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.
The JT-60/60U facility, which has been a prominent contributor to the IEA LT activities over the past two decades, was terminated in August 2008 to prepare for the construction and installation of the new superconducting tokamak JT-60SA which is expected to be operational in about eight years.
Current scientific foci of large tokamak experiments are: ITER baseline discharge simulation (start-up, flat-top, landing); candidate steady state scenarios for ITER and DEMO (long-duration sustainment of high plasma pressure, high bootstrap current discharges); qualification of hybrid mode for ITER; cross-machine experiments on plasma edge, TF ripple, Internal Transport Barriers; control of disruptions using massive gas injection; control of edge localised modes by perturbing magnetic configuration and using pellet injection; characterisation of plasma instabilities (resistive wall modes, neoclassical tearing modes); material erosion, migration re-deposition and fuel retention; effect of plasma rotation on confinement and MHD.
The objective of these investigations is to advance the scientific basis for the burning plasmas in tokamaks and contribute to the resolution of the issues identified in the ITER Research Programme accompanying construction 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 modelling 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), operating under the auspices of ITER, identifies high priority research tasks for ITER in close coordination with the ITER Organization, and proposes experiments and modelling activities to resolve them. The IEA LT IA holds annual workshops, in close cooperation with the IEA Poloidal Divertor (PD) and IEA Plasma Wall Interaction in TEXTOR (PWIT) IA”Ēs, the tokamak leaders and the ITPA on "Implementation of the ITPA coordinated research recommendations". The 2008 annual workshop was held at MIT, Cambridge, MA, US on 11-13 December 2008. In this 7th 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 preparations for the installation of the ITER-like Wall materials in JET in late 2009/10), as well as diagnostics improvements. In general, it is considered that the interactions between IEA/ITPA/ITER work well, with the primary path for the proposal of experiments being the ITPA Topic Groups.
In the EU, JET has completed successfully seven Experimental Campaigns (April 2008 to April 2009; 224 days of two-shift operation, comprising 151 days for scientific experiments and 73 days for commissioning the ITER-like ICRH antenna). There was a strong ITER focus (High level commissioning of new systems; issues that could impact on the design of ITER components; preparation of integrated operating scenarios for ITER; physics issues essential to the efficient exploitation of ITER). 51 ITPA ITER high priority coordinated experiments (45% of overall run-time) included cross-machine studies with JT-60U, DIII-D and ASDEX Upgrade. The strong involvement of the EU (760 visits; 319 people; 21 countries; 65ppy total) was complemented by strong International collaborations (US, Russian Federation, Japan, Korea; nearly 5ppy). An intervention for plasma control upgrades will be followed by experiments in June-October 2009. Major R&D activities and procurements took place for enhancements of high scientific value and strategic importance (eg. ITER-like combination of first wall materials (tungsten divertor and beryllium wall); NB Power Upgrade (34MW for 20s)) which will be installed from end-2009. Feasibility studies for ELM control coils and ECRH (~10MW) were established, with US and RF involvement. From early 2011, the focus will be on preparation of ITER operational scenarios at high power with acceptable plasma/wall interactions. 132 JET FT tasks have been launched since 2000 (~21M€; ~2.7M€ in 2009), concentrating on tritium in tokamaks, tritium processing and waste management, plasma facing components, engineering, and neutronics and safety.
On 29 August 2008 JT-60U experiments were completed successfully, fulfilling its core objective of developing an advanced tokamak concept with high integrated performance for ITER and DEMO. Towards this goal, significant progress was made during 2008, including long sustainment of high ¦ĀN plasmas exceeding the ¦Ā-limit of free boundary plasma with no conducting wall. International and domestic collaborations were extremely effective, with many researchers (19 from foreign institutes and more than 50 from Japanese universities and institutes) participating during the last two months of JT-60U operation. The activities and structure of JT-60 team then shifted substantially towards the construction of the superconducting tokamak JT-60 Super Advanced (SA) as a part of the Broader Approach (BA) Agreement between Japan and EU. JT-60SA is at the core of the Japanese domestic programme for tokamak development, contributing to ITER and the technical preparations for DEMO construction decision. Component procurements for JT-60SA are shared by the Implementing Agencies (JAEA and F4E). Of those launched in 2007 and 2008 (supply of PF magnet conductor, a coil manufacture building, vacuum vessel and materials for in-vessel components), the coil manufacture building and PF conductor manufacturing line have been completed on schedule. First plasma is planned for the end of FY2015.
Under the new U.S. Administration, significant changes in the scientific leadership of the DoE occurred during the last year with Dr. Stephen Chu, a Nobel Laureate, becoming the new Secretary of Energy and Dr. Ed Synakowski being appointed the new Associate Director for Fusion Energy Sciences (FES). ITER continues to be the major focus, with the US providing scientific support for the ITER design and R&D needs, and participating fully in the International Tokamak Physics Activity (ITPA). The three facilities (DIII-D, C-MOD, and NSTX) plan to operate for about 10-15 weeks in FY2009. The Fusion Simulation Program (FSP), which is a computational initiative led by FES with collaborative support from the Office of Advanced Scientific Computing Research (ASCR) is in its planning stage. It is aimed at the development of a world-leading, experimentally validated, predictive simulation capability for fusion plasmas in the regimes and geometries relevant for practical fusion energy. Following the Fusion Energy Sciences Advisory Committee (FESAC) report (October 2007) on "Priorities, Gaps, and Opportunities: Towards a Long-Range Strategic Plan for Magnetic Fusion Energy", the National Research Council of the Academy of Sciences published its review "Plan for U.S. Fusion Community Participation in the ITER Program (July 2008)". A Long-Range (15-20 year) Strategic Overview Plan, including the mission, vision, strategic goals and research activities to guide the FES program, is being developed for transmission to Congress in March 2009. A community workshop (Research Needs Workshop-ReNeW) is scheduled for 7-13 June 2009 to provide input to this plan.
The physics-related work in the collaboration is conducted under eight Task areas, seven of which cover the Topic areas used in the ITPA. These are Transport and ITB Physics, Confinement database and modelling, MHD, Edge and pedestal physics, SOL and divertor physics, Steady State Operation and Other issues (including Diagnostics, and also Power Supplies). In addition, Tritium and Remote Handling Technologies are conducted in Task Area 7. Accomplishments in these Task Areas are described in Attachment A2.
The joint Ex-Co meeting of the LT and PD IA”Ēs on 21-22 May 2009 provided the opportunity to discuss the future strategy of IEA IA summarised briefly below (with further details available in the minutes of the meeting):
The joint Ex-Co meeting of the LT and PD IA”Ēs also provided the opportunity to discuss several issues relating to tokamak activities. These are briefly summarised below (with further details available in the minutes of the meeting):
The close coupling between the ITPA, the ITER organisation, the IEA FPCC, and the IAEA IFRC provide the opportunity to streamline international collaborations in fusion, with its priority for the success of ITER in achieving its key scientific and technological objectives. In recognition of the change of the world fusion programme into this new era, symbolised by the establishment of the ITER Organisation, collaborations inside/outside IEA have to be strengthened in view of support and supplement ITER towards DEMO. The IEA LT homepage (http://www-jt60.naka.jaea.go.jp/lt/) is open to all IEA IA's and the public.
The IEA Large Tokamak Implementing Agreement remains one of strongest fusion IA”Ēs 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”Ēs. With the accession of the Government of Korea as a Contracting Party, the extension of the term of the Agreement by another five years, and the amendment of the Agreement - including a change of title to "Implementing Agreement for Co-operation on Tokamak Programmes (CTP)" - tokamak-related activities in FPCC will be streamlined. Productive interactions with ITPA, IO and the IFRC will be further enhanced if other fusion countries such as China, the Russian Federation, and India join the CTP Agreement in the future in order to facilitate science and technology exchanges among the domestic programmes of all ITER Members.
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.
Experiments on JT-60U were completed on 29 August 2008. A core objective of JT-60U experiments was to develop an advanced tokamak concept with high integrated performance for ITER and DEMO. Towards this goal, significant progress has been made on several issues in 2008 experiments. One of the most important challenges towards DEMO is long sustainment of high beta plasma exceeding the beta limit of free boundary plasma with no conducting wall. In this experimental campaign, it was found that decrease in toroidal plasma rotation velocity could bring about growth of Resistive Wall Mode (RWM) and then plasma disruption. It was also found that a newly found instability, Energetic particle driven Wall Mode (EWM), induced by perpendicular neutral beam injection could also trigger RWM. Based on these findings, operation scenario to drive plasma rotation in the direction of the plasma current has been developed by using tangential NB injection while keeping perpendicular NB power as low as possible. As a result, βN ~ 3.0, which exceeds no-wall beta-limit, was successfully sustained for ~5 sec, which is significantly longer than resistive current diffusion time(>3τR).
As for the development of steady-state operation scenario for the DEMO, we have been investigating Reversed Shear (RS) plasma with high confinement and high bootstrap current fraction (fBS). Since one of the issues in RS plasma is relatively low MHD stability limit, we tried to optimize the discharge keeping plasma boundary close to the conducting wall (rwall/rplasma~1.3) for stabilization of RWM. We successfully achieved the normalized beta of ~2.7, exceeding the no-wall beta limit (~2); operation region in RS has been significantly expanded to the region expected in future fusion power plants where fBS ~ 70-90% and the safety factor (q) ~5. In addition, the achieved plasma shows the high integrated performance which almost satisfies requirements in ITER Steady-State operation scenario; key parameters, such as the confinement improvement, the density normalized by Greenwald density, the fraction of radiation loss power and the fraction of non-inductively driven current, have the values to be required in ITER Steady-State operation scenario.
Furthermore, significant progress has been made on many issues, such feedback stabilization of neoclassical tearing mode, study on impact of plasma rotation, momentum transport, assessment of toroidal field ripple in ITER, scaling of edge pedestal width, development of supersonic molecular beam injection (SMBI), edge density and current density measurement with Li beam, improvement for NNB operation, fast power modulation of ECH, etc.. The SMBI system installed in JT-60U in collaboration with CEA-Cadarache under Large Tokamak IA was successfully operated after the completion of the test at Cadarache for improvement of the seal material inside the injector head. The successive density jumps associated with the SMBI pulses were observed, indicating high capability of the SMBI for the plasma control even in large tokamak devices. In achieving above-mentioned results, international and domestic collaborations were extremely effective, and lots of researchers, 19 from foreign institutes and more than 50 from Japanese universities and institutes, participated during the last two months of JT-60U experiments.
After the shutdown of JT-60U, the activities and the structure of JT-60 team have been substantially shifted towards modification to the superconducting device, JT-60SA, while the team also continues physics studies and plasma evaluations for JT-60SA, ITER and Demo based on existing JT-60U data. Objective of JT-60SA programme, which is promoted as one of joint programmes under Broader Approach (BA) agreement between Japan and EU and also a domestic core programme of tokamak development in Japan, is to contribute to ITER Project and also to technical preparations for the decision of DEMO construction, with enhanced performance in duration of plasma discharge, plasma shaping control, heat exhaust and particle control, stability control, and heating & current drive capabilities. According to the Integrated Design Report (IDR) approved in the 4th BA Steering Committee (December 2008), the first plasma is planned in the end of FY2015. Procurements of JT-60SA components are shared by Japan and EU. Existing infra-structure and equipments, such as heating & current drive systems, cooling facilities, power supplies, diagnostics etc., will be utilized as many as possible. The Integrated Project Team, consisting of the Project Team, JA Home Team and EU Home Team, was organized and is conducting the modification to JT-60SA. In accordance with provisional Work Programme 2007-2008, the procurement arrangements were launched between the Implementing Agencies, JAEA and F4E, for the supply of PF magnet conductor, a coil manufacture building, vacuum vessel and materials for in-vessel components in 2007 and 2008. The coil manufacture building and PF conductor manufacturing line have been completed in schedule.
The scientific leadership of DOE has changed with the new administration. The new leadership includes:
The U.S. Fusion Energy Sciences (FES) program status and significant program activities since the last IFRC meeting in October 2008 are as follows:
The last 12 months on JET have been a period of intense operations with the successful completion of the C20-C26 campaigns (9th April 08 to 7th April 09) comprising a total of 224 S/T days in two shift operation. Of those, 151 days were devoted to the scientific programme and 73 days to the commissioning of the new ITER-like ICRH antenna. The machine performance was very good, with reliable operation of the NB injection systems at high power. Other new hardware systems have been tested and exploited during the campaigns, including the high frequency pellet injector, the disruption mitigation valve and two TAE antenna arrays, as well as multiple enhanced diagnostics. In addition, major R&D activities and procurements related to components and buildings for enhancements to be installed on JET during 2009/10 (ITER-like Wall, neutral beam power upgrade, ...) have been undertaken and significant Fusion Technology Tasks conducted.
The scientific programme has 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). The development of ELM-resilient RF heating systems (3 dB Couplers and External Conjugate T for the A2 antennas; Internal Conjugate T for the ITER-like Antenna) has been a success with up to 8.5 MW RF coupled power achieved in ELMy H-mode. LH coupling was demonstrated at ITER-relevant power densities (24.1 MW/m2, scaled for 3.7 GHz). Cross-machine studies with JT-60U, DIII-D and ASDEX Upgrade addressed pedestal, toroidal field ripple and ITB comparisons. The influence of plasma shape and the use of heating during the current rise phase for ITER scenarios has been explored. Vessel forces and heat loads during disruptions were measured with the new Halo current sensors and wide angle IR camera. Different gas mixtures were tested with the disruption mitigation valve. In ITER baseline scenario studies, emphasis was given to the high current programme development, with plasma pulses operated up to 4.3MA at low triangularity. After significant effort, JET hybrid scenarios had a major breakthrough, achieving H98=1.3-1.4 at βN ~3. Candidate scenarios for ITER SS operation at q95~5 were explored. A number of studies was carried out in preparation for the ITER-like wall: the power footprints of regular and large (> 0.6 MJ) ELMs were characterised with the newly installed divertor IR camera. Material migration and fuel retention were quantified in reference pulses to be later repeated with the ITER-like wall. Different methods for ELM amelioration (EFCCs, vertical kicks, pellets) were explored and compared in dedicated experiments. From the physics side, the ion temperature critical gradient length "threshold" for anomalous transport was determined and inward momentum pinch was measured.
Once more, the campaigns were characterised by a strong involvement from the European Associations (760 visits by 319 people from 21 countries; 65 ppy total; 47 working days per person on average) which was complemented by strong International (non-EU) collaborations (with the US, the Russian Federation and Japan, equivalent to 4.5ppy or 6 percent of staffing in 2008). A total of 51 ITPA ITER high priority coordinated experiments was targeted, requiring 45 percent of overall run-time.
After a ten week intervention to carry out plasma control system upgrades for higher resilience against ELMs, an Experimental Campaign (C27) is foreseen for the period June-October 2009. 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 end-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 early 2011, the experimental programme will focus on preparation of ITER operational scenarios at high power with acceptable plasma/wall interactions and optimization 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. In addition, feasibility studies for ELM control coils and ECRH on JET (~10MW) were established, with US and RF involvement.
Since 2000, 132 JET FT tasks have been launched (total resources ~21M€ (~2.7M€ in 2009)), 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. The key areas of focus are (i) transport dependence on Ti/Te ratio in hybrid and steady-state scenario TP-3.1, (ii) QH/QDB plasma studies TP-5, (iii) scaling of spontaneous plasma rotation TP-6.1, (iv) measure ITG/TEM line splitting and compare to codes TP-7, (v) ITB similarity experiment between JT-60U and JET TP-8.3.
TP-3.1: Determine transport dependence on Ti/Te ratio in hybrid and steady-state scenario plasmas
In JT-60U weak shear plasmas (target for ITER hybrid scenario), the Ti-ITB degraded significantly when stiffness feature was strong in the Te profile against ECH, resulting in increase of Te/Ti. On the other hand, Ti-ITB unchanged or even grew, when stiffness feature was weak in the Te profile. Density fluctuation level in the frequency range of ITG seemed to be unchanged during ECH, however, correlation length became longer in the Ti-ITB degradation case and shorter in the Ti-ITB unchanging case. In DIII-D, analysis continued of the EC heated hybrid discharges obtained in 2007. The rise in Te with ECH can be reproduced with the TGLF transport code, which predicts that the electron transport in the EC heated case is dominated by higher-k fluctuation wave numbers. Experimental turbulence measurements of intermediate wave number, TEM-range fluctuations using Doppler backscattering show an increase with ECH, while low wave number ITG-range fluctuations also increase. Significant effects of ECH on fluctuations, resulting in confinement and ITB degradation, were observed in hybrid scenario plasmas. Systematic understanding of physical mechanisms responsible for change in fluctuations during ECH is required.
TP-5: QH/QDB plasma studies
Experiments were carried out in both DIII-D and JT-60U to study QH-mode operation. Of particular importance was to develop an operating regime with co-rotation in which the QH-mode could be achieved. While this was done in both devices, the QH-mode with co-rotation appears not to be robust, but, rather, its existence depends critically on just the right plasma configuration for optimizing edge stability. The edge particle transport in QH-mode plasmas can be controlled by changing the edge rotation. QH-mode periods up to 1 s long in have now been seen in plasmas with strong co-rotation in which there is strong toroidal rotation shear at the plasma edge. This is consistent with a model in which the EHO is an edge kink-peeling mode that is destabilized by shear in the edge toroidal rotation at an edge current density slightly below that on the ELM boundary calculated with zero rotation.
TP-6.1: Scaling of spontaneous plasma rotation with no external momentum input
The scalar database has been expanded significantly. Most of the data required for the initial empirical scaling studies have been acquired, though some more data are requested for 2009, particularly at higher βN, from DIII-D with balanced beams and from JET with high power ICRF. Data from NSTX and EAST are solicited. Emphasis for the future will be for full velocity profiles and to continuing and expanding the similarity experiments.
TP-7: Measure ITG/TEM line splitting and compare to codes
Experiments on AUG from previous campaigns show evidence for a transition from ITG to TEM turbulence as the density/ν* is varied. Experiments in DIII-D in the 2007 campaign managed to obtain large variations of the logarithmic electron temperature gradient, but no clear related modifications of the fluctuation spectra were observed. In previous campaign T-10 and TEXTOR observed two separate maxima in the spectra due to the simultaneous existence of two quasi-coherent modes. Experiments in the 2008 campaign in DIII-D have explored the conditions of dominant electron and ion heating and found that no obvious signature of a shift in the turbulence frequency could be observed because the predicted mode frequency is more than one order of magnitude smaller than the frequency shift due to the ExB rotation. In other DIII-D experiments applying the ECH swing technique, a modulation of the frequency and amplitude in density fluctuations has been observed in response to a local modulation of the logarithmic electron temperature gradient.
TP-8.3: JT-60U/JET ITB similarity experiment
Similarity experiments have been performed in both JT-60U and JET, focusing the triggering and sustainment of ITBs, with near identical configuration, heating waveforms, and normalized quantities. On JT-60U experiment, the reversed shear plasmas with qmin~3, 2 and the optimized shear plasmas with q0~2 were performed by changing the timing of high power NB heating. In torque scan, the identical temperature and density profiles while different toroidal rotation profile were obtained with identical power deposition and q profiles. On JET experiment, the reversed shear plasmas with qmin~3, 2 and the optimized shear with q0~2 were also performed. The TF ripple was varied at 0.08% (JET standard), 0.3% (similar to JT-60U) and 0.75% (higher ripple) in reversed shear scenarios. Many similar results on the triggering and sustainment of ITBs were obtained in 2008 and are currently being analyzed in detail.
As a result of the reorganization of ITPA activities, the groups on Confinement Database and Modeling (CDB) and Transport Physics (TP) were merged. The combined Transport and Confinement (TC) group had its first meeting in October, 2008. Many CDB tasks were merged with TP tasks or closed out at this time; the remaining tasks were renumbered as TC tasks. The LTA report on CDB consists of the final description of reports on ITPA led activities in this area; in the future, reports on transport activities will be found in the LTA report on TC. Below are the major ITPA elements associated with the CDB task agreement, spanning the period up to October 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 β dependence to a strong β degradation. The difference in the confinement scaling as a function of β appears to be associated with shape through edge (pedestal) stability. Theory and modeling show that core transport should be ~β-independent for experiments, but different pedestal behavior could be accounted for by peeling-ballooning mode theory. Since new experiments are planned on JET, AUG, MAST and NSTX, this task will continue in the new TC group.
CDB-4 Scaling with Collisionality and Greenwald Fraction
Prior experiments on JET, DIII-D, and C-Mod showed that collisionality rather than Greenwald density fraction was the key scaling parameter for transport. Recent experiments on JET also showed that the increase in normalized confinement as collisionality is reduced appears to saturate at low collisionality. Further progress on this topic will need to wait until C-Mod is able to achieve the needed normalized beta for the collisionality scan in the required shape. Therefore it was decided to close out this task.
CDB-6 Aspect Ratio Scaling
Analyzing the MAST H-mode energy confinement database in DND configuration using both statistical method on global variables and by transport analysis of single parameter scans confirm weaker then linear plasma current dependence and stronger then linear toroidal field dependence. In addition strong favorable collisionality dependence of energy confinement and transport is inferred from a two point scan. These findings are in line with previous NSTX results. NSTX found that Lithium coatings of the plasma facing components lead to a significant improvement on energy confinement time in ELM-free H-mode. NSTX found little effect of rotation/rotation shear on global confinement. However, rotation shear is acting on a small part of plasma by improving local transport with increasing rotation/rotation shear. This task and TP-9 will be individually closed out and merged into a new JEX.
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. The primary devices for this task are JET and C-Mod. However, C-Mod must obtain good density control through cryopumping to be able to reduce the collisionality to ITER levels, which has proven difficult. Within the ITPA it has been decided that this is not a high priority experiment for ITER. As a consequence, it has been decided to drop this experiment from the ITPA list.
CDB-9 Density Peaking Dependence on Collisionality
Four devices (JET, AUG, C-MOD, JT-60U) reporting very similar behavior of peaking increasing as collisionality decreases in H-mode. The emphasis has shifted to identifying other, less obvious dependencies and understanding behavior in devices and confinement modes that appear to deviate from these observations. It has been known that L-modes exhibit a clear dependence on q95 or other measures of magnetic shear, for example TCV reported neo/<ne>≅j0q0/<j> with no or little collisionality dependence. Now a similar dependence has been reported in JET, MAST and NSTX L-modes. C-MOD H-modes also exhibit a dependence on q95, which is highly correlated to the above parameter. However evidence for a q95 or li dependence in AUG and JET H-modes has until recently been faint at best. In a new set of JET 2006-07 data dependences on additional parameters have been identified. For νeff<0.5, the best empirical fits obtained are
n0.2/<ne>=1.43±0.10 - (0.63 ± 0.07)νeff0.5 + (1.92±0.32)Γ'+ (0.34±0.09)li and
R/Ln=2.3±0.6 - (2.7 ± 0.4) * νeff0.5 + (12 ± 2) Γ' + (1 ± 0.5) li, where Γ'=eTiS/Qi is the ratio of particle flux to the ion heat flux at mid-radius. No dependence on R∇Ti/Ti was observed. Γ' and Ti/Te are strongly correlated in the dataset and their respective influence cannot be resolved. R/Ln is taken between ρ=0.2 and ρ=0.8. A dataset of more than 1000 linear GS2 simulations with parameter scans for νeff, R/Ln, ŝ (shear), R∇Ti/Ti, Ti/Te representative of the JET operating domain, was produced, with Γ' as the result of each simulation. A fit of these GS2 simulations yields
R/Ln≅2.9-1.8νeff0.5+21Γ'-1.1 Ti/Te + ŝ. Using the experimental relations to eliminate Ti/Te and to express ŝ in term of li, it was found that R/Ln≅2.1-2.1 νeff0.5 + 13 Γ' + 0.9 li. The agreement with the experiment is surprising in view of the fact that former linear simulations have predicted R/Ln≅0 at finite collisionality. The latter however were done for the mode with maximum γ/k2, while the present ones were for max γ. Non-linear simulations using a wide spectrum of modes confirm that for particle transport, modes with max γ are most representative.
With these results, the expectation of a moderately peaked density profile in the ITER, n0/<ne> ~1.5, appears to be on a firm footing. It was therefore elected to close out CDB-9. Questions remain, such as the effect of alpha heating and the understanding of the differences between H- and L-mode behavior. However the pursuit of these questions does not easily lend itself to joint experiments. Impurity transport certainly lends itself to joint experiments and database activities. The general field of impurity transport is so vast that it would represent an open-ended investigation. We therefore chose to open an investigation focusing only on He transport in advanced scenarios (hybrid and steady-state plasmas).
CDB-10 H-mode Power Threshold Hysteresis and Access to Good Confinement
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. No experiments in this area have been done yet, but experiments are planned on JET, AUG, MAST, TCV and NSTX. This important task will continue in the new TC group.
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. Recent C-Mod experiments 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. ASDEX-Upgrade has characterized the low-density behavior of H-mode thresholds in both D and 4He plasmas at fixed values of plasma current and toroidal field, and found both species give similar values for the minimum density. Experiments to check the BT and IP dependence of the low density threshold are planned for JET, DIII-D, AUG and TCV. This task will carry over in the new TC group.
CDB-12 H-mode operation in H and He
During ITER”Ēs 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. In 2008, experiments to investigate the H-mode power threshold and confinement were performed with hydrogen plasmas in DIII-D and helium-4 plasmas in ASDEX Upgrade. In DIII-D, there is a clear trend of increasing H-mode threshold power with increasing injected torque, with the H-mode threshold power in hydrogen being approximately a factor of 2 greater than in deuterium plasmas. The confinement time is over a factor of 2 greater in deuterium than in hydrogen. In ASDEX Upgrade, experiments were performed in pure He plasmas using ECH, and in plasmas with Deuterium injection into either Deuterium or Helium plasmas. The power threshold was found to increase linearly with toroidal field and the power threshold was observed to be the same for 4He and deuterium over a large target density scan. In addition, the density of minimum L-H threshold power was the same regardless of species. The confinement time in helium-4 was lower than in deuterium by a factor of 0.7. Future experiments on DIII-D, AUG, JET and NSTX are planned, and this task will continue in the new TC group.
Resistive Wall Modes
New experimental results from NSTX, JT-60U and DIII-D are providing better insights into mechanisms governing the excitation, damping and control of RWMs. JT-60U has observed a new branch of the RWM - an energetic particle excited wall mode (EWM). NSTX has demonstrated global mode feedback control (maintaining a long pulse plasma over the ideal limit) and DIII-D has reported stable operation beyond the no-wall βN limit at nearly zero plasma rotation. On JET an improved understanding of the role of edge stability on resonant field amplification, has been achieved through detailed modelling.
Non-resonant Magnetic Braking
Most tokamaks with external error field coils have observed non-resonant magnetic braking e.g. on DIII-D, NSTX and JET. Experiments on DIII-D have shown the theoretically expected off-set of the velocity, in the electron direction, under non-resonant braking. Studies also show the density scaling of the non-resonant torque is not that expected from the neo-classical toroidal viscosity (NTV) theory, though this may be partially resolved by including the plasma response to the applied magnetic field. On NSTX studies using n=2 applied fields do show braking increasing with Ti5/2 as expected from NTV theory. These NSTX experiments also show the distinctly different character of resonant and non-resonant braking. On JET the non-resonant torque has been simulated with a model including momentum diffusivity and velocity pinch terms, giving the torque profiles for comparison with NTV theory.
Neoclassical Tearing Modes
Experimental data from several tokamaks (DIII-D, JET, NSTX, JT-60U) show clear evidence of the effect of plasma rotation on the threshold βN for the onset of an NTM; the threshold decreases as the amount of rotation is decreased. However, DIII-D shows the threshold continues to decrease in the counter rotation direction, particularly in the low rotation region; a result that is not yet well understood. Current theoretical models do not explain fully these experimental results. Recent experimental results from JT-60U emphasise the importance of phase matching in the control of NTMs using ECCD.
Recent experiments on C-Mod have revealed interesting new results on the confinement/loss of LHCD enhanced runaways during gas-jet triggered disruptions. The lack of avalanching during these disruptions may be due to enhanced loss of Runaway Electrons due to other mechanisms, but more studies are needed to confirm this. Massive Gas Injection experiments on ASDEX-U have examined the dependence of fuelling efficiency on a number of parameters including the valve-plasma distance, the gas pressure, the type of gas, the gas quantity, the plasma energy etc. and found it to be independent of the gas pressure and the gas quantity, as well as the plasma energy. On JT-60U the disruption current quench rates have been reproduced based on measured temperature and Zeff profiles, using an inductive circuit model that includes the time variation of the internal inductance.
With regard to future plans from June 2009 to May 2010, it is expected that joint experiments on Disruptions, Neoclassical Tearing Modes, Resistive Wall Modes and non-resonant magnetic braking will continue, together with the related exchanges of personnel.
PEP-1+3 Dimensionless identity experiments in JT-60U and JET studies of ripple effects and rotation: G. Saibene, V. Parail, J. Lonnroth and A. Loarte (JET); N. Oyama, H. Urano, K. Kamiya, K. Shinohara (JAEA). New experiments carried out in September 2008 on JET explored identity plasmas up to 1% level of TF ripple. No further experiments are planned in JET. H. Urano (JAEA) will visit JET for 1 month in July 2009 to carry out the detailed analysis of the new data, with the aim of producing a joint publication by the end of the year.
PEP-2 Pedestal gradients in dimensionally similar discharges and their dimensionless scaling: A. Loarte, M. Kempenaars, G. Saibene, M. Beurskens (JET), T. Osborne, A. Leonard, M. Fenstermacher (DIII-D), C. Maggi, E. Wolfrum (AUG). The experiments have been carried out successfully in JET and DIII-D. Excellent identity points have been obtained in the two devices and the range of ρ* explored between the two machines is ~3.2 (M. Beurskens, invited paper at the 2009 EPS). The main result is that the pedestal Te and ne widths scale with machine size and no ρ* dependence is found, in agreement with earlier experimental results obtained in JET and DIII-D in 2002/03. Density and Te profile alignment is found in JET, while this is not the case in DIII-D. The neutral penetration model does not fit the measured pedestal widths, in contrast to earlier results that indicated that in DIII-D variations of the density pedestal width could be explained on the basis of simple atomic physics. ELM size scaling was compared between JET and DIII-D, with very different behaviour being found in the two devices: in JET, the ELM size does not scale with ρ* (this is consistent with the bulk of the JET data), while in DIII-D it is found that ELM size increases quite strongly with ρ*. These results and differences are not explained, and are under study. AUG will carry out experiments in the near future (from June 2009). An early completion of the AUG part may lead to a few pulses being repeated in JET to improve the match, if necessary.
PEP-6 Pedestal Structure and ELM stability in DN: H. Meyer (MAST), W. Suttrop (AUG), I. Nunes (JET), R. Maingi (NSTX), J. Hughes (C-Mod). Further experiments were carried out only on ASDEX-Upgrade (July and November 2008). A new scenario was developed within the limitations imposed by the damaged EZ2 fly-wheel generator. However, the scenario suffered from impurity accumulation during the DN phase because of improved confinement and lower ELM frequency. Measures to mitigate the impurity accumulation and influx by increasing the ECRH power and the NB power respectively were unsuccessful for various reasons. Nevertheless, the considerable progress made in this very demanding scenario development warrants new experiments planned for late 2009, when more ECRH power is available. The planned MAST experiments were deferred to 2009 because of the limited power available in 2008. Experiments on MAST are planned for Autumn 2009, when both NB systems are fully commissioned. New experiments are also planned in AUG for 2009, when the upper "divertor" is expected to be well-conditioned and one additional gyrotron has been installed. The experimental aim is to characterise the pedestal in DN under stable conditions and compare to LSN and USN at high confinement.
PEP-13 Comparison of small ELM regimes in JT-60U, AUG and JET: N. Oyama, H. Urano (JT-60U), E. Wolfrum (AUG), G. Saibene (F4E), A. Loarte (ITER), I. Nunes (JET), J. Hughes (C-Mod). No experiments were carried out or planned for on JET. In AUG, experiments are planned for later in 2009. No Type II or grassy ELMs were carried out in the last few years, because of the damaged fly wheel generator. Before carrying out the experiments, high triangularity, and high power scenarios will have to be tested at AUG, possibly in combination with protective measures for the W divertor, such as impurity seeding.
PEP-16 C-MOD/NSTX/MAST small ELM regime comparison: J. Hughes (C-Mod), R. Maingi (NSTX), H. Meyer (MAST). No new experiments were performed in the reporting period. The data from previous experiments were published at the 22nd IAEA Fusion Energy Conference, 13-18 October 2008 (EX/P6-4) and will be published in a refereed journal paper later this year. There is a clear difference in the structure of the NSTX Type V ELMs and the small ELMs at high collisionality and low βped seen on MAST and NSTX in DN, as well as the ELMs observed on C-MOD. So far no common parameter has been identified suggesting that the small ELMs in these three devices are of similar physics origin. No Type V ELMs have been observed on MAST; experiments are planned in Autumn 2009 to access Type V ELMs on MAST.
PEP-19 Basic mechanisms of edge transport with resonant magnetic perturbations in toroidal plasma confinement devices: T. Evans (DIII-D), J. Hughes (C-Mod), O. Schmitz (TEXTOR), R. Maingi (NSTX), A. Kirk (MAST), Y. Liang (JET), J. Moret (TCV). On MAST, experiments have been performed looking at the effect of the internal RMPs on L-mode plasmas. A large change in the edge turbulence characteristics has been observed during the application of the coils. The implications on the edge transport are being investigated. On JET, the plasma density drop by ~20% at core and edge (so-called density pump-out) has been observed with the application of either n=1 or n=2 magnetic field perturbationss. Reduction of the pressure gradient in the pedestal by ~20% (mainly due to a drop of density gradient rather than a change of temperature gradient) has been observed during ELM control with n=1 magnetic field perturbations. However, in contrast to DIII-D, no splitting of the strike point has been observed in the heat load profiles on the outer divertor even with a Chirikov parameter close to 1 for a normalised flux >0.915 (ν*~0.2). The measurement of particle flux with a fast visible camera may be made in the next JET experimental campaign in July-October 2009.
PEP-20 Documentation of the edge pedestal in advanced scenarios Pedestal: C. Maggi (AUG), R. Groebner (DIII-D), M. Beurskens (JET), N. Oyama (JT-60U). This was formally a joint activity between the Pedestal and the SSO ITPA groups, but continues now only as PEP-20. The experiment provides a good database of pedestal structure across machines. Analysis of AUG and DIII-D power scans in hybrid discharges (2005-2008 on AUG and 2007 on DIII-D) is complete and a paper will be submitted to Nuclear Fusion. JET has performed similar power scans for hybrid discharges and the data are being analysed. The low triangularity plasma shape is the AUG plasma shape, which was also reproduced at DIII-D. JT-60U obtained new pedestal data in hybrid discharges in 2008, which can be provided for comparison with the other machines. AUG and DIII-D will perform new experiments in 2009: a q95 scan in AUG (changing Ip at fixed Bt) and hybrid operation at higher density (similar to AUG densities) in DIII-D; here the main aim is to compare the pedestal structure, especially the pedestal density width, which is observed to be independent of β in AUG, whereas it is found to scale as (βpol,ped)1/2 in DIII-D.
PEP-21 The spatial and temporal structure of Type II ELMs: A. Kirk (MAST), T. Eich (AUG), R. Maingi (NSTX). The spatial and temporal evolution of small ELMs was measured on MAST during 2008. The next stage is to produce Type II ELMs on AUG, measure their characteristics and compare them to the characteristics of the small ELMs on MAST. The aim is to compare their characteristics with those of Type I ELMs established through PEP-10. The AUG experiments are awaiting the repair to the EZ4 fly-wheel generator which should be available from Summer 2009.
PEP-22 Controllability of pedestal and ELM characteristics by edge ECH/ECCD/LHCD: N. Oyama (JT-60U), A. Leonard (DIII-D), J. Hughes (C-Mod). No new experiments were carried out. New AUG experiments with CW ECRH are planned for 2009. 14 good pulses have been assigned, but they will be carried out only after boronisation which will take place after high priority tungsten experiments with unboronised wall have been completed.
PEP-23 Quantification of the requirements for ELM suppression by magnetic perturbations from internal off mid-plane coils: T. Evans (GA), A. Kirk (MAST), Y. Liang (JET), A. Loarte (ITER), R. Maingi (NSTX). The aim is to verify the suppression requirement derived from DIII-D by achieving ELM suppression on other devices that are equipped with off-midplane coils. At present, the only other machine equipped with off-axis coils is MAST. ELM suppression has not been achieved on MAST, even though vacuum modelling shows that either the coils can produce a region for which the Chirikov parameter is greater than 1 wider than that required to produce ELM suppression in DIII-D. Hence, the width of the edge ergodised region may be a necessary but not sufficient condition to ensure ELM suppression on any device. Albeit with different coil set-up, experiments in JET showed that no complete ELM suppression was obtained by application of n=1 or n=2 magnetic field perturbations with a Chirikov parameter close to 1 for a normalised flux >0.915 which is one of the important criteria for the design of ITER ELM suppression coils.
PEP-24 Minimum pellet size for ELM pacing: P.T. Lang (ASDEX Upgrade), B. Alper (JET), L. Baylor (DIII-D). New or modified pellet injectors are under development in AUG, JET and DIII-D to enable the launch of small sized pellets optimised for ELM pacing. In AUG, a blower gun is being commissioned for launch speeds in the range 100-200m/s and repetition rates up to 100Hz. Injection from the Low (magnetic) Field Side and tangential injection is being set up. At DIII-D, the pellet dropper system was equipped with a deflection plate directing pellets perpendicular to the plasma and hence increasing the usable pellet speed. Even with this improvement, pellets were found accelerated toroidally by fast ions and do not penetrate sufficiently in H-mode plasmas to trigger ELMs. At JET the new High Frequency Pellet Injection system was installed and commissioned. Pellets were launched at up to 10Hz repetition rate and speeds up to ~200 m/s from the LFS and the vertical High Field Side track in a variety of different scenarios. ELM pacing was demonstrated at the maximum available pellet rate of 10Hz. Detailed studies of ELM triggering by pellets in different plasma regimes were performed. The dynamics of triggered and spontaneous ELMs was compared in the perturbative regime (fpellet < fELM natural). First results for triggered ELMs indicate slightly reduced peak heat fluxes in line with the observation of a slower decay of the power arriving at the target plates after the peak heat fluxes. First investigations confirm the potential of pellets to trigger ELMs under appropriate conditions. Reliable ELM triggering is also maintained with increased toroidal field ripple. Strong MHD activity is observed during pellet ablation. Ablation dynamics and its correlation with MHD activity and ELM onset are investigated by means of a new fast framing camera.
PEP-25 Inter-machine comparison of ELM control by magnetic field perturbations from midplane RMP coils: Y. Liang (JET), T. Evans (GA), A. Kirk (MAST), A. Loarte (ITER), R. Maingi (NSTX), W. Suttrop (ASDEX Upgrade). Work across devices is on-going for setting up a database including the effect of the magnetic perturbation field on ELM size and frequency and pedestal profile dependence on plasma parameters, i.e. q95, plasma rotation, poloidal beta, collisionality, and toroidal mode numbers for JET (n=1,2), DIII-D (n=1,2,3), NSTX (n=1,2,3), MAST (n=1,2) and AUG (n=1,2,3,4). The existing experimental results from most of these devices show common behaviour of ELM control with mid-plane coils. No complete ELM suppression has been observed so far in all devices. Density pump-out and increasing ELM frequency has been observed in JET, NSTX and MAST with applied perturbation fields. Plasma toroidal rotation braking due to an application of perturbation fields has been observed in JET, DIII-D and NSTX, but not on MAST. Multiple resonances in the ELM frequency as a function of the edge safety factor q95 with n=1 and n=2 perturbation fields were observed for the first time on JET. Additional joint experiments for completing the database are planned (JET in October 2009, DIII-D in 2010, NSTX in 2009-2010, MAST in 2009-2010 and AUG after 2012). Combining the experimental results with modelling and making a more detailed comparison of the experimental results on the devices will be carried out in parallel with experiments over the next years.
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 2008 - April 2009.
The new organization of ITPA has been implemented, strengthening links with ITER IO. Several sessions of the 11th ITPA meeting (Nagasaki, September 2008) were initiated by ITER IO. Presentations at ITPA meetings are now on the ITER IDM website.
T retention in ITER: calculations made by the US and EU during the ITER review period used different assumptions and methods. Bruce Lipschultz and Jochen Roth organized an effort to resolve these differences. 13+ scientists came to an ”ĘITPA ITER Tritium Inventory Assessment workshop”Ē, June 23-24, 2008 at MIT. Agreement was reached on a number of underlying parameters such as the flux of ions and atoms to all surfaces, the co-deposition rate at various surfaces and the neutron damage to W which leads to D retention deep within the W bulk. Most projections to ITER have been redone. However the effect of neutron damage on retention deep within the W is still in the process of being modeled. As expected, the present selection of PFCs for ITER (Be/C/W) lead to large uncertainties in retention (4.5 - 260g at 105s of full operation or 250 shots). Taking into account only ion implantation into W (ignoring nuclear damage) leads to a lower projected range of T retention for an all-W ITER (3-30g for the same period of operation).
Fuel retention in carbon: a wide range of retention fraction is observed, from 20-30% (JET) to 50% (TS) and up to 80% (recent results from DIII-D in un-pumped discharges) of the injected gas, while it is close to 0% on JT60U for saturated walls conditions. Because of varying pumping conditions, the absolute value of the retention rate, rather than the retention fraction, should be used. Recent work on JET and Tore Supra show that the fuel retention estimated from post mortem analysis and particle balance can come to a reasonable agreement (factor 2 in TS), provided they are carried out for similar plasma conditions and that an extensive post mortem analysis is performed. New JT-60U studies showed that the erosion areas of the main wall had significant D in the near surface that, although low per unit area, may rival the divertor retention. Co-deposition on tile sides in high erosion areas of the castellated Tore Supra limiter dominates the co-deposition in remote areas. Deep diffusion of D into CFC graphite was evidenced in erosion zones, and shown to be ~ 10% of overall retention in TS.
Fuel retention in high Z materials: the number of retention processes are quite varied. Bubbles and blisters can be good as they often release their trapped gas and inhibit diffusion deep into the bulk; however, such deformation damages the surface and may degrade material properties re heat load handling and shock resistance. The effect of neutron damage which creates traps for T retention throughout the tile is just starting to be examined. Such damage can lead to a maximum of ~ 1% [D+T]/W. Then, it is a matter of when the D/T implanted at the surface can diffuse to those traps and be retained. The simultaneous implantation of He along with the D/T fuel can affect the diffusion and trapping of D/T.
Dust: The present ITER dust safety strategy relies on measurements of gross erosion, and conservatively assumes a conversion factor fd between gross erosion and mobilisable dust equal to 1. Carbon machines report values of fd between 1-15%. For C, the main source of dust seems to be the peeling of thick deposited layers. The C gross erosion as measured by spectroscopy is often much larger than the collected dust but it is not clear whether they are related. For high Z, preliminary results indicate that gross erosion estimated from spectroscopy is marginal to explain the dust produced; transients may play a central role as the collected metallic dust is often spherical, due perhaps to droplets formed during transients (disruptions, strong ELMs, arcing). Further studies somehow directly linking dust types/quantities to erosion mechanisms is needed before projections to ITER can be made. Dust can have serious effects on plasma operation, e.g. Tore Supra compared operational space (and UFOs) with/wo extensive cleaning of dust and C flakes from all PFCs. Cleaning led to a dramatic reduction in UFOs and the operational space (heating power limits) were greatly increased. Expansion of such cleaning/non-cleaned comparisons to diverted machines and Be walls (JET ILW) will be crucial. It is proposed to launch a new DSOL task on coordinated dust injection experiments and associated modelling for code benchmarking. However, this is limited to carbon (eventually W) dust, as no Be dust is readily available.
ELMs: The characterization of mitigated ELMs is just starting. Pellet-induced ELMs in AUG are similar to intrinsic ELMs in terms of SOL plasma and heat loads at the same frequency. The measurements of the effect of ELMs mitigation with RMPs (Resonant Magnetic Perturbations) on divertor heat loads is primarily from DIII-D as other tokamaks either do not have RMP capability or have no ELMs. Experiments are hampered by lack of IR measurements at the outer strike point and matching non-RMP and RMP equivalent discharges. The general assessment is that mitigated discharge heat fluxes are similar to averaging over the ELM, between-ELM periods of a non-mitigated discharge. Given the minimal amount of information available and the importance to ITER it is clear much more information from DIII-D and now JET is needed.
Disruptions: New, fast bolometry systems are just coming online at several machines to diagnose the wall and divertor heat loads with mitigated disruptions (AUG, C-Mod, TCV). Complex behaviour occurs, with pronounced toroidal /poloidal asymmetries (peaked near the injection location) during the initial gas penetration. While this raises concerns for localized heating of the Be wall in ITER, the measurements show that when the radiation is highest (thermal quench) the asymmetries drop to ~ 1 (C-Mod). Different gas mixtures are under study for optimisation. Understanding the controlling physics is critical to progress further (cross machine comparison essential). It is recommended that the ITER system remains flexible in terms of the nature/amount of gas injected.
Material migration: Chemical erosion of C is now being studied for detachment parameters. The effect of seeding gases (N2, Ne, Ar) needs investigation. In JT-60U a localized set of W tiles served as the W source. W was found to be redeposited locally toroidally, while C (from 13C injection experiments) is transported further toroidally. The amount of W in the core was strongly dependent on core rotation, depending even more sensitively on the particle transport properties than the W edge source. In contrast, a complete ring of W divertor tiles in C-Mod has led to no observable level in the core. Be migration was also studied in JET using Be evaporation.
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 CDBM-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.
And the followings are by the IOS-TG:
IOS-1.1: ITER baseline, at q95=3, βN=1.8, ne=0.85nGW
IOS-1.2: Study seeding effects on ITER baseline discharges
IOS-2.1: ECRH breakdown assist at 20 toroidal angle
IOS-2.2: Ramp-down from q95=3
IOS-3.1: Beta limit for SS with ITER recommended q-profile.
IOS-3.2: Define access conditions to get to SS scenario
IOS-4.1: Access conditions for hybrid with ITER-relevant restrictions
IOS-4.2: ρ* dependence on transport and stability in hybrid scenarios
IOS-5.1: Ability to obtain and predict off-axis NBCD
IOS-5.2: Maintaining ICRH coupling in expected ITER Regime.
IOS-6: Modulation of actuators to qualify real-time profile control methods for hybrid and steady state scenarios.
Significant progress was made in the period June 2008 to May 2009 in view of development of steady state operation and integrated operation scenarios for ITER.
For the steady-state scenario (SSO-1, IOS-3.1, and IOS-3.2), experiments to study beta limit and access conditions were carried out, showing improved performance significantly in DIII-D, JET and JT-60U by optimizing current profiles. In DIII-D, fully non-inductive operation at high βN ~4 and fBS~0.7 was established using shape optimizations for stability, confinement and density control that can be applicable to ITER. High li operation at qmin~1 was also demonstrated at much higher βN>4.5 with fBS~0.9 and fCD>1, showing possibility of access to βN >3 in ITER without wall stabilization. In JET, using Ip overshoot technique, high βN~3 and good confinement HH~1.2 achieved with qmin>2. In JT-60U, fully non-inductive CD with relaxed current profile has been demonstrated maintaining qmin above 2 using off-axis LHCD in order to avoid NTM (m/n=3/2 and 2/1), but below the no-wall limit. High βN=3 discharge sustained for 5s and high bootstrap fraction (fBS=0.9) discharge, both exceeding the no-wall limit, were demonstrated.
For the hybrid scenario (SSO-2.1, SSO-2.2, SSO-2.3, IOS-4.1 and IOS-4.2), experiments to study the physics of stability and access condition in hybrid plasmas were performed. DIII-D demonstrated successful ELM suppression using n=3 RMP in hybrid plasma at βN=2.5. Access to hybrid scenario using large bore startup with early X-point formation was also demonstrated in DIII-D at βN=2.9 and G=0.42 exceeding ITER Q=10 target. In JET, strong current ramp-up followed by ramp-down produced significant broadening of q profile, where good performance was obtained at βN ~3, HH~1.2, and qmin~1.3. JT-60U have demonstrated steady hybrid scenario discharge at βN=2.6 and HH=1 (G=0.54) for as long as 25s (about 14 times current relaxation time).
Joint experiment on study of ITER baseline scenario (IOS-1.1 and IOS-1.2) is newly proposed in IOS. DIII-D demonstrated βN=1.8 scenario in ITER shape, however, the scenario was close to stability limit of n=1 tearing mode leading to frequent disruptions. Dedicated baseline scenario demo experiment up to Ip=3.5MA at q95=3 is planned in JET in April 2009. Nitrogen seeding in H-mode was investigated in JET, showing strong reduction of heat load at outer strike point.
For development of feedback control for advanced scenarios (SSO-3 and IOS-6), analysis of JT-60U high fBS~0.7 discharges contributed to identify eigenmodes of actuators on q, Ti and toroidal rotation, using a state-space model developed in JET.
The documentation of the edge pedestal in hybrid scenarios (SSO-PEP-1) has progressed in JET, where systematic documentation of pedestal as a function of edge q (or q profile) in baseline and AT scenarios is ongoing.
On simulation and validation of ITER startup to achieve advanced scenarios (SSO-5) and on startup with ECRF assist (IOS-2.1): Analysis on ITER like low voltage breakdown in AUG, DIII-D, Tore Supra, JET and JT-60U shows breakdown less than 0.3V/m is possible in Tore Supra and JET even without RF assist, while in all devices with RF assist. Threshold power for startup assisted by second harmonic ECRF was between 0.7 and 1.1 in DIII-D.
Ramp down scenario from q95=3 (IOS-2.2) is a new proposal in IOS. Ip ramp down with βp change at H-L transition was demonstrated in DIII-D. In JET, Ip ramp down scenario is to be examined in ITER baseline scenario discharge (IOS-1.1), planned in April 2009.
On the ability to obtain and predict off-axis NBCD (SSO-6 and IOS-5.1): Results on experiments in 2008 was reported from DIII-D, and JT-60U. Good agreement between measured and calculated off-axis NBCD observed in DIII-D, using up/down shift of plasma with respect to NB lines. Since JT-60U previously reported good agreement between measured and calculated off-axis NBCD except on the CD location, analysis on the CD location for off-axis NBCD, including slight off-axis NBCD using N-NB as in ITER, is reported. It seems measured and calculated CD location agrees fairly well, however, further analysis in detail has to be done.
On maintaining ICRH coupling in expected ITER Regime (IOS-5.2), documentation of the requirements for improving wave coupling and system parameters needed for ITER ICRF system is intended. Dedicated experiment was carried out in JET in 2009. Gas injection level for optimized coupling strongly varied with plasma shape. High power ICRH experiment planned for later in 2009 in DIII-D, where electron density measurements directly in front of the antenna will be performed.
Concerning personnel exchange, J. Ferron, T. Luce, M. Murakami and JM Park (US) participated remotely in many JET experiments, and T. Luce visited JET in June and September 2008, where the topics covered fall under IOS-3.1, 3.2, 4.1 and 4.2.
De-tritiation of plasma facing components at JET
A number of candidate methods for de-tritiation of first wall components have been developed on JET (laser, flash-lamp, plasma torch, Inside Gap Plasma Generator), and in 2008 there has been a move towards practical applications of the techniques. Laser de-tritiation has continued to be optimised for access to awkward areas, e.g. by using the mock-up of the corner of the JET divertor (inner divertor tiles 3 and 4). The laser power thresholds for removing deposited films and substrate material have been explored, so that power levels may be set to remove deposits but not damage the substrate. The efficiency of each of the techniques has been tested for the removal of deposits from castellations, as this may be an important issue because of the limits on deposition on hot surfaces in ITER. During 2008, further experiments were carried out in JET of Laser Induced Breakdown Spectroscopy (LIBS), which offers the promise of obtaining the chemical composition of co-deposited layers, including the concentration of T/D/H in-situ. In previous trials, a laser-induced plasma was generated, but with insufficient intensity to generate the characteristic spectroscopic lines. For the 2008 experiments, the focussing of the JET LIDAR laser onto the inner divertor tiles was greatly improved and the laser fluence increased. Characteristic spectroscopic lines were observed. Ex-situ calibrations are now needed, together with modelling activities, so that the concentrations of the species in the layer can be extracted from the optical spectra. 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 within the JET Vacuum Vessel.
Surface analysis of plasma facing components from JET
Analysis of tiles removed during the 2007 Shutdown, and modelling the 13C transport, has continued following the puffing of 13CH4 on the last day prior to the 2007 Shutdown. A further puffing experiment is planned for the last day before the 2009/10 Shutdown for the installation of the ITER-like wall (ILW). This may be the last opportunity to puff 13C into JET since during the shutdown all carbon-based plasma-facing surfaces will be removed. The analysis of mixed deposited materials is a major topic for ITER. JET provides a supply of Be-C mixed deposits, and many more samples will become available following the ILW Shutdown. Development of techniques for characterisation of these films such as the Ion Micro-probe and XPS/AES has continued, and will also be useful for Be-W and Be-W-C films coming from JET. Marker layers for Be substrates (using a Be coating on a nickel interlayer) and for W-coated divertor tiles have been developed, and will be applied to a selection of ILW tiles as soon as they are available. A programme for the collection of dust in JET during the ILW Shutdown is being developed, in consultation with ITER. This will be a unique opportunity to make a detailed assessment of deposition in the machine, and will provide data for the ITER safety case.
Dr. Grisham made four trips to collaborate with the negative ion beam groups at JAEA during the course of the past year. As in previous years, the principal thrust of this collaboration is to understand the physical mechanisms which have limited the performance of negative ion neutral beam systems, and to use this knowledge to improve future systems.
Most of the work this year was focused on finding ways to improve the voltage holding in accelerators of the sort being developed for the ITER and JT-60SA negative ion neutral beam systems, and on the design and validation of a beamlet steering concept for the JT-60SA and ITER accelerators. Improved modeling of accelerator geometries planned for these devices is now showing that, with the large acceleration gaps required for voltage holding in high voltage negative ion accelerators, each beamlet is influenced by the electric fields from many other beamlets as well as from the accelerator support structure. This means that any change to any part of the accelerator structure can easily affect the overall focusing of the total beam envelope. This requires either very careful control and optimization of the accelerator and the accelerator support structure, or alternatively, finding a way to reduce the size of accelerator gaps through improved voltage holding. A concept, magnetic insulation, was developed for doing this, as well as improving the operability of accelerators such as those needed for ITER, but it requires testing on a high voltage component test facility before it can be incorporated into accelerator designs. It is discussed in the first reference listed below.
This work resulted in a number of publications, including the following:
(1) "Magnetic insulation to improve voltage holding in electrostatic accelerators," L. R. Grisham, accepted for publication in Physics of Plasmas (2009).
(2) "Lithium jet neutralizer to improve negative ion neutral beam performance," L. R. Grisham, accepted for publication in AIP Conference Proceedings of Negative Ion Beam Symposium, Aix-en-Provence (2009).
(3) "Compensation of beamlet repulsion in a large negative ion source with a multi aperture accelerator," M. Kashiwagi, T. Inoue, L. R. Grisham, M. Hanada, M. Kamada, M. Taniguchi, N. Umeda, K. Watanabe, accepted for publication in AIP Conference Proceedings of Negative Ion Beam Symposium, Aix-en-Provence (2009).
IEA Large Tokamak Cooperation
|SUBJECT:||Development of high β scenarios for ITER|
|Date:||24-25 June 2008|
|Place:||General Atomics, San Diego CA, U.S.|
|Name(s) of attendees:||(All names of attendees are listed in the attachment.)|
Brief description of the activities in the Workshop W68
This Workshop was focused on discussions of recent experimental and activities that are in support of high β operation in ITER. This was a joint Workshop of the IEA Large Tokamak and Poloidal Divertor Implementing Agreements. The Workshop was attended by ~ 25 participants and 4 by teleconference from Japan and Europe. Many of the major facilities were represented.
The meeting consisted of 1-1/2 days of presentations by various participants focused on areas of important research including:
In addition, the recently developed ITER Research Plan for steady-state development was discussed. The final half-day of the meeting was devoted to open discussion of the outstanding issues for steady-state and hybrid regime development and the development of a coordinate plan to address these issues.
There was general consensus amongst the participants that significant progress has been made in demonstrating key elements of high β operating regimes for ITER. In particular, multiple experiments presented results in which sustained operation at the ideal MHD limit, high fractions of self-generated bootstrap current, and excellent confinement quality had been achieved. There was also general consensus that the development of these regimes would be aided measurably by coordinated multi-tokamak experiments in order to determine the common (and extrapolable) elements important in achieving sustained high b operation. Near-term plans for personnel exchanges were discussed and developed along these lines.
The discussion of the ITER Research Plan pointed out towards several potential deficiencies in ITER's plan for steady-state operation including: insufficient scenario development time in early phases, insufficient ability to control resistive wall modes at high β, and insufficient off-axis current drive for sustaining fully non-inductive operation.
The full agenda for the Workshop can be found at the Workshop Web Site http://fusion.gat.com/conferences/w68/.
|Shunsuke Ide||JAEA (remote)|
|Yutaka Kamada||JAEA (remote)|
|David Campbell||ITER-IO (remote)|
|Wayne Houlberg||ITER-IO (remote)|
|Phil Snyder||General Atomics|
|John Ferron||General Atomics|
|Mickey Wade||General Atomics|
|Craig Petty||General Atomics|
|Charles Greenfield||General Atomics|
|Peter Politzer||General Atomics|
|Tim Luce||General Atomics|
|SUBJECT:||Seventh Joint Workshop on Large Tokamak, Poloidal Divertor and TEXTOR IA's "Implementation of the ITPA Coordinated Research Recommendations"|
|Date:||11-13 December 2008|
|Place:||Plasma Fusion Science Center (PFSC, Massachussetts Institute of Technology (MIT), Cambridge, MA, U.S.|
|Name(s) of attendees:||(All names of attendees are listed in the attachment.)|
Brief description of the activities in the Workshop W69
The Workshop was the seventh in the series and was held jointly by the three tokamak-related IEA Implementing Agreements (IAs) and the International Tokamak Physics Activity (ITPA). The ITPA has been operating under the ITER auspices since February 27, 2008. The ITER Science and Technology (S&T) Department was well represented at the meeting with the attendance of the Deputy Director General (DDG) and the Assistant DDG at the meeting, and with the participation of several ITER S&T staff by televideo from Cadarache. The scope of the workshop was expanded to include also discussions of a broad range of ITER physics R&D needs in addition to the planning of Joint Experiments. The workshop duration was increased by an additional day in order accommodate the increased scope of the workshop. While recognizing that the ITPA is the most effective international body in place for generating coordinated experiment plans across a wide range of fusion research topics, the Workshop aimed to stimulate and facilitate increased multi-machine Joint Experiments amongst the various tokamak programmes.
The Workshop was attended by ~ 34 participants on site and 4 by televideo from Cadarache, including the Chairs and additional ExCom members of the three tokamak-related IEA IA”Ēs, the Chair and additional members of the ITPA Coordinating Committee, the Chairs (or their representatives) of the six ITPA TG”Ēs, 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 tokamaks were unable to attend the Workshop. Specifically, the Workshop:
Minutes of the 24th Executive Committee Meeting for
the IEA Large Tokamak Cooperation Programme
21-22 May 2009, EFDA-JET, Culham Science Centre, Abingdon,
Oxfordshire, OX14 3DB, UK
HOW Room, Building K1
|Attendees:||M. Mori (JA)||: Member|
|Y. Kamada (JA)||: Member|
|K. Kamiya (JA)||: Secretariat|
|E. Oktay (US)||: Member|
|R. Hawryluk (US)||: Member|
|E. Marmar (US)||: Alternate|
|T. Taylor (US)||: Alternate|
|P. Gohil (US)||: Expert|
|R. Wilson (US)||: Expert|
|F. Romanelli (EU)||: Member|
|M. Watkins (EU)||: Alternate|
|H. Zohm (EU)||: Expert (remote)|
|U. Samm (EU)||: Expert|
|C. Gill (EU)||: Expert|
|G. Sips (EU)||: Expert|
|M. Kwon (KO)||: Expert|
The Twenty-fourth Executive Committee Meeting for the IEA Implementing Agreement on Cooperation among Large Tokamak Facilities was held at 21-22 May 2009, EFDA-JET, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, UK (HOW Room, Building K1). 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. F. Romanelli as the chairman until the next meeting (Dr. M. Mori has been replaced by Dr. M. Kikuchi, and Dr. R. Nazikian has been replaced by Dr. R. Wilson.). 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, E. Marmar, and R. Hawryluk), and M. Mori. 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;
Workshops and personnel assignments completed in the period of June 2008 - May 2009 are listed in Appendix E1. Two workshops on "Development of high βN scenarios for ITER (Combined Workshop with PD)" (W68), and "7th joint WS of LT (W69) PD and Textor IA's on implementation of the ITPA Coordinated Research Recommendations”É (W69) were carried out. The total number of personnel assignments completed in the period was 27. All personal exchanges were for review tours (less than 4 weeks) without any participation of more than 4 weeks (see Appendix E2). Subjects are summarized as follows (see Appendix E3): Task 1 (Transport and ITB Physics) was 6 (22%); Task 2 (Confinement database and modeling) was 1 (4%); Task 3 (MHD, disruptions and control) was 2 (7%); Task 4 (Edge and pedestal physics) was 6 (22%); Task 5 (SOL and divertor physics) was 1 (4%); Task 6 (Steady State Operation) was 3 (11%); Task 7 (Tritium and RH Technologies) was 1 (4%); and Task 8 (Other) was 7 (26%). The reports on the workshops (FORM C) and the short reports for review tours are attached as Appendices E4 and E5, respectively.
We should enhance the task activities, especially for Task 2, 5 and 7.
Proposed Workshops and Personnel Assignments for June 2009 - May 2010 are listed in Appendix F. This includes two Workshops (W70: "Key ITER disruption issues" and W71: "8th joint WS of LT, PD and Textor IA's on "Implementation of the ITPA Coordinated Research Recommendations"). The Committee discussed these proposals and authorized their implementation.
Following a clear presentation of what is required from the LT Executive Committee on this matter, Dr. F. Romanelli presented the latest version (See. Appendix G) of the draft of the new "Implementing Agreement for Co-operation on Tokamak Programs", and the attendees for the LT Executive Committee meeting extensively discussed the draft and made suggestions for revision.
The LT Executive Committee agreed that, after the meeting, the Chair will circulate the proposed amendments (incorporating the changes from the discussion at 22nd-May-2009) to all Executive Committee representatives. The Executive Committee representatives will then approach their respective authorities for approval of the Amendments.
The LT Executive Committee unanimously resolves to invite the Government of the Republic of Korea, or any entity it may designate, to join the Implementing Agreement on Co-operation on the Large Tokamak Facilities as a Contracting Party.
The LT Executive Committee unanimously resolves to request a five year extension from 15 January 2011 to 14 January 2016. The formal agreement will be sent by June 19 from each party.
There was extensive discussion on how to amend articles on three issues:
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 2009.
The next Executive Committee Meeting will be held in May 2010 in JAEA, Naka-site (Japan). It will be a joint meeting with PD IA.