail (EJ44) participated in the experiments. The main result of this experiment was that a change in ELM size, which resulted in higher PELM/Psep, was clearly observed, while no significant improvement of pedestal performance has been observed. Based on these new plasma parameters, new similarity experiments were carried out in February 2007 in JET. The experiments included reference discharges without ripple, as well as ripple scans including a match to the JT-60U values (at the plasma midplane). N Oyama and H Urano (JAEA) participated in the experiments. Data analysis is in progress, and data were obtained on the effect of ripple on ELM and pedestal parameters, as well as on plasma response to fuelling, for comparison with existing JT-60U data. Note that the preparation for the JET TF ripple experiments benefited from the extensive use of the JAEA Orbit Following Monte Carlo (OFMC), used for the analysis of JET heat loads due to fast particle losses (by J Lonnroth on JAEA computers form JET).
PEP 6: AUG/MAST/NSTX pedestal structure and ELM comparison in double null
No new experiments have been conducted so far, because of the fly-wheel generator incident on AUG and the lack of available power on MAST. Initial experiments in 2007 on MAST during PEP-16 revealed a difficulty in accessing the H-mode in L-SN, δrsep<-15mm, even with PNB<3.4MW. This behaviour is reflected on NSTX and needs further investigation since, previously, the H-mode has been readily accessed at lower power in L-SN. More experiments on MAST are planned in June 2007. New experiments on AUG are planned for late 2007.
PEP 9: Dependence of the H-mode Pedestal Structure on Aspect Ratio (DIII-D/MAST/NSTX)
New data were obtained on MAST with higher NB power than before, allowing access to lower pedestal collisionality ~0.4 and higher pedestal local beta >7%. Stability analysis is commencing on these new data, as well on existing NSTX data to build on the results presented at the 2006 IAEA conference. The need for new data on DIII-D at matching pedestal collisionality is being assessed.
PEP 10: Collaborative experiments between MAST and ASDEX Upgrade on the effect of pedestal parameters on ELM radial extent
Experiments have been carried out on MAST in February and April 2007 and are planned for AUG for July 2007.
The MAST experiments were conducted at different current, toroidal field, ne and PNB, and the radial efflux due to ELMs was measured. During the February experiments, MAST was equipped with the AUG IR camera which, being identical to the MAST camera, allowed the simultaneous observation of the upper and lower strike points. In addition, the magnetic signature of the ELMs has been studied. To complete the study, the required data from AUG will be available from the end of July 2007. During these experiments the MAST fast visible camera will be installed on AUG to obtain images of the ELM filaments.
PEP 13: Comparison of small ELM regimes in JT-60U, ASDEX Upgrade and JET
New experiments carried out in 2007 at JET included a density scan with the identity plasma parameters and a ρ* scan, to define the operational space for the Type II ELM regime. Type II ELMs were obtained at a Greenwald fraction at the pedestal top between 0.7 and 0.8. ΔWELM/Wped was ~4 to 5%, while the minimum ν* required for the Type II onset is ~0.23, still far from the projected ITER value. Experiments to extend the regime to lower ρ* and ν* are planned for the next experimental campaign.
PEP 16: C-MOD/MAST/NSTX small ELM regime comparison
Following the successful shape development in 2006 and initial experiments on all three devices, further experiments were performed on MAST in May 2007 and NSTX in February/ April 2007. On both devices, β scans were performed in configurations close to DN (-6mm<δrsep<2mm). Good pedestal data were obtained for a wide range of βped and υ*ped. In both STs the small ELMs vanish at high input power corresponding to low collisionality and high β at the top of the LFS pedestal. On MAST, small ELMs are still observed at PNB=1MW, whereas on NSTX the small ELMs are only observed between 2MW<PNB<3MW. In LSN with δrsep=-15mm H-mode access was impossible on MAST and NSTX despite NB powers of PNB<3.4MW on MAST and PNB<6MW on NSTX. This behaviour is of particular interest, since H-mode access in LSN is generally no problem with PNB>2MW on NSTX with κ>1.9 and needs further investigation. The structural information obtained for the small ELMs close to DN on NSTX and MAST is distinctly different to those obtained for the Type V ELMs observed on NSTX. Further experiments in C-MOD are planned for the summer 2007. For this, a shape closer to DN needs to be developed to match the shape on NSTX and MAST.
D/T retention: Retention in gaps (DSOL-13) due to co-deposition with C and B is a special concern since such codeposits are hard to access for tritium recovery. Studies of inner divertor gaps in AUG showed that the inventory was predominantly on plasma-exposed areas, however, and that there was ~identical inventory with/without gaps: the inner divertor gaps do not increase inventory but change the spatial distribution. In the outer AUG divertor inventory was very small and there was no difference between plasma-exposed and plasma-shadowed sides. Regarding more remote areas in AUG: maximum inventories were in line-of-sight to strike points with only small inventories on shadowed areas, indicating that deposition is by particles with high (~1) sticking probability and there was negligible inventory in pump ducts. D deposition in castellated grooves in Be JET limiters (from the 1980s) is always associated with C. In Tore Supra the D retention rate decreased with increasing radiation fraction (variable impurity seeding). Deep penetration of D in CFCs varies significantly with CFC type: T depth profiles in JET (2D CFC) divertor tiles showed about 40% of the T was retained at depths > 1 mm, while only a few % was found at such depths in the TFTR 4D CFC tiles. Mechanism for deep retention in CFC (Garching): hydrocarbon molecules formed at end of range migrate through interconnected pores into bulk with codeposition on pore surfaces. ~50% of D injected into C-Mod is retained initially in Mo tiles (if no disruptions), decreasing to ~ 25% after a few shots, Retention does not appear to saturate for further discharges. Lab experiments with pristine Mo show much lower retention and saturation; however, new lab (DIONISOS) experiments appear better able to explain deep retention due to a mechanism dependent on plasma flux density, e.g. surface D pressure driven by large incidence fluxes (MIT hypothesis) or due to D self-aggregation in clusters due to stress field created by implanted D, creating traps (Garching). Atomic displacements cause by DIONISOS MeV ion beam simulates 14 MeV n-damage and enhances bulk retention of D in Mo, raising issue of dynamic response of Mo/W to n-damage and T retention. Neutron damage will create strong trapping in vacancies distributed over the entire W thickness, which may become accessible for T via diffusion in tension field (Garching).
Multi-code, multi-machine edge modelling and code benchmarking (DSOL-14).
A joint Pedestal and Divertor/SOL session was held on modeling. The session covered a wide range of modeling techniques, applied to a large number of machines. Topics included interpretive modeling, extensively using all available experimental data and making an 'empirical reconstruction' of the plasma by means of the OSM-EIRENE code. The insight and knowledge of plasma transport coefficients gained in such studies can then be leveraged for predictive simulations. A large amount of effort is also devoted to the more standard approach of predictive 2-D edge fluid plasma codes, often coupled to Monte-Carlo packages for neutrals. The issue of fuelling of the pedestal region (DSOL-16), in both DIII-D and AUG, was tackled with the UEDGE code package and in Alcator C-Mod by neutral kinetic calculations. Matching the large SOL flows seen in experiment remains a challenge for such codes and is only achieved in special cases. The difficulty current codes have at reproducing all the features of divertor detachment, in particular the particle flux rollover, were seen to be of concern for proper extrapolation of current models to the ITER scenario of semi-detached divertor operation. Detailed SOLPS modelling of well documented AUG plasmas reveals a tendency for code solutions to predict colder and denser plasma in the divertor (target Te profiles flatter than in expt, peak Te smaller). Simulated Er in the SOL (SOLPS, EDGE2D) << than in experiment. Codes underestimate of Er is consistent with their underestimate of experimental parallel SOL ion flows. Leading hypothesis: divertor, Er and flows discrepancies are all related to each other and caused by non-local effects of parallel electron transport in the SOL and divertor. It was pointed out by the ITER Team that modeling needs will be shifting from scenario development to diagnostic design support as the period for procurement of the hardware moves on from the divertor itself to the diagnostics. More generally, the feeling of the community was that priority in the modeling effort should be given to understanding and reproducing detachment physics (including its in/out divertor asymmetry), which are essential for ITER operation, followed by work on kinetic and transient effects such as ELMs, and reproduction of the experimental signatures of radial electric field and strong Mach flows in the SOL.
ELMs. (DSOL-1 & DSOL-19). On JET, ELM-filaments follow pre-ELM B-lines. Toroidal separation Δφ~2πn; for high triangularity, n~5-20 (upper dump plate), n~20-50 (outer limiter); for low triangularity, n~10-15 (outer limiter), similar to previous observations in AUG and MAST. Average normalized width δθ/Δθ~δφ/Δφ~ 0.6, expected to hold also in ITER. High Ip and low fueling produce a few large ELMs (ΔWELM <0.9 MJ) at ITER ν*ped. These cause post-ELM radiation spikes of ~50% of ΔWELM and increased impurity influxes into the bulk plasma. In DIII-D ELM suppression studies, a correlation is seen between width of vacuum island overlap region and ELM suppression: Δstocvac/Δpped > 3.5 gives ELM suppression. JET studies show that type-I ELMs can be mitigated by the application of a low n (1 or 2) external perturbation field.
ITER Session. Professor Karl Lackner chaired a session on ITER, that was particularly well attended and included presentations on the ITER Design Review, in progress. He lead a lively and productive discussion, concluding that there was a consensus that ITER should have the capability for an (at least once in a lifetime) exchange of the first wall material. There was a general recognition that at some stage ITER should test a fully DEMO/reactor relevant material combination for both first walls and divertors, which in all likelihood appears to be tungsten. It would be highly desirable if the present design of the first wall could be modified in a way to give more flexibility for this, including the introduction of protection limiters. The majority of the group does not feel that presently available experimental evidence would justify now the start of ITER operation with all-tungsten plasma facing components.
International collaborative experiments coordinated through IEA IAs have made significant progress and expanded multi-machine data sets for further analysis in view of steady state operation development. Concerning new data during the period of this review: JT-60U operated June, July and September, 2006. JET operated from November, 2006 until April, 2007. DIII-D operated from June, 2006 until September, 2006 and after a vacuum opening, operations resumed in February, 2007 and will continue until July, 2007.
On preparation of ITER steady-state scenario (SSO-1): In JT-60U, in real time q profile control experiments, it was shown that as qmin raised by off-axis LHCD above 2, which is a targeted steady-state profile, the m/n =2/1 NTM disappeared and βN retrieved. Full non-inductive current drive solely by the bootstrap current (fBS ~ 100%) had been pursued to document it's self-regulating feature. And ramp-up of Ip was observed in a bootstrap dominant discharge, that is bootstrap over-drive. In JET, in a discharge with q95 ~ 5 and δ >> 0.4, reaching βN > 4xli. βN >> 3 has been sustained for duration of order the resistive time with βN controlled using NBI power. The achievable βN increases with heating start time (or decreasing qmin) with βN being limited by large n=1 MHD. The no-wall limit has been probed using resonant field amplification. In DIII-D, experiments for exploring the performance boundaries in the steady-state scenario focused on the optimization of the plasma boundary shape. It was clearly demonstrated that changes in the shape of the outer boundary with fixed triangularity and elongation led to significant variation in the maximum stable beta for long-pulse operation. A clear optimum was observed, similar to model calculations prior to the experiment.
On preparation of ITER hybrid scenario (SSO-2): In JT-60U, sawteeth were induced by central ECCD and effect of the sawteeth on NTM was investigated in high βp discharges and it was found that central ECCD enhanced sawteeth but suppressed the m/n =3/2 NTM. In JET, qualifying the hybrid scenario for ITER concentrated on: (1) Extending operation to q95=3.2 and q95 = 2.7. (2) Systematic comparisons with the H-mode scenario (no apparent differences are observed on confinement yet). (3) Extending the duration at βN = 2.5 to 20 s. (4) Operation at higher density, up to the Greewald density with ~10% loss of confinement. (5) Employing ELM mitigation techniques (Type III ELMs) using radiative fraction control. The role of MHD in hybrid scenarios was studied: The confinement without 3/2 NTMs is close to H98(y,2)=1, while with a 3/2 mode the confinement is degraded by typically 15%. The hybrid scenario has been extended to βN = 3.6 (βN,TH = 2.6), above the no-wall limit and diagnosed by MHD spectroscopy. A βN scan has been carried out at same ρ* (and υ*) in hybrid discharges. In DIII-D, experiments examined the dependence of transport on rotation and beta and the role of MHD in the current profile evolution. The confinement increases significantly with increasing rotation. This increase is described well by gyro-fluid model simulations that include the effect of ExB shear. The beta scaling experiments were carried out and are under analysis. Fast ion transport was ruled out as a leading candidate to explain the anomalous current evolution in hybrid discharges through a series of experiments varying the mixture of co- and counter-injected NBs.
On real-time q-profile control in hybrid and steady state scenarios (SSO-3): In JT-60U, real-time q profile control by means of off-axis LHCD had been investigated. It was confirmed that qmin could be controlled well by this scheme. And as described above, effectiveness of the q profile control for optimization of steady-state operation was shown. Real time control of spatial gradient of the Ti profile (∇Ti) by combining on- and off- axis NB and real time measurement of Ti profile by filter CXRS. And real time control of ∇Ti was demonstrated. In JET, for preparation of future implementation on JET control system, response of plasmas to actuator modulation.
On documentation of the edge pedestal in advanced scenarios (SSO-4): In JT-60U and JET, related data had been collected in the related experiments. In DIII-D, experiments to measure the pedestal dependence on power, density, and rotation were carried out in hybrid discharges. The analysis is yet underway, but preliminary analysis showed that the pedestal clearly rose with increasing power input and was not limited at the same value by ELMs.
On Simulation and validation of ITER startup to achieve advanced scenarios (SSO-5, new in 2007): In JET showed that a much higher li at the end of the current rise phase than predicted by ITER simulations. In DIII-D, it was found that the outer wall startup with correlated shape evolution to maintain constant qlimiter during the current ramp result in much higher li than the present ITER simulations. Limiter heat loading and the influence of density are still under analysis.
On Ability to obtain and predict off-axis NBCD (SSO-6, new in 2007): In JT-60U, off-axis NB driven current profile, this indicates effectiveness of off-axis NBCD.
Concerning personnel exchange, there were many participation on JET advanced tokamak related experiments and/or preparation for them both from US, Dr.s T. Luce, J. Ferron, J. Menard and M. Murakami (UE315), and Japan, Dr. S. Ide (JE139). Dr. P. Gohil participated to JET real time control experiments from US (UE322).
As for the future plan from June 2007 to May 2008, the research on steady state operation will be continued. During that period, new data will be available from DIII-D, JET and JT-60U. The ITPA "SSO" TG is proposing researches on the ITER steady state relevant plasmas and hybrid operation relevant ones in more depth.
De-tritiation of plasma facing components at JET
Several activities have been launched in the EU related to in-situ detritiation.
Apart from the flash lamp technique, laser detritiation is the most highly developed technique and in 2006 a laser detritiation system has been used in the JET Beryllium Handling Facility (BeHF) to clean several tiles. The most important conclusions of this task are that all the co-deposited layers were removed in one laser pass and that almost all the products obtained after treatment were dust. The efficiency of the treatment was confirmed using IBA analysis which has shown the complete removal of the film as well as the fuel trapped in the co-deposited layer. No fuel diffusion was observed in the bulk material after laser detritiation. It was also shown on a mock-up of the inner divertor (Tiles 3 and 4) that the laser can access and treat remote surfaces. In 2008, this activity will continue and trials will be undertaken in order to optimize and control the laser detritiation. Moreover, a tool must be designed and developed in order to install this system on the JET Remote Handling boom to test the hardware in the JET Vacuum Vessel itself.
A small plasma torch was also developed during 2006 and tested on laboratory samples. This torch operates with active gases such as Nitrogen, and the plasma plume allows the sample surface temperature to be raised to more than 500C. From the first trials, it appears that the removal rate is much lower than that for laser detritiation. However, the chemical processes induced by the torch could allow treatment within castellations. In 2008, it is planned to test this new small plasma torch on castellated samples and samples with thick films in the JET BeHF in order to assess its efficiency with real JET samples.
Also, a new system has been developed in the past year. This system, called Inside Gap Plasma Generator (IGPG), could be mounted on a robot and used to clean castellations using an RF Argon plasma. First trials have shown that the plasma penetrates the void and gaps and its efficiency is under assessment. It is proposed to use this IGPG on real samples in the JET BeHF in the future; 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 parallel with these tests and assessments, tasks devoted to the in-situ physical and chemical characterization of plasma facing components (PFC) have been launched. In 2006-2007, the first results on JET of in-situ Laser Induced Breakdown Spectroscopy were obtained. These can be used to determine the chemical composition of the PFC. It seems possible to obtain the composition of the co-deposited layers including the concentration of T/D/H. However, to be as quantitative as possible, ex-situ calibrations are needed. It is also possible to use code calculations to extract from the optical spectra obtained during LIBS the concentration of the species in the layer. This modelling activity will be supported in the JET 2008 Workprogramme.
Surface analysis of plasma facing components from JET
Surface analysis of JET divertor tiles removed during the 2004-5 shutdown for 13C studies has been completed. 13CH4 was puffed in the outer divertor on the last day prior to the shutdown. Two different poloidal sets of tiles have been scanned, with toroidal variations being found only close to the injection points. Modelling work on the transport is continuing. A third 13C puffing experiment in JET was carried out prior to the 2007 shutdown. This time puffing was from the outer mid-plane into H-mode discharges. During the shutdown, a poloidal set of divertor tiles and selected tiles from the main chamber will be removed for analysis. This task will be scheduled in 2008.
A set of W-coated tiles have been exposed in JET during 2005-7 in order to check on predicted lifetimes for coatings during the planned ITER-like wall experiment. Areas of interest are the outer strike point for plasmas with high delta, and neutral beam shine-through and beam re-ionisation areas. The tiles have a 3µm W coating, similar to the ones exposed at the outer divertor during 2001-4, and will be removed for analysis during 2007-2008.
Analysis of mixed deposited materials has been flagged by EFDA as a major topic of concern for ITER. JET provides a source of Be-C mixed materials, and in future will provide Be-W and Be-W-C films. Existing analysis techniques such as IBA and SIMS will need to be supplemented by high resolution techniques such as SEM and Ion Micro-probe and chemical analysis techniques such as XPS/AES.
Marker layers have proven to be very useful in determining the extent of erosion/deposition in JET in recent years. The pattern of erosion/deposition is likely to be changed considerably with the introduction of the ITER-like wall, which will have plasma-sprayed 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 marker coatings will need to cope with the rough W surface, so analysis methods will require optimisation, whilst for the Be tiles, marker coatings of Be and onto Be substrates bring complications due to handling the toxic material. All developed marker systems will require high power flux testing to ensure they will survive in JET, and analysis of the samples before mounting in JET (and again after their eventual removal) is implicit in the development.
In the frame of the ITER-like wall project on JET (which is configured specifically to have direct relevance to ITER PFC), several activities devoted to the preparation of IR surface temperature measurements in a metal machine environment have been launched. Based on active IR measurements techniques, they are currently under assessment. In 2008, activity will be launched in order to select the most suitable technique for JET and the ITER tokamak. An assessment of practical methods of implementation in JET could be included in this task.
Surface analysis of plasma facing component from JT-60
In JT-60, erosion/deposition analyses of the plasma-facing wall have shown that local carbon transport to the in-board divertor was appreciable, in addition to long-range transport. The total deposition and erosion rates in the divertor region were ~10x1020 C atoms/s and ~-6x1020 C atoms/s, respectively. 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 6x1019 atoms/s, respectively. In the plasma-shadowed area beneath the divertor region at around 420K, 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 wall covered with boron layers. Their (H+D)/C and H+D retention were ~0.16 and ~10x1022 atoms/m2, respectively, for the vacuum vessel temperature of 570K. Integrating this retention over the whole main chamber wall results in a significant 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 had a peak 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 the carbon transport more closely, 13C deposition in the gap of the divertor and dome-wing tiles will be analyzed.
Tungsten has been selected for plasma facing materials such as the divertor baffles for fusion reactors. For the case in which the energy of impinging particles can be kept below the sputtering threshold, it has been believed that the plasma impurity problems can be avoided by using tungsten. However, blistering can occur at the tungsten surface, even if the ion energy is too low to create displacement damage such as vacancies. This tungsten blistering could lead to an instability in the plasma due to impurity release into the core plasma. 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.
In the framework of the Bi-lateral Agreement between the EU and US, new diagnostic systems, installed on JET during the last years, have been commissioned and/or exploited. First, a set of nine Faraday cups, located in 5 different poloidal positions in the lower outer part of JET have produced the first scientific results, showing a clear poloidal dependence of the fast particle losses. These detectors provided data also during the Toroidal Field Ripple experiments and the analysis is under way. Second, the JET Charge Exchange Recombination Spectroscopy (CXRS) system, to which ORNL and PPPL contributed with two high through-put, transmission grating spectrometers and two fast CCD cameras, has been fully commissioned and systematically exploited during the last Experimental Campaigns. Third, GA filter spectrometers and PPPL amplifiers for the High Resolution Thomson Scattering system on JET have worked properly during commissioning and operation of this new system.
One session during high level commissioning was devoted to Ohmic discharges for the calibration of JET MSE with a technique proposed by GA. The raw data are good but the analysis has not yet started.
Since the installation of new waveguides the S/N ratio of JET reflectometer system has improved by about 20dB. Therefore, a more coordinated collaboration has been activated with PPPL for the exploitation of the PPPL correlation reflectometer on JET (one band in the interval 100-105GHz) and for the cross-validation of these data with the other JET systems.
Negative Ion Neutral Beam Technology
Continuous progress is being made in the development of the next generation of negative ion based neutral beam systems. A key contributor to this work under the LTA agreement is Dr. Larry Grisham in collaboration with Japanese laboratories at NIFS and JAEA. Dr. Grisham made three trips to collaborate with the JT-60U negative ion beam group to engage primarily in the development of techniques to steer more reliably the many beamlets that comprise a beam, and to find ways to compress the beam envelope. This will be incorporated into the upgraded accelerators for JT-60SA, and will be essential for designing the ITER beam accelerators. The collaboration also studied processes precipitating breakdown in the accelerator column, and the evolution of oxygen contamination in the source plasma. The paper "Beamlet deflection due to beamlet-beamlet interaction in a large area multi-aperture negative ion source for JT-60U" by Kamada, Hanada, Grisham, Jiang was submitted to the Int. Conf. on Ion Sources (Korea, 2007), and also a paper to the Int. Conf on Vacuum Breakdown (Japan, 2006), "Correlation between voltage holding capability and light emission in a 500keV electrostatic accelerator utilized for fusion application" by Hanada, Ikeda, Kamada, Kikuchi, Komata, Mogaki, Umeda, Grisham, Kobayashi.
IEA Large Tokamak Cooperation
Workshop Number: W64
SUBJECT: IEA Large Tokamak IA Workshop on "Edge Transport in Fusion Plasmas"
Date: 11-13 September 2006
Place: Dom Polonii, Kraków, Poland
Name (s) of attendees: (All names of attendees are listed in the attachment.)
Brief description of the activities in the Workshop W64
The workshop was held in Krakow, Poland from the 11th to 13th of September 2006, in the cultural center "Dom Polonii". A detailed list of the 30 participants is provided in the appendix, along with agenda of the workshop. The first day (Monday) focused on edge turbulence with experimental results being reported in the morning session and theoretical work in the afternoon session. The second day (Tuesday) was similarly divided (morning - experiment, afternoon - experiment) with the topic being edge localized modes or ELMs. The final day (Wednesday) contained a morning session dedicated to challenges in edge modeling with fluid codes, and an afternoon session including a plenary discussion followed by smaller working group discussions. The outline of each session is presented below.
In the Monday morning session (chaired by W.Fundamenski), S.Zweben presented a comprehensive overview on recent observations of edge turbulence with fast cameras, pointing out that turbulent structures in the edge are a sincere issue for ITER. Besides results he also reviewed available diagnostic methods for the edge and the emerging comparison with detailed turbulence calculations. J.Horacek reported on results from TCV on detailed investigations using probes of edge and SOL properties with good comparison to ESEL modeling. M.Hron showed recent results from the Czech tokamak CASTOR including biasing experiments in the SOL leading to improved confinement. The stochasticity in magnetic field lines at the ergodic divertor at TEXTOR and its impact heta and particle deposition was the topic of a talk by M.Jakubowski, depicting the relevance of the length of the ergodic field lines for heat deposition and the question of stochastic transport vs. collisional transport. First results and the construction on the FTU liquid Lithium limiter including observations of high density peaking at reduced Zeff were presented by V. Pericoli-Ridolfini. P. Scarin reported on structures in the edge turbulence on RFX showing scalings in the structures and the spectra with density.
The Monday afternoon session (chaired by V.Naulin) dealt with theoretical progress made in edge turbulence simulations. B.Scott presented an overview of recent developments in edge turbulence codes and the requirements for energy transfer terms required for realistic numerical simulations. A.Kendl reported on the influence of magnetic field geometry on the turbulence, while T.Ribeiro presented first calculations showing the change of the turbulence characteristic from the drift wave type on closed magnetic field lines to interchange type on open field lines. T.S.Hahm showed the importance of turbulence spreading from the edge into the core and in general into stable regions. A.Nielsen presented ESEL simulations for JET showing fair agreement in predicting e-folding length and radial transport profiles, while stressing that corresponding ITER simulations will see a tenfold increase in computational resources.
In the Tuesday morning session (chaired by W.Fundamenski), K.Kamiya presented an overview of experimental observations of edge localized models (ELMs) in present day large tokamaks, including recent results on the effect of toroidal field ripple (varied with ferromagnetic insets) on ELM size on JT-60U. Peter Balan showed recent measurements for Reynolds stress and particle fluxes in the quiescent inter-ELM phasesand compared these with numerical simulations of 2D SOL interchange turbulence for ASDEX Upgrade parameters. K.H.Finken reported on the Influence of the Dynamic Ergodic Divertor (DED) in the Textor tokamak on ELM behaviour, including the suppression of ELM activity at even moderate DED currents. S.Zweben, reporting for R.J.Maqueda who could not attend for personal reasons, showed High Speed Images of Edge Plasmas in NSTX; the images revealed an apparent continuum in intermittent activity and phenomenology of filamentary structures from ohmic through L-mode to ELMy H-mode conditions. E.Nardon discussed the simulations of ELM control by resonant magnetic perturbations (RMP) for ITER, including the expected response of the plasma to the RMP as predicted by the non-ideal MHD code JOREK. The screening of the RMP due to toroidal rotation was suggested to explain the fall of the edge density with RMP in the DIII-D I-coil experiments. In the final talk, G.Kirnev reported on Investigation of nonlinear interactions of plasma fluctuations in the edge of the T-10 tokamak, showing a two wave decomposition of the observed fluctuatios into several pairs of the frequency components (with frequencies f1 and f2) obeying the rule f1+f2 = f0.
The Tuesday afternoon session (chaired by T.Rognlien) dealt with progress in the theory and simulations of ELMs. G. Huysmans presented an overview talk in which he described several approaches at modeling the ELM. Non-ideal MHD codes were shown to be fairly successful in modeling the early non-linear phase of the ELM evolution. X-point geometry and plasma dissipative effects were shown to be important in this context. O.E.Garcia presented dynamical simulations of isolated blobs and ELM filaments in SOL plasmas based on electrostatic interchange dynamics. He showed that It is demonstrated that such plasma filaments develop dipolar vorticity and electrostatic potential fields, resulting in rapid radial acceleration and formation of a steep front and a trailing wake. A.H.Kritz discussed the Modeling of ELM Dynamics in ITER. He addressed the important question of integrated modeling of ELMy H-mode discharges, namely how much plasma and current density is removed during each ELM crash. Non-ideal MHD simulations using the NIMROD code indicate the formation of ELM filament structures, which are observed in many existing tokamak experiments, in marginally stable ITER equilibria. Differences between ELM dynamics for ELMs triggered by ballooning instabilities and by peeling instabilities are described. M.Tokar discussed modeling of mechanisms responsible for ELM mitigation by external magnetic field perturbations. He demonstrated that through non-linear interaction, leading to the generation of side bands which suck energy from the main mode, such perturbations can raise the threshold of MHD instabilities. D.Kalupin presented simulations of the effect of pulsed gas puffing as a trigger for the ETB formation, using the 1-D transport code RITM. The results indicate that the turbulent transport is transiently suppressed by the shear of the radial electrical field, which emerges at the plasma edge due to the formation of the steep pressure gradient driven by the gas injection. Finally, P.Belo discussed the effect of the edge transport barrier on impurity transport using the 2D transport code EDGE2D. She found that the decrease in the deuterium puff level lead to an increase in the parallel classical thermal force, such that the sum of the parallel friction and thermal forces was directed away from the divertor region allowing impurities to concentrate in the plasma core.
In the Wednesday morning session (chaired by V.Naulin), was devoted to numerical approaches to edge modeling. R. Schneider presented an overview on current challenges in edge modeling pointing out the successes in edge modeling over the last decade, but also lasting complex questions that need addressing in an integrated way. He showed the scope of edge modeling from molecular dynamics simulations via kinetic and gyrokinetic calculations to the more ubiquitous fluid models. He also called for integrated modeling as is happening in other engineering and physics fields with well defined measures for quality assessment and code integration. C.S. Chang presented the SciDAC program on integrated simulation of edge transport in fusion plasmas and showed first results from the full f gyrokinetic XGC-1 PIC code. He called for an integrated approach to turbulence and neoclassic in the edge as conventional neoclassical theory often breaks down for the SOL. Coupling of the code to an MHD code is underway. T. Rognlien presented the progress on a continuum gyrokinetic code for edge plasmas and results from early physics benchmarks from modules of the code. . H. Kawashima reported on JT60U/JT-60SA simulations and development of a integrated SOL/divertor code in JAEA. An effort to integrate particle and fluid codes for the edge is underway. Results on sputtering and divertor pumping were presented. . Kalentiev presented results from different numerical approaches to 3D transport modeling of fusion devices, focusing on three competitive codes under development for W7-X . One problem being the mapping of a complex, evolving magnetic field structure to a usable and computational beneficial numerical grid. J. Brooks considered sputtered impurity edge plasma transport modeling for ITER parameters and the spatial distribution of sputtered material from the ITER wall. Large uncertainties in the material redeposition properties were reported.
The concluding discussion session on Wednesday afternoon was chaired by W. Fundamenski, V. Naulin and R. Zagorski. Topics of the previous days were taken up in a vivid discussion on the convergence of large scale MHD models towards fluid turbulence models. The need for better benchmarking of codes was stressed as the need for development and maintenance of larger code projects. It was agreed that some progress will be made on the inclusion of atomic physics into transport codes and the coupling of these to turbulence codes. The experimental side has benefited from better diagnostics in the edge, with a clear call for improved spatiotemporal diagnostics and quantification of fast camera data.
This report received by Secretariat on day mo y
|EU:||Balan Petru (Innsbruck University)
Belo Paula (CFN/IST Lisbon, Portugal)
Finken Karl Heinz (IPP, Forschungszentrum Juelich,Germany)
Fundamenski Wojtek (Euratom-UKAEA)
Garcia Odd Erik (Riso National Laboratory, Denmark)
Horacek Jan. (Institute of Plasma Physics, Czech Republic)
Hron Martin (Institute of Plasma Physics, Czech Republic)
Huysmans Guido (Association Euratom/CEA Cadarache, France)
Jakubowski Marcin (IPP, Forschungszentrum Juelich,Germany)
Kalentiev Oleksander (IPP Greifswald, Germany)
Kalupin Denis (IPP, Forschungszentrum Juelich,Germany)
Kendl Alexander (University of Innsbruck)
Nardon Eric (Association Euratom/CEA Cadarache, France)
Naulin Volker (Riso National Laboratory, Denmark)
Nielsen Anders Henry (Riso National Laboratory, Denmark)
Pericoli-Ridolfini Vincenzo (ENEA - C.R.FRASCATI, Italy)
Ribeiro Tiago (IPP - Garching, EURATOM Association, Germany)
Scarin Paolo (Consorzio RFX, Enea-Euratom, Italy)
Schneider Ralf (IPP Greifswald, Germany)
Scott Bruce (IPP - Garching, EURATOM Association, Germany)
Tokar Mikhail (IPP, Forschungszentrum Juelich,Germany)
Zagórski Roman (EURATOM-IPPLM, Warsaw, Poland)
|USA:||Brooks Jeffrey (Argonne National Laboratory,USA )
Chang Choong-Seoc (Courant Institute of Mathematical Sciences, New Jersey, USA)
Hahm Taik Soo (PPL, Princeton University, USA)
Rognlien Tom (Lawrence Livermore National Laboratory, USA)
Zweben Stewart (PPL, Princeton University)
|JAPAN:||Kamiya Kensaku (Japan Atomic Energy Agency)
Kawashima Hisato (Japan Atomic Energy Agency)
|RUSSIA:||Kirnev Gennady (RRC, Kurchatov Institute, Moscow, Russia)|
IEA Large Tokamak Cooperation
Workshop Number: W65
SUBJECT: Fifth Joint workshop of Large Tokamak , Poloidal Divertor and Textor IA's on "Implementation of the ITPA Coordinated Research Recommendations"
Date: Nov. 30 - Dec. 1, 2006
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 W65
The 5th annual Workshop for the planning of ITPA/IEA joint experiments to implement the ITPA high priority research tasks was held at JAEA Naka on November 30 and December 1. About 33 participants (see Attachment #1) included Executive Committee members of the three tokamak related IEA IAs, Chairs, Co-Chairs, and some members of the ITPA Coordinating Committee and its Topical Groups, and Leaders or their representatives of ~ 14 tokamak programs in the seven ITER Parties (CN, EU, IN, JA, KO, RF, and US). The previous workshops for the planning of Joint Experiments were held at MIT, USA (Nov 02), Naka, Japan (Dec. 03), and Oxford, UK (Dec. 04), General Atomics, USA (Nov. 05).
The meeting agenda is shown in Attachment 2. F. Romanelli (EFDA) made opening comments as the Chair of the IEA LT ExCo. P.Gohil (GA), representing T. Taylor (GA), reported briefly on the 4th Workshop held at GA, USA in 2005. Ron Stambaugh, Chair of the ITPA Coordinating Committee reported on the status of Joint Experiments that were planned at the last meeting. For preparation of the discussion on the joint experiments, the present status of the world tokamak devices were introduced by updating information on the new hardware capabilities, their operating schedules, and major topical areas of their research programs.
After the last meeting in 2005, 63 joint experiments were active. Among them, new results have been obtained in 34 experiments.
In this meeting, the seven ITPA Topical Groups proposed 11 new joint experiments and updated their proposed research on the experiments to be continued into 2007. Thus there were 65 joint proposals for CY2007 for the participants to review and plan for CY 2007. The TG Chairs or Co-Chairs provided one page summaries of these proposals, including the progress made previously, and their recommendations for tokamaks that should participate in these experiments to provide the parameter range for the experiments.
These presentations were then followed with comments by the tokamak leaders or their representatives considering their tokamak programs and schedules, and technical details and the necessary basis of proposed joint experiments.
The last session was dedicated to a brief review each proposal, receiving an expression of interest from the tokamak leaders to participate in the specific joint experiment, and to identify spoke-persons for the joint experiments and the names of the contact persons from the participating tokamaks. This activity led to the development of the Joint Experiments Plan for FY 2007, which is shown in Attachment 3. The commitments from some of the tokamaks are not yet firm (indicated in green), pending finalization of Research Forums of the individual programs and their process to finalize their experimental program plans. Most tokamaks (in red) were committed to the experiments. The tokamaks participating in the joint experiments, but completed their experimental runs earlier are shown in black.
The process of developing joint experiments has matured during these past four years, and most proposals were well defined and involved coordinated joint experiments. These are labeled as ready for Experiments (E). Some proposals, categorized as D, required additional discussions for joint experiments, partly because tokamak leaders expressed an interest to join the experiment at the meeting and their participation needed to be defined. Only a few proposals were labeled as P, indicating that this is an ongoing programmatic activity, such as collecting database, with minimal coordination among the participating tokamaks.
This report received by Secretariat on day mo y
|Attendees:||M. Kikuchi (JA)||: Member|
|Y. Kamada (JA)||: Member|
|Y. Koide (JA)||: Alternate|
|S. Ide (JA)||: Alternate|
|Y. Nakamura (JA)||: Expert|
|Y. Ueda (JA)||: Expert|
|K. Kamiya (JA)||: Secretary|
|E. Oktay (U.S.)||: Member|
|T. Taylor (US)||: Alternate|
|P. Gohil (US)||: Expert from PD|
|R. Nazikian (US)||: Expert|
|S. Clement-Lorenzo (EU)||: Member|
|M. Watkins (EU)||: Alternate|
|H. Zohm (EU)||: Expert from PD|
|J.-H. Han (KO)||: Expert from PD|
|R. Jha (IN)||: Expert from PD|
|B. Wan (CN)||: Expert from PD|
|L. Yan (CN)||: Expert from PD|
The Twenty second Executive Committee Meeting for the IEA Implementing Agreement on Cooperation among Large Tokamak Facilities was held at Naka-site in 21 - 22 May 2007.
The Committee elected Dr. M. Kikuchi as the chairman until the next meeting. (Dr. R. Stambaugh had been replaced by Dr. T. Taylor as a US alternative member, Dr. S. Ishida had been replaced by Dr. Y. Kamada as a JA member, and Dr. H. Kimura had been replaced by Drs. Y. Koide and S. Ide as a JA alternative member. 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, H. Zohm, E. Oktay, (T. Taylor, P. Gohil, and R. Nazikian), and M. Kikuchi. The status reports are attached as Appendix C. Also, the status and plans of CHINA (EAST and HL-2A), KOREA (KSTAR), and INDIA (SST-1) were presented by Drs. B. Wan, L. Yan, J.-H. Han, and R. Jha. The detailed presentations will be uploaded on the following Web-site;
The present status of the Task Activities in each device was reported; JET (M. Watkins), AUG (H. Zohm), JT-60U (Y. Kamada), DIII-D (T. Taylor), NSTX (R. Nazikian), and C-mod (P. Gohil). In addition, Reports on the 8 Task Areas were briefly introduced. 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.
Workshops and personnel assignments completed in the period of June 2006 - May 2007 are listed in Appendix E1. Two workshops on "Edge Transport in Fusion Plasmas" (W64), and "Fifth Joint workshop of Large Tokamak (W65), Poloidal Divertor and Textor IA's on "Implementation of the ITPA Coordinated Research Recommendations"" (W65) were carried out. The total number of personnel assignments completed in the period was 21. All 21 PAs were for review tours (less than 4 weeks) (see Appendix E2). Subjects are summarized as follows (see Appendix E3): 2 on Task1: Transport and ITB Physics (10%); 0 on Task 2: Confinement database and modeling (0%); 6 on Task 3: MHD, disruptions and control (28%); 2 on Task 4: Edge and pedestal physics (19%); 0 on Task 5: SOL and divertor physics; 5 on Task 6: Steady State Operation (24%); 0 on Task 7: Tritium and RH Technologies; and 4 on Task 8: Other (19%). The reports on the workshops (FORM C) and the short reports for review tours are attached as Appendices E4 and E5, respectively.
We should enhance the task activities, especially for Task 2, 5, and 7.
Proposed Workshops and Personnel Assignments for June 2007 - May 2008 are listed in Appendix F. These include three new Workshops (W66: 6th joint WS of LT (W66) PD and Textor IA's on "implementation of the ITPA Coordinated Research Recommendations", W67: Control of ELMs and RWM linked with US-Japan WS (K. Yamazaki and M. Okabayashi coordinators), and W68: Development of high βN scenarios for ITER (Combined WS with PD)). The Committee discussed these proposals and authorized their implementation. The Committee agreed equally that exchanges between EU and JA related to JT60-SA design activities could take place. Also, we agreed to consider organising under this Implementing Agreement the H-mode Workshop, if support was discontinued by IAEA, and the committee may want to consider if this Workshop should be included in our planning.
We considered the establishment of a single IEA-IA for multilateral exchanges enabling collaboration in experimental and theoretical tokamak research [1, 2]. This IA would consolidate the activities of the present LT, PD and TEXTOR IAs and could provide a legal framework for voluntary tokamak research activities for ITER and next-step facilities worldwide, such as currently defined by the ITPA. The structure of collaboration activities could be organized according to topics. It may be preferable to implement hardware exchanges, including diagnostics and related personnel assignments, under separate bilateral agreements between Parties.
 The present LT, PD and TEXTOR IAs expire before 2012.
 Collaboration with alternative concepts on tokamak-specific issues is welcome under this IA.
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 2007.
The next Executive Committee Meeting will be held in May, 2008 in GA (U.S.). It will be a joint meeting with PD IA.
J-1. Report from the IEA FPCC meeting:
1) ITER - IEA interactions and the future role of ITPA (S. Clement-Lorenzo)
2) Restructuring of Agreements (H. Zohm)
3) SSO WG matters (E. Oktay).
"The 36th meeting of the IEA Fusion power Coordinating Committee (FPCC) was held in Paris on February 27-28, 2007. This meeting was especially important because the ITER Director General Nominee Ikeda and the Assistant DDG for Science and Technology David Campbell accepted the FPCC invitation to participate in the meeting to exchange views on working relationship between ITER and the IEA FPCC. The FPCC represents the collaborative work of international fusion community through nine IEA Implementing Agreements (IAs) in three broad topical areas: Tokamak related (Large Tokamaks-LT, Poloidal Divertor-PD, and Plasma Wall Interactions-PWI), Technology and safety related (Fusion Materials-FM, Nuclear Fusion technology-NFTR, and Environment-Safety & Economics-ESE), and Alternate Concepts (Stellarators, Reversed Field Pinch-RFP, and Spherical Torus-ST). The FPCC aim is how best to structure its activities in order to better serve fusion, and specifically ITER and the Broader Approach, in the changing world fusion program.
Ikeda and Campbell made presentations on ITER status and the ITER Scientific Program and they witnessed the workings of the FPCC through the annual reports from the Executive Committee Chairs of implementing agreements. The meeting was very successful in providing the opportunity for discussion of collaborative fusion research in the international fusion community and its continuing contributions to the ITER scientific program. These discussions provide the foundations for a close scientific relationship between ITER and the FPCC.
In addition to the presence of the ITER representatives, FPCC for the first time have invited representatives from China and India. Prof. Kaw from the Indian Institute for Plasma research attended the meeting and gave the presentation on the Indian fusion program. Representatives from China, Prof. Li and Prof. Pan, were not able to attend this time due to schedule conflicts and they expressed high interest in joining the FPCC and the Implementing Agreements.
The meeting was held in three, half-day sessions. The annual reports by the Executive Committee Chairs of the Implementing Agreements (IAs) were presented during the first session; the presentations on the ITER Status, ITER Scientific Program, FPCC Implementing Agreements, and the Broader Approach (BA) during the second session, and the reports on related fusion work in IAEA, Nuclear Energy Agency, and on other FPCC activities such as preparations for the IEA Technology Fair during the third session".
The participants of the Naka meeting had extensive productive exchange of views on the first two topics. These are indeed complex issues that have to take into account the interests and decision making in other forums such as the ITER enterprise (international ITER Organization, ITER Council, ITER Domestic Agencies, and research programs in ITER Members), ITPA Coordinating Committee, and FPCC.. These discussions and exchange of views will continue through the coming weeks and months through e-mail exchanges as the ITPA - ITER interactions are discussed at other forums. The issue of the restructuring IEA implementing agreements should be resolved in the next 3-4 years as the renewal of the IEA LT/PD agreements will be considered in 2011.
J-2. Discussion on restructuring and enhancement of IEA PD & LT activities and Participation of India and China in IEA PD
Following the report from the FPCC meeting, a discussion was held on possibilities to restructure the IEA-IAs. The role of the IAs as a vehicle to enable the co-prdinated experiments proposed by ITPA was felt to be very important. As an option, a merging of the three tokamak-related IAs (LT, PD, TEXTOR) was discussed. It was pointed out that the time scale for such an action would be until around 2010, when the present IAs expire. This was also felt to be an appropriate time scale to learn how the interaction between ITER and the various international bodies evolves, an information that will influence the decision how to go ahead with the IAs. It was proposed to assess commonalities / differences between the three IAs to guide such a decision. Also, it may be interesting to find out from IEA-IAs in other areas what they assess to be the added value of their respective IAs. Historically, the question of merging the tokamak IAs had been discussed in 2001 after a specific request from CERT, and the answer had been to increase the interaction between the IAs, but keep them seperated.
Several options were discussed, one of them being to use one large IEA-IA for coordinated (ITPA-type) experiments and bilateral ones for other actions between two parties, like work on diagnostics or exchange of equipment. It was pointed out that such bilateral actions are very important and there must be a framework for them as well. This would need more bilateral agreements, but could be achievable on the 2010 timescale (action item: make a list of existing agreements). It was pointed out that Russia is at present not included in the three IAs and it was felt that the contribution of the Russian programme would be important to have. Also, representatives from LHD had pointed out that the stellarator community is interested in keeping in touch with the tokamak progress. It will have to be discussed how these two items can be resolved.
One worry about a single large IA is the fact that this may mean an excessive list of reports from the individual machines. This could be avoided if the work in the IA was structured according to physics topics rather than single machine contributions, with reports given according to these topics.
Another boundary condition will be the change of the role of IAEA in fusion, such as heavily reducing the number of TMs. The issue was raised if the IEA-IAs should take up the role of hosting these TMs. At this occasion, a list of TMs should be compiled and a discussion is needed if all of them should be continued.
It was felt by the participants that this initial discussion was very helpful in raising many points, but no definitive conclusion was achieved yet (and is also not needed now). A very helpful input to further discussion would be a list of pros and cons of the present structure versus a large single IA (action item: compile such a list, someone to take the lead!).
J-3. Introduction of ITER Design Review
"ITER Design Review" was briefly introduced by Dr. Y. Kamada (JAEA), including 1) An ITER Research, 2) Sensitivity studies, 3) Identify ripple requirements for the attainment of the physics objectives, 4) Review PID requirements for disruption, VDEs and runaways, 5) Choice of PFC, 6) Requirements for tritium breeding blanket in phase II, 7) Develop startup scenarios, 8) Assure that space for RWM and ELM control windings, 9) Impact of increased gas loads due to pellet pacing and disruption mitigation on pumping systems affecting building size, 10) Ensuring the reliability of ITER systems - engineering practices, configuration control, and QA to support the procurements, 11) Assessment of the Maintenance Requirements inside the Cryostat, 12) Auxiliary Heating and Current Drive, 13) Wall conditioning, 14) Measurements, There are 383 "on-going" issues, and the total number of issues is 502 (Present Status; Design Review CM, 23-24 April 2007, Cadarache - P. Thomas, ITPA PEP-TG meeting, V. Chuyanov, May 7-9, IPP).
J-4. Issues of W- first wall material
Prof. Y. Ueda (Osaka-Univ., Japan) made a presentation on "Issues of W-first wall material", referring the presentation of "Irradiation Effects of He on Plasma Facing Metals" of Prof. N. Yoshida (Kyushu-Univ., Japan).
Discrepancy between the laboratory and tokamak experiments (H. Zohm, AUG) was pointed out. Therefore, more comprehensive understanding is necessary. There was a discussion on the choice of W- first wall material as ITER first wall material. The EU outlined its 'step-ladder' strategy to use W in AUG and then W and Be in JET to prepare the use of W-first wall material in ITER since the W- first wall material is at present the sole candidate for DEMO though there remain a lot of unsolved issues. Japanese side (M. Kikuchi) questioned about the use of W without sufficient confidence.
K. Review of rule concerning providing with program and database in research cooperation on JT-60U
"Review of rule concerning providing with program and database in research cooperation on JT-60U" (Appendix K1) was briefly introduced by Dr. M. Kikuchi, including, a copy of which, in addition to "Guidelines for International Collaborations on JET" (Appendix K2), is attached to these minutes.