How To Understand Fusion

Transcription

1 FUSION Technology Programme Technology Programme Report 1/2007 Final Report

2 FUSION Technology Programme Final Report Seppo Karttunen and Karin Rantamäki (Eds) Technology Programme Report 1/2007 Helsinki 2007

3 Tekes Your contact for Finnish Technology Tekes, the Finnish Funding Agency for Technology and Innovation, is the main funding organisation for applied and industrial R&D in Finland. Funding is granted from the state budget. Tekes primary objective is to promote the competitiveness of Finnish industry and the service sector by technological means. Activities aim to diversify production structures, increase production and exports and create a foundation for employment and social well-being. In 2007, Tekes will finance applied and industrial R&D in Finland to the extent of 460 million euros. The Tekes network in Finland and overseas offers excellent channels for cooperation with Finnish companies, universities and research institutes. Technology programmes part of the innovation chain Tekes technology programmes are an essential part of the Finnish innovation system. These programmes have proved to be an effective form of cooperation and networking for companies, universities and research institutes for developing innovative products, processes and services. Technology programmes boost development in specific sectors of technology or industry, and the results of the research work are passed on to business systematically. The programmes also serve as excellent frameworks for international R&D cooperation. Copyright Tekes All rights reserved. This publication includes materials protected under copyright law, the copyright for which is held by Tekes or a third party. The materials appearing in publications may not be used for commercial purposes. The contents of publications are the opinion of the writers and do not represent the official position of Tekes. Tekes bears no responsibility for any possible damages arising from their use. The original source must be mentioned when quoting from the materials. ISSN ISBN Cover: Oddball Graphics Oy Page layout: DTPage Oy Printers: Libris Oy, 2007

4 Foreword In the end of 2006 there was a culmination of decades effort by the fusion community. Formal agreement to begin construction of the ITER experimental fusion reactor to Cadarache in France was signed. The FUSION technology programme has provided the national framework for fusion research and development activities of Association Euratom-Tekes. The objective of the programme was to promote collaboration between research institutes and the industry in R&D work for ITER. The programme has been a fully integrated project in the European Fusion Programme of the 6th Framework Programme. Tekes contribution has been focussed on technology: material technology (multimetal components, joining and beam welding methods and cutting robots), remote handling (virtual simulation, water-hydraulic tools and manipulators for divertor maintenance). In fusion physics, we have gained a lot of scientific visibility by participating in and co-ordinating experiments and research projects in JET. The programme has produced a lot of excellent and internationally recognised research results. Another important part of the programme has been education and training which have had a significant role from the very beginning of the programme. Many questions, such as climate change, energy crises, security, quality of life and well being are linked to the theme of energy and environment and thus, have a direct impact on the issue of fusion energy. From the good experiences our national technology programme, we are looking forward to contributing to a technology-driven ITER project, which should lead to an energy source that is both economically and socially acceptable. The future of fusion research and technology development in Finland will be very closely connected to international co-operation between industry and other big science projects like CERN and ESO. Tekes, the Finnish Funding Agency for Technology and Innovation, would like to express its sincere thanks to all the individuals, enterprises and institutes who have contributed to the programme. This gratitude is extended also to the international scientific and industrial fusion community, with special thanks being given to the Head of the Tekes Research Unit, Dr. Seppo Karttunen from VTT, who has carried out the programme in such an excellent way. Helsinki, January 4th, 2007 Tekes, the Finnish Funding Agency for Technology and Innovation

5 Summary This report summarises the results of the FU- SION technology programme during for the period between FUSION is a continuation of the previous FFusion and FFusion2 technology programmes that took place from 1993 to The FUSION technology programme was fully integrated into the European Fusion Programme in the sixth Framework Programme (Euratom), through the bilateral Contract of Association between Euratom and Tekes and the multilateral European Fusion Development Agreement (EFDA). The Association Euratom- Tekes was established in At the moment, there are 26 Euratom Fusion associations working together as an European Research Area. There are four research areas in the FUSION technology programme: (1) fusion physics and plasma engineering, (2) vessel/in-vessel materials, joints and components, (3) in-vessel remote handling systems, and (4) system studies. The FUSION team consists of research groups from the Technical Research Centre of Finland (VTT), the Helsinki, Tampere and Lappeenranta Universities of Technology and the University of Helsinki. The co-ordinating unit is VTT. A key element of the FUSION programme is the close collaboration between VTT, the universities and the industry, which has resulted in dynamic and sufficiently large research teams to tackle challenging research and development projects. The distribution of work between research institutes and industry has also been clear. Industrial activities related to the FUSION programme are co-ordinated through the Big Science Project by Finpro and Prizztech. The total expenditure of the FUSION technology programme for amounted to 14,9 million in research work at VTT and the universities with an additional 3,5 million for projects by the Finnish companies including the industry co-ordination. The funding of the FUSION programme and related industrial projects was mainly provided by Tekes (37%), Euratom (38%) and the participating institutes and industry (24%). The FUSION research teams have played an active role in the EFDA JET and Technology Workprogrammes. There are two clear focus areas in the FUSION programme: (i) EFDA JET work and (ii) EFDA Vessel/In-Vessel technology. Work on theoretical and computational fusion physics at VTT and the Helsinki University of Technology has been very productive and on a high scientific level. The main emphasis has been on participating in the JET Task Forces. Two persons have acted as Deputy Task Force Leader in TF-H (heating) and TF-T (transport). Several JET experiments have been co-ordinated by the Tekes association and the scientific contribution to the tokamak database has been important and visible. A remarkable arsenal of simulation codes have been developed, which has secured us a firm position in the European Fusion Programme. The principal topics have been plasma turbulence and transport studies, radio-frequency heating, edge plasma phenomena and plasma-wall interactions. The Association Euratom-Tekes participated actively in the European Task Forces PWI (plasma-wall interaction) and ITM (integrated tokamak modelling). Collaboration with other Euratom associations on the same topics has been active, too. In fusion technology, the focus has been on remote handling systems and the vacuum vessel and in-vessel materials, components. A major achievement during the programme period was the hosting of the ITER divertor test platform (DTP2), which was offered to Tekes association. The basic structure of DTP2 by a Finnish company TP-Konepaja Oy has been just completed at a VTT research hall in Tampere. Full size divertor

6 cassette mock-ups and the mover will arrive later in 2007 and testing can be started. Water hydraulic maintenance tools and manipulators have been developed and prototyped by the Tampere University of Technology in collaboration with Hytar Oy and Adwatec Oy. The hardware has successfully been tested demonstrating the feasibility of water hydraulic systems in the fusion reactor environment. A virtual design of various systems and operations has been a key element in development work for remote handling maintenance. Virtual models can reveal problems and weak points in the design, providing substantial savings in the development phase. Research on joining techniques and multimetal first-wall components, including the manufacturing and characterisations of potential materials and joints, has been carried out in collaboration with VTT, Metso Powdermet Oy and Luvata Oy (former Outokumpu Poricopper Oy). Mechanical testing of reactor materials under neutron irradiation started in collaboration with SCK-SEN and Risø with a VTT-designed radiation rigs installed in the BR-2 research reactor at Mol. Hot isostatic pressing (HIP) is a very promising joining method for the first wall components with cooling tubes. Small and medium-size mock-ups have been manufactured, tested and characterised, proving the good quality of joints. Advanced welding for vacuum vessel assembly have been studied by VTT and a design and prototyping of a cutting/welding robot was carried out by the Lappeenranta University of Technology. Development of superconducting niobium-titanium and niobium-tin wires for the ITER magnets is carried out mainly as industrial activity by Luvata Oy. In addition, the Association Euratom-Tekes participated in the socioeconomic studies and safety analysis for conceptual power plant in close collaboration with the other associations. The Finnish fusion research unit has prepared and contributed to over 225 articles in scientific journals and over 285 conference articles during the programme period A fair number of university degrees have been completed in the same period: seven Doctorates and nineteen Masters. The previous programme period was very busy with international conferences culminating to the big 22nd Symposium on Fusion Technology in Helsinki, September The most important international meeting in the present programme period was the 10th European Fusion Theory Conference, Helsinki, 8-10 September Several smaller meetings and workshops were hosted by the Association Euratom-Tekes in Two general seminars ITER Challenge for Industry, Espoo, 2003 and ITER Opportunities for Finnish Industry, Road Show, Tampere, 2006 were organised to promote the ITER project. The international ITER agreement was signed on 21 November 2006 and the project is now running. The Finnish industry sees the construction of ITER as a major opportunity and is ready to provide the technology and expertise for the project. The subsequent know-how and technology transfer gained from participating in ITER construction will strengthen the industry and make it more competitive in future technology markets. The future Fusion technology activities commencing in 2007 will continue to foster co-operation with the industry and to support it in technology development and in the challenge of building the ITER.

7 Contents Foreword Summary 1 FUSION Technology Programme Background European Fusion Research Programme and ITER FUSION Programme Objectives FUSION Research Areas Participating Institutes and Companies Tekes, the Finnish Funding Agency for Technology and Innovation Finnish Fusion Research Unit Industrial companies National Steering Committee FUSION Programme Funding International Collaboration Association Euratom-Tekes Participation in the committees of the EU fusion programme European and other international collaboration Public Information Fusion Physics And Plasma Engineering Transport Studies in Candidate Operation Scenarios ITER baseline scenario: ELMy H-mode Interpretive transport analysis of toroidal momentum transport in the ELMy H-mode scenario Transport simulations of toroidal momentum transport Predictive transport modelling of transient transport experiments Predictive transport modelling of hybrid scenario plasmas Predictive transport modelling of advanced tokamak scenarios with ITBs Predicting the ITB dynamics with the measured values of the poloidal rotation velocity on JET Effect of toroidal ripple on ion transport MHD Stability and Plasma Control Edge localised modes (ELMs) Real-time control techniques to control q and the strength and location of the ITB in predictive transport simulations Stochastization of magnetic fields and magnetic reconnections...30

8 2.3 Plasma Heating and Current Drive Ion cyclotron heating experiments in JET LHCD experiments at JET Gyrotron theory Gyrotron development for ITER Energetic Particle Physics Confinement of and heating by fusion alphas in standard H-mode and advanced scenario discharges in JET Fast Ion Distribution in H- and QH-mode operation of AUG Simulation of the high energy NPA signal in AUG Particle fluxes and power loads on plasma-facing components due to co- and counter-nbi injection in AUG Fusion-born tritium in the plasma-facing components of a fusion device Edge poloidal rotation and radial electric field due to NBI-generated ions in AUG Modelling of ITER divertor load distributions using ASCOT Theory and Code Development First principles turbulence simulations with ELMFIRE code Development of the ASCOT code Particle-in-cell modelling of plasmas ERO plasma-wall code development Development of JET Codes Diagnostics Upgrading of AUG neutral particle analyser Design and fabrication of NPA detector for JET Optimisation of W7-X interferometer Smart tiles for first wall diagnostics Micromechanical magnetometer Fusion Reactor Materials Research Characterisation of Copper Alloy to Stainless Steel HIP Joints Tensile and fracture behaviour Stress corrosion cracking susceptibility of tube joints Thermal fatigue behaviour of small scale FW mock-ups In-reactor Material Testing under Neutron Flux Tensile tests Creep fatigue tests Advanced Welding of ITER Vacuum Vessel Sectors Keyhole laser and hybrid laser welding of thick sections Conduction-limited hybrid laser welding of thick sections Root control of electron beam welding Testing of a intersector welding robot (IWR) by laser welding Radiation Damage in EUROFER FeCrHe Thermodynamics Fusion Neutronics Radiation transport and shielding calculations...78

9 3.5.2 ITER neutronics IFMIF neutronics Neutronics summary Material Transport and Plasma-Wall Interactions Material transport and erosion/re-deposition in JET and AUG tokamaks Surface analyses of JET and AUG divertor tiles Erosion and deposition Material transport in SOL Simulations of trace methane puffing experiments at AUG ERO simulations at JET MD simulations of plasma-wall interactions Flake formation of re-deposited amorphous carbon films Deuterium irradiation induced defect concentrations in tungsten Remote Handling Systems Divertor Test Platform for ITER Introduction Design Implementation design Safety analysis Future PREFIT Development of Water Hydraulic Tools and Manipulators for ITER Divertor Maintenance Development of water hydraulic manipulators Development of divertor maintenance equipment (CMM) Conclusions Development of a High Precision Intersector Weld/Cut Robot System Studies Socio-Economic Studies TIMES modelling of global energy systems External costs of Fusion Waste disposal or clearing of materials Time horizon questions Fusion Power Plant Conceptual Studies Safety assessment of conceptual fusion power plants Safety and economy requirements of fusion power reactors in co-existing advanced fission plants Industrial Projects Development of ITER Superconducting Wires Superconducting wires from Luvata Pori Oy Gross-checking of the superconducting strand acceptance tests Advanced Fabrication Methods for Vacuum Vessel Company background Hollming Works Finnish technology partner in fusion energy development...126

10 ANNEXES A FUSION Projects and EFDA Tasks B Participating Institutes and Companies C Doctoral, Licentiate and Graduate Theses D Publications and Reports Fusion Physics and Plasma Engineering Publications in Scientific Journals Fusion Plasma Physics Conference Articles Fusion Plasma Physics Fusion Technology Publications in Scientific Journals Fusion Technology Conference Articles Fusion Technology Research Reports Fusion Technology General Articles E Abbreviations Tekes Technology Programme Reports in English

11 1 Fusion Technology Programme VTT, Technical Research Centre of Finland Materials Performance Seppo Karttunen, FUSION Programme Manager 1.1 Background Most projections show world energy demand doubling or even trebling in the next 50 years. This derives from fast population growth and rapid economic development. Energy sources that are not yet fully tapped include biomass, hydropower, geo-thermal, wind, solar, nuclear fission and fusion. All of them must be developed to meet future needs. Each alternative has its advantages and disadvantages regarding the availability of the resource, its distribution globally, environmental impact, and public acceptability. Fusion is a good candidate for supplying base load electricity on a large scale. Fusion has practically unlimited fuel resources, and it is safe and environmentally sound. The next step in the fusion development work is the International ITER project, which will open the way ( iter in Latin) for the future by demonstrating the scientific and technical feasibility of fusion energy production. Harnessing fusion energy is one of the most challenging missions of mankind, so global collaboration, such as ITER, is the best approach to meeting this challenge. In Europe, fusion research is a fully integrated and co-ordinated programme that makes good use of the resources in the most efficient manner. Joint experiments such as JET joint European torus have made Europe the leader in fusion research around the world. During the sixth Euratom framework programme, the European fusion development agreement (EFDA), JET implementing agreement (JIA) and the contracts of association were the main tools for implementing the Euratom fusion programme. EFDA covers both JET experimental work and technology activities. The Association Euratom-Tekes was established on March 13, Tekes became the 14th Euratom association in the EU fusion programme. At the end of the FP6, there are 26 Euratom fusion associations from the European Union, Switzerland, Romania and Bulgaria. The FUSION technology programme, from 2003 to 2006, was a continuation of the previous FFusion 2 technology programme. The national programme structure has been very useful in organising the national activities and bringing together smaller research groups to form more competitive units for the European fusion programme. This has clearly strengthened the role of Finnish fusion research in the international fusion community. The national programme period corresponded to the framework programme cycle of four years, which helped in adapting to new trends and priorities, rules and funding schemes. The main objectives of the FUSION technology programme are to carry out high-level scientific and technology research and development in the key areas of fusion physics and reactor materials and to provide technology for ITER construction and to benefit from technology transfer by applying the know-how to other technology fields. An effective collaboration among universities, research institutes and industry is crucial to meet these objectives. The research activities are focused on a few important topics in fusion plasma physics, reactor in-vessel materials and components as well as remote handling maintenance systems. Approximately one third of the FUSION programme is devoted to fusion physics and plasma engineering and two thirds to fusion technology. Industry is involved in the most technology projects and in some physics related projects such as diagnostics development, plasma facing materials and coatings for the present experiments and ITER. 1

12 The emphasis in the fusion physics research has been on the joint experimental work at the JET facility, the leading fusion experiment in the world. This has provided challenging opportunities, tasks and responsible positions for our young scientists and engineers to work in a fully international environment and produce new exciting data in support to the ITER project. The task force activities in high priority areas at JET and ASDEX Upgrade in Germany have considerably increased the productivity and visibility of the Finnish fusion research. Fusion technology activities are focused on the ITER vessel and in-vessel field including remote handling systems, materials, joining methods and multi-metal components. One of the highlights during the programme period is the hosting of the ITER divertor test platform (DTP2) at VTT, which underlines the high quality of the Finnish R&D in remote handling technology. Other activities cover water hydraulic tools and manipulators for remote maintenance operations and welding robot development, in which virtual design plays a key role. Regarding the materials research, manufacturing processes are developed in the industry and research institutes are responsible for testing and the characterisation of materials, joints and components. In-reactor mechanical testing of materials under neutron radiations and modelling of radiation damage effects gave new very interesting results. The industry plays a major role in the ITER construction, but it needs support from fusion technology experts in design, testing, quality control and consulting in nuclear and fusion specific issues. However, in some technologies the fusion research community, i.e., Euratom fusion associations will be the main contractors for ITER procurements. These are, for instance, diagnostics, plasma heating and current drive and remote handling systems. 1.2 European Fusion Research Programme and ITER The EU fusion programme is a fully integrated programme that includes all magnetic fusion research carried out in the Member States and Switzerland. The community funding for the 6th Framework Programme (FP6) was about 800 million. The total financing, including national funding, has been approximately 450 million per annum during FP6 ( ). The main elements of the European fusion programme are the association programmes that are defined by the bilateral contracts of association (CoA) with Euratom and the multilateral European fusion development agreement (EFDA), which covers fusion technology activities and the exploitations of the JET Facilities. JET is still the largest fusion research installation in the world, holding the world record fusion power of 16 megawatts. Since the beginning of 2000, JET facilities have been exploited by the research teams from the associations and the Euratom association UKAEA is responsible for operating the JET machine. JET research activities and operation are implemented by the multilateral JET implementing agreement (JIA) and by the JET operating contract (JOC) between UKAEA and Euratom. European participation in the scientific work on JET is co-ordinated by the EFDA close support unit (CSU) in Culham. The EFDA technology programme is co-ordinated by the EFDA CSU in Garching. The emphasis of the technology work programme is in the next step (ITER) activities to support ITER design. The main areas are physics integration, vessel/in-vessel technology, magnets, tritium breeding and materials. Other EFDA technology fields are the test blanket development, IFMIF (international fusion materials irradiation facility) design activities and system studies including safety and environmental issues, socio-economic research, conceptual power plant studies and public information. A third EFDA close support unit was established in Barcelona where the main activity in 2006 was the ITER site preparation. It will be replaced in 2007 by the Joint European Undertaking for ITER and the Development of Fusion Energy, which is a legal entity taking care of the European in-kind contributions to ITER and fusion technology development for ITER and Demo. A significant proportion of European fusion research is carried out in national laboratories under the contract of associations. There are several 2

13 Figure 1. Organisation of the European fusion programme during FP6 ( ). It consisted of the integrated programmes in the Euratom associations and the EFDA JET and EFDA technology activities. medium-size and small tokamaks and other plasma devices in the associated laboratories. A large superconducting fusion device Wendelstein 7-X stellarator is under construction at IPP Greifswald, Germany. The international ITER agreement was signed at Elysée Palace in Paris on 21 November 2006 and the project started from December The parties are European Union, Japan, Russian Federation, United States, People s Republic of China, South Korea and India representing over half of the world population. ITER will be built at the European site Cadarache in Southern France. The direct construction cost of ITER, including spares and R&D during construction, is nearly 5 Billion. ITER should start operation by the end of The overall objective of ITER is to demonstrate the scientific and technical feasibility of fusion as an energy source. The fusion power reaches megawatts with a power amplification of Q >10 in the standard scenario andq>5 in the steady-state mode with non-inductive current drive. New technologies and manufacturing methods have had to be developed during the ITER engineering design activities by the differ- Figure 2. JET Plasma chamber and a remotely operated manipulator changing the divertor tiles on the bottom of the machine. Radiofrequency heating antenna is seen on the left side. 3

14 Figure 3. ITER is a global project for demonstrating the scientific and technical feasibility of magnetic fusion. ITER parties are: European Union, Japan, Russian Federation, United States, People s Republic of China, India and South-Korea. ent industries and the research sector; for example, superconductors, multimetal first-wall components, various remote handling techniques, plasma heating systems and diagnostics methods. Regarding the recent progress in tokamak physics, the ignition (Q > 30) in ITER is not excluded. 1.3 FUSION Programme Objectives The FUSION technology programme is fully integrated into the European fusion programme, as the Association Euratom-Tekes. EU fusion programme has set a long-term goal of the joint creation of safe, environmentally sound prototype reactors, which should result in the construction of economically viable power stations. The national objectives of the FUSION programme are: to carry out high-level scientific and technology research and development in the key areas of fusion physics and reactor materials to strengthen competitiveness in the ITER project by effective collaboration among universities, research institutes and industry to provide technology for ITER construction and to benefit from technology transfer by applying the know-how to other technology fields. Active participation in the EFDA JET and technology programmes have provided challenging tasks and projects for Finnish science and technology community and industry. 1.4 FUSION Focus Areas of Research The FUSION technology programme is focused to the following research areas: joint experiments and modelling of the leading European fusion experiments such as EFDA JET and ASDEX-Upgrade research and development of in-vessel materials, joining techniques and manufacturing of multi-metal components development of ITER divertor maintenance systems and hosting of the European divertor test platform (DTP2). 4

15 In addition, there is a smaller effort on fusion neutronics, system studies and global energy modelling and testing of superconducting wires. Development and manufacturing of the superconducting wires for ITER is industrial activity by Luvata Oy (former Outokumpu Poricopper Oy). The physics programme is carried out at VTT, the Helsinki University of Technology (TKK), and the University of Helsinki (UH). The research areas in fusion plasma physics are: particle and energy transport, MHD phenomena and ITER operational scenarios fusion plasma engineering on radio-frequency heating and plasma diagnostics plasma-wall interactions and surface studies of plasma facing materials. The emphasis in fusion plasma physics has been on participating in the S/T Order and Notification work of the EFDA JET workprogramme. The contribution of the Association Euratom- Tekes consists of the scientific co-ordination of radiofrequency heating experiments, preparation, modelling and data analysis of experiments under the following Task Forces: H (heating), S1 and S2 (confinement and advanced scenarios), M (MHD), E (exhaust), T (transport), and FT (fusion technology). In addition, Tekes association has provided two deputy task force leaders for plasma heating and transport. Work has been carried out both on the JET site and by remote participation from VTT and TKK with an access to the JET computers and database. Plasma facing materials research and coating development include erosion/re-deposition studies, material transport in the edge region and tritium retention issues at JET and ASDEX Upgrade. Industrial partner in the coating technology is Diarc Technology Oy. Other European collaborations in fusion physics include IPP Garching and Greifswald (Germany), UKAEA Culham (UK), CEA Cadarache (France), IPP-CZ Prague (Czech Republic), FZJ Jülich (Germany), CRPP Lausanne (Switzerland) and VR (Sweden). The ITER tasks, dealing with the R&D and design of radio-frequency systems, have been partly performed under the physics programme or by the EFDA Technology Tasks and Contracts. These activities include optimisation of the ion cyclotron heating antenna, analysis of fast electrons and hot spots in the lower hybrid launcher as well as the design of the coaxial gyrotron for ITER. In addition, neutronics calculations and a nuclear analysis for various heating systems have been carried out. The technology programme of the Association Euratom-Tekes has been carried out at VTT, TKK, the Tampere University of Technology (TUT) and the Lappeenranta University of Technology (LUT), in close collaboration with the Finnish industry. The technology research covers the following fields: Remote handling systems and the hosting of the ITER divertor test platform (DTP2) ITER vessel/in-vessel materials, coatings, joining, beam welding and multi-metal components Development, manufacturing and testing of ITER superconducting wires System studies and global energy scenarios. The highlight of the programme period was the EFDA decision to give the hosting of the DTP2 facility to the Tekes association. The DTP2 facility is now under construction at VTT in Tampere. One reason for DTP2 hosting is our wide experience and active research in remote handling systems and tools over the past ten years. The focus of the vessel/in-vessel materials research is in the first-wall structures, joining techniques such as HIP and beam welding, multi-metal components and characterisation of materials and joints. Metso Powdermet Oy and Hollming Works Oy are the main companies involved in the vessel/in-vessel components. Mechanical testing of structural materials under neutron irradiation is an important topic and new interesting results have been obtained. Superconductor development is mainly industrial activity by Luvata Oy and TUT tests and characterises superconducting wires from European suppliers. System studies consist of global energy scenarios, estimates for the external costs of fusion and conceptual power plant studies. 5

16 Table 1. Expenditures in the main research areas of the FUSION technology programme in The detailed EFDA Technology Task and Contract summary is given in Annex A. Activity Projects 2003 (k ) 2004 (k ) 2005 (k ) 2006 (k ) Total (k ) FUSION Co-ordination Fusion Physics Underlying Technology EFDA Physics Integration EFDA Vessel/In-Vessel EFDA T-Breeding & Materials EFDA System Studies EFDA Technology Contracts EFDA JET Technology EFDA JET S/T EFDA Secondments Staff Mobility Visits Industrial Projects Total (keuro) The respective volumes of the FUSION research projects and industrial projects in the main research areas are given in Table 1. The summary of the EFDA Technology Tasks and Contracts in is given in Annex A. The industry is involved in the most technology tasks related to vessel/in-vessel materials and remote handling systems. In addition, there have been some industrial ITER design tasks through the EFET Consortium (European Fusion Engineering and Technology) in which Fortum Oyj is a partner. The underlying technology in reactor in-vessel materials includes the further development of fracture resistance test methods and verification of specimen size effects and the development of non-destructive examination techniques applicable to the inspection of primary wall modules. The other activity in underlying technology is the development of water hydraulic tools and manipulators. Collaboration in fusion technology with the EFDA CSU Garching (five Finnish secondees) and the Associations CEA (France), ENEA (Italy), FZK (Germany), Risø (Denmark), SCK- CEN (Belgium), VR (Sweden), UKAEA (UK) and CRPP (Switzerland) has played an essential role in the activities of the Association Euratom- Tekes. 1.5 Participating Institutes and Companies Tekes, the Finnish Funding Agency for Technology and Innovation Tekes, the Finnish Funding Agency for Technology and Innovation, is the main funding authority and co-ordinator for technological research and development activities in Finland. The fusion research co-ordinators in Tekes are Technology Manager Reijo Munther and Senior Technical Adviser Juha Lindén. 6

17 1.5.2 Finnish Fusion Research Unit Research activities in the FUSION technology programme are carried out at VTT, universities and industry. The co-ordinating institute is VTT and the Head of Research Unit of the Association Euratom-Tekes is Seppo Karttunen who was the programme manager of the FUSION technology programme. The Finnish Fusion Research Unit consists of the following research groups from the institutes and universities, which have been participating in the fusion research that occurred in : VTT, Technical Research Centre of Finland VTT Processes (FUSION co-ordination, fusion plasma physics, plasma-wall interactions, neutronics, socio-economic studies) VTT Industrial Systems (reactor materials, remote handling, welding) VTT Electronics (diagnostics). Helsinki University of Technology (TKK) Department of Engineering Physics and Mathematics (fusion plasma physics, diagnostics) Automation Laboratory (remote handling). University of Helsinki (UH) Accelerator Laboratory (plasma-wall interactions). Tampere University of Technology (TUT) Institute of Hydraulics and Automation (remote handling) Laboratory of Electromagnetics (superconductors). Lappeenranta University of Technology (LUT) Laboratory of Machine Automation (remote handling, welding robots). VTT was reorganised from the beginning 2006 and fusion related activities are in the following new VTT technology centres: Materials Performance (co-ordination, fusion plasma physics, plasma-wall interactions, reactor materials, diagnostics); System Engineering (DTP2, remote handling); Risk and Reliability Management (DTP2, remote handling); Industrial Automation (DTP2, remote handling); Production Systems (beam welding); Micro- and Nanoelectronics (diagnostics); Nuclear Energy (fusion neutronics, safety); Sensors (diagnostics); Energy Systems (energy scenarios, external costs); Industrial companies Three industrial groups were on the list of qualified companies for the EU fusion programme and ITER activities: 1. The Finnish Remote Handling Group: Advatec Oy, Creanex Oy, Fortum Oy, Hytar Oy, PI-Rauma Oy, Platom Oy, Rocla Oy and Tehdasmallit Oy. (Technology: 11. Qualification of Standards and Tools) 2. The Finnish Blanket Group: Aker Mäntyluoto Oy, Diarc Technology Oy, Fortum Oy, High Speed Tech Oy, Hollming Works Oy, Metso Engineering Oy, Metso Powdermet Oy, Luvata Oy, Patria Finavitec Oy and PI-Rauma Oy. (Technologies: 5. Plasma Facing Component Mock-Ups, 6. Vacuum Vessel, Shield and Tritium Breeding Blanket Segment Mock-Ups) 3. Outokumpu Poricopper Oy (from 2005 Luvata Oy) / Superconductors. (Technology: 7. Strand). Fortum Oyj is a partner in the European Fusion Engineering and Technology (EFET) Consortium. The EFET partners are: Ansaldo Richerche (Italy), Belgatom (Belgium), Fortum (Finland), Framatome ANP GmbH (Germany), Framatome ANP SAS (France), IBERTEF (Spain) and NNC (UK). Finland was represented in the Fusion Industry Committee by Juho Mäkinen from Outokumpu Oy. Fusion related industrial activities are co-ordinated by the Tekes BigScience project managed by Finpro and Prizztech 1.6 National Steering Committee The national steering committee of the FUSION technology programme prepares the Finnish fusion research strategy, advises in the planning of fusion research and promotes collaboration with the Finnish industry. 7

18 The members of the FUSION steering Committee are: Chairman Harri Tuomisto, Fortum Nuclear Services Oy Members Iiro Andersson, Prizztech Oy Ben Karlemo, , Outokumpu Poricopper Oy (Luvata Oy) Pentti Kotiranta, Finpro ( ) Juha Lindén, Tekes Reijo Munther, Tekes Olli Naukkarinen, , Outokumpu Poricopper Oy Pentti Pulkkinen, Finnish Academy Jouko Pulliainen, Metso Powdermet Oy Rauno Rintamaa, VTT Rainer Salomaa, Helsinki University of Technology Arto Timperi, 2003 Creanex Oy, , VTT Seppo Karttunen, VTT, Head of Fusion Research Unit Jukka Heikkinen, VTT, Secretary Tuomas Tala, VTT, Secretary There have been 15 FUSION Steering Committee meetings in the programme period between 2003 and FUSION Programme Funding The FUSION technology programme/activities of the Association Euratom-Tekes is financed by Euratom and by the national institutions of Tekes, the Finnish Academy of Sciences, the participating institutes (VTT, TKK, TUT, LUT and UH) and the industry. Figure 4 shows the yearly expenditures of the FU- SION technology programme from 2003 to 2006, including industry projects. The distribution of the total funding between the different organisa- Physics and PWI Underlying Technology Technology PS & Contracts EFDA JET S/T Technology Tasks Industry Projects Figure 4. Yearly expenditures by research fields (in millions of ) of the FUSION technology programme in including industry projects (co-ordination, R&D, EFDA Art. 7 contracts). The overall expenditure is 18.4 million. 8

19 tions during the four-year period is shown in Figure 5. The total funding of the FU- SION research activities for is approximately 14.9 million. The volume of the industrial activities related to the FUSION programme is about 3,5 million for the same period. Thus, the overall expenditure in is about million. The relative volume of the research and development work in the participating institutions can be seen in Figure 6. VTT accounts for about 38% of the research volume, the universities 41%, and the industry 21%. FUSION technology programme funding Euratom 38 % Tekes 37 % Universities 14 % VTT 9 % Academy 1 % Industry 1 % Figure 5. Distribution of the funding of the FUSION technology programme and the related industrial R&D projects between the different organisations for the period Euratom funding includes EFDA Art. 7 contracts with industry. The total value of the funding is approximately 18,4 million. FUSION technology programme research volumes LUT 6 % Industry 19 % VTT 42 % UH 5 % TUT 12 % HUT 16 % Figure 6. Research volumes of the participating institutions, VTT, the universities and the industry in

20 1.8 International Collaboration Association Euratom-Tekes The FUSION technology programme is fully integrated into the European Fusion Programme. The Association Euratom-Tekes was established when the contract of association between Euratom and Tekes was signed in Helsinki, on March 13, The present contract of association extends to the end of The Tekes association is the responsible organisation and partner in the multilateral agreements EFDA, JIA and staff mobility. The FUSION technology programme covers all research activities of the fusion research unit of the Association Euratom-Tekes. Association Steering Committee The activities of the Association Euratom-Tekes are steered by the association steering committee. It supervises the execution of the contract of association, adopts the details of the programme, ensures the progress of the research activities and steers them towards the programme objectives. The association steering committee also appoints the head of research unit on the proposal of Tekes. The members of the association steering committee were: Umberto Finzi, EU Commission, Research DG, 2003 Hardo Bruhns, EU Commission, Research DG (Chairman in 2004), Johannes Spoor, EU Commission, Research DG, 2003 Edward Rille, EU Commission, Research DG, 2004 Yvan Capouet, EU Commission, Research DG, Christopher Ibbott, EU Commission, Research DG, Marc Cosyns, EU Commission, Research DG, Doug Bartlett, 2006, EU Commission, Research DG (Chairman in 2006) Reijo Munther, Tekes (Chairman in 2003, 2005) Markku Auer, VTT, 2003 Jouko Suokas, VTT, Harri Tuomisto, Fortum Nuclear Services Jukka Heikkinen, VTT (Secretary) Seppo Karttunen, VTT (Head of Research Unit). The association steering committee has had 4 meetings during the period The steering committee approves annual accounts, yearly budgets and research programme, and the annual reports of the research unit Participation in the committees of the EU Fusion Programme The Finnish representatives on the various committees of the EU Fusion Programme are given below. Consultative Committee for the Euratom Specific Research and Training Programme in the Field of Nuclear Energy Fusion (CCE-FU): Seppo Karttunen, VTT Reijo Munther, Tekes EFDA Steering Committee (EFDA SC) Seppo Karttunen, VTT Reijo Munther, Tekes Administration and Financing Advisory Committee (AFAC) Juha Lindén, Tekes Rainer Salomaa, TKK Science and Technology Advisory Committee (STAC) Seppo Karttunen, VTT Rauno Rintamaa, VTT Rainer Salomaa, TKK Fusion Industry Committee (CFI) Juho Mäkinen, Outokumpu Oyj EFDA sub-committee on Public Information (EFDA CPI) Seppo Karttunen, VTT (CPI, Chairman) 10

21 The following fusion committees and expert groups have Finnish representatives: Reijo Munther is a members of the IEA Fusion Power Co-ordinating Committee. Jukka Heikkinen is a member of the Coordinating Committee for Fast Wave Heating (CCFW). Karin Rantamäki is a member of the Coordinating Committee for Lower Hybrid Heating and Current Drive (CCLH). Harri Tuomisto is a member of the International Organising Committee, of the Symposium on Fusion Technology (SOFT). Rainer Salomaa is a member of the ASDEX- Upgrade Programme Committee Rainer Salomaa was a member of the Programme Committee of the 30th EPS Conference on Controlled Fusion and Plasma Physics, St. Petersburg, Russia. Jukka Heikkinen is a the Chairman of the Scientific Committee of the International Workshop on Plasma Edge Theory (PET) Jukka Heikkinen is a member of the Scientific Committee of the European Fusion Theory Conference and he was the Chairman of the Local Organising Committee of the 10th European Fusion Theory Conference held in Helsinki, September Taina Kurki-Suonio is a member of the Programme Committee of the 34th EPS Conference on Controlled Fusion and Plasma Physics, Warsaw, Poland, Ad-Hoc-Group (AHG) memberships: Jukka Heikkinen in the AHG for the FOM Magnum- PSI project, Seppo Karttunen in AHG preparing the new agreements EFDA and CoA, Rainer Salomaa in AHG for monitoring the enhancements of JET heating and current drive systems European and other international collaboration In plasma physics and plasma-wall interactions, the Association Euratom-Tekes participates in the EFDA JET and ASDEX Upgrade work programmes. In addition, Tekes association participated actively in the two European task forces: PWI (plasma-wall interactions) and ITM (integrated tokamak modelling). ITER gyrotron development work is carried out in collaboration with the associations FZK Karlsruhe and CRPP Lausanne and ITER ICRF antenna design with ERM in Brussels. Optimisation of the interferometer diagnostics for Wendelstein 7-X stellarator was carried out in collaboration with IPP Greifswald. In fusion technology, Tekes association hosts the ITER divertor test platform (DTP2) which is under construction at VTT. DTP2 will be an European test facility for the development of ITER divertor maintenance systems. Other joint research projects are e.g. the in-situ materials testing under neutron irradiation together with the associations Risø and SCK-CEN in Belgium, development of water hydraulic manipulators and welding robots with CEA and fusion neutronics with FZK. The staff mobility scheme of the EU Fusion Programme has offered excellent opportunities for the exchange of scientists and engineers in Europe. There have been 35 staff mobility visits of 1 to 4 months in corresponding to 110 visit months. Longer visits have been made to JET for the task force leader duties. The longer assignments to the EFDA Close Support Units in Culham and Garching and UKAEA JET Operators Team were: Ben Karlemo at EFDA CSU Garching, 2003 Hannu Kaikkonen at EFDA CSU Garching, Hannu Rajainmäki at EFDA CSU Garching, Pertti Pale at EFDA CSU Garching, 2006 Johnny Lönnroth at UKAEA JOC, Herkko Plit at EFDA CSU Garching, 2003 Antti Salmi at UKAEA JOC, , 2006 Marko Santala at UKAEA JOC, Some collaboration with non-eu countries has also taken place, e.g., visits to JT-60U tokamak at Naka, Japan and DIII-D tokamak in San Diego, US related to the IAE large tokamak agreement, collaboration with the Ioffe Institute in St. Peters- 11

22 burg (fusion theory, Globus tokamak), the Institute for Applied Physics in Nizhny Novgorod (gyrotrons). Annual fusion symposiums between TKK, VTT and the Ioffe Institute have been organised. Fusion theory conference and several international workshops were held in Finland during the programme period: the 10th European Fusion Theory Conference, Helsinki, Finland, 8-10 September 2003; Finnish-Russian Symposium on Fusion Research and Plasma Physics, Espoo, Finland, September 2003 and November 2005; Workshop on New Coating Development for ITER and Other Demanding Applications, Siuntio, Finland November 2003; 7th International Conference on Computer Simulation of Radiation Effects in Solids, Helsinki, Finland, 28 June 2 July 2004; Workshop on Diagnostics in Fusion Devices, Espoo, Finland, 31 January 2005; 1st German-Finnish Workshop on Carbon Migration in Fusion Devices, Tervaniemi, Finland, February 2005; 10th Workshop on Multi-Scale Modelling of FeCr Alloys for Nuclear Applications, M/S Silja Serenade Cruise on the Baltic See, August Public Information Fusion research and ITER attracted increasingly interest in Finnish media during the programme period driven mainly by ITER negotiations. Several general articles and interviews on fusion energy and fusion research were published in newspapers and weekly journals. National TV channel YLE1 produced a 30 min science program Knights of Fusion Energy on fusion research in which several Finnish scientists and engineers were interviewed. In addition, a few shorter fusion reports in YLE1 Prisma science programme and interviews in morning TV took place. The ITER site decision and the signature of the ITER agreement received a lot of positive publicity in Finnish media. The Commission and EFDA public information activities provided a lot of excellent material such as EFDA Newsletters, JET Bulletin, fusion brochures (some translated in Finnish) and CD/ DVDs, which were widely distributed in seminars, conferences and various occasions. Other public information actions were: The summary seminar of the FFusion2 programme period was held in Espoo, February The second day of the seminar ITER-Challenge for Industry was devoted to ITER including an invited talk by the ITER Director Robert Aymar. The FUSION technology programme brochure on the research activities of the Tekes Association, Euratom and ITER development were published in Finnish 2003 and in English 2004 by Tekes and a brochure Finnish Industrial Interest in ITER was published by Prizztech in More detailed technology cards on plasma-wall and coating technology research and Finnish virtual reality expertise rated high related to design of remote handling systems and DTP2 facility. A lecture series in fusion technology and plasma physics at the Helsinki University of Technology in 2003 and 2005 and the Lappeenranta University of Technology in 2004 and ITER Opportunities for Finnish Industry Road Show, Tampere, Finland, 21 August Mini Expo at in Espoo and Tampere 2003 and at the Tampere University of Technology, August Studia Generale Lectures and Seminars for a broader audience. Invited talks on fusion and ITER in the Millennium New Materials Seminar, Mikkeli, Finland, September 2005 and in the Conference on Powder Metallurgy and Advanced Materials, Tampere, Finland, January FUSION Newsletter has appeared 2 3 times per annum during and Annual FUSION Seminars in with invited speakers from other fusion associations and EFDA CSUs took place in Tampere, Helsinki. 12

23 2 Fusion Physics and Plasma Engineering VTT, Technical Research Centre of Finland, Materials Performance Seppo Karttunen (FUSION Programme Manager), Jukka Heikkinen (Project Manager), Matti Kortelainen (TKK), Johnny Lönnroth (TKK), Markus Nora (TKK), Karin Rantamäki, Tuomas Tala (Deputy TFL T at JET) Helsinki University of Technology (TKK) Department of Engineering Physics and Mathematics, Laboratory of Advanced Energy Systems Rainer Salomaa (Head), Pertti Aarnio, Markus Airila, Otto Asunta, Olgierd Dumbrajs, Antti Hakola, Svante Henriksson, Ville Hynönen, Salomon Janhunen, Timo Kiviniemi, Pia Kåll, Taina Kurki-Suonio, Susan Leerink, Mervi Mantsinen (Deputy TFL H at JET), Francisco Ogando (Euratom Fellow, seconded from UNED), Antti Salmi, Marko Santala, Seppo Sipilä Students: Leena Aho-Mantila, Timo Ikonen, Simppa Jämsä, Tommi Kokki, Ville Tulkki Contributors to the EFDA JET workprogramme The ASDEX Upgrade Team 2.1 Transport Studies in Candidate Operating Scenarios To achieve burning plasma conditions for long periods of time, the operating regime in a fusion reactor will have to comply with a number of challenging conditions. The degraded plasma confinement is one of the main problems en route to a fusion reactor. This has led to both theoretical and experimental research of transport processes in the parameter range relevant for fusion energy production. To achieve a viable and economical way to produce fusion energy, different type of operation scenarios have been developed in several tokamaks. For the time being, the following three scenarios are considered as the most promising ones, and the transport studies in those operation scenarios are reported in this section. In the so-called ELMy H-mode Scenario, the plasma current is driven inductively. This mode of operation is very robust, routinely produced in practically all major tokamaks and under a wide variety of conditions, and it has been chosen as the primary operation mode or named as the baseline scenario in ITER. It takes advantage of the H-mode (for High confinement) with Edge Localised Modes (ELMs) governing the physics of the edge of the plasma. This scenario would indeed meet the goal of fusion gain of Q = 10 in ITER for a few hundred seconds. However, this performance is limited at moderate plasma pressure by the triggering different type of magnetohydrodynamic (MHD) instabilities. As a consequence, a new scenario, called as hybrid scenario has been developed in several tokamaks. It aims to eliminate or mitigate the deleterious MHD modes by modifying the plasma current profile or the q-profile (the q-profile or the safety factor, q ~ 1/current). In addition to improved MHD stability, it may possess some minor improved confinement and thus able to produce the same fusion performance at a lower plasma current as the ELMy H-mode Scenario. This scenario provides thus a promising route to ITER operation with reduced flux consumption (longer plasma discharges), high fluence and lower potential damage associated with disruptions. The third tokamak operation scenario is called the Advanced tokamak scenario. The main objectives of the Advanced Tokamak (AT) scenarios for ITER and, also, for an attractive tokamak reactor, are to increase the fusion power density and confinement and to render the tokamak compatible with continuous plasma operation. All these effects can reduce the size of the tokamak and as a consequence, reduce the cost of the tokamak reactor significantly. The main means to achieve these goals are to optimise the shape of 13

24 the current and the pressure profiles so that confinement improves. In order to improve confinement, plasma regions with reduced or killed turbulence are needed. These regions are called Internal Transport Barriers (ITBs). In order to be able to optimise the current density and pressure profiles, understanding the plasma transport and the onset and evolution of the ITBs as well as how they are affected by the current density profile, is crucial ITER baseline scenario: ELMy H-mode Predictive transport modelling of the effect of ripple-induced thermal ion transport on H-mode performance: Extensive MHD stability analysis of a series of dimensionless pedestal identity experiments at JET and JT-60U revealed that there is no significant difference in MHD stability between discharges from the two machines. Comparable JET and JT-60U discharges often both have access to second stability. The effects of the aspect ratio and of plasma rotation were explicitly included in the analysis, because JT-60U features a 15% larger major radius than JET and typically a rather different toroidal rotation profile. It was concluded that a difference in MHD stability is not the reason for why JT-60U plasmas generally have lower pedestal performance than JET identity plasmas. Given this result and the fact that JT-60U plasmas are characterised by far stronger toroidal magnetic field ripple than the corresponding JET identity plasmas, ripple losses were identified as a possible cause for the discrepancy in plasma performance. Losses of thermal ions, in particular, may play an important role in the performance of H-mode plasmas, because ripple-induced ion thermal transport can be significant in comparison with the residual level of transport in the pedestal. The influence of ripple losses on ion thermal transport has been studied in orbit-following simulations. These simulations have shown that ripple-induced transport can indeed considerably exceed the level of ion neo-classical transport and thereby affect the physics of the H-mode pedestal. Depending on the collisionality, both diffusive and convective (direct) losses can be important. Diffusive losses have been found to lead to a wide distribution of enhanced ion thermal transport, comparable in magnitude to or larger than the level of ion neo-classical transport, extending from the separatrix well beyond the top of the pedestal. Convective losses, on the other hand, have been found to be very edge-localised. The results of the orbit-following calculations have been applied in predictive transport simulations of ELMy H-mode plasmas. These simulations indicate that toroidal magnetic field ripple can influence the ELM behaviour and plasma performance very sensitively. In the case of convective losses, the so-called τ approximation has been used, in which a convective energy sink term is included into the continuity equation for the ion pressure. In these simulations, a deterioration of plasma confinement as well as an increase in the ELM frequency have been observed for a level of losses consistent with the orbit-following simulations. This result suggests that convective losses of thermal ions might play an important role in explaining the modest pedestal performance and high ELM frequency characterising many JT-60U plasmas. It also shows that toroidal magnetic field ripple could become an important tool for ELM mitigation. In predictive transport simulations with additional ion thermal transport matching the wide distribution of diffusive losses seen in the orbitfollowing simulations, an improvement in confinement and a reduction of the ELM frequency were observed. A somewhat similar improvement of performance and reduction of the ELM frequency were observed at the start of the H- mode phase, when the ripple amplitude was increased only slightly, in a series of experiments with enhanced toroidal magnetic field ripple at JET in A further increase in the ripple amplitude led to a deterioration of confinement and an increase in the ELM frequency, followed by a back H-L transition, which could suggest an interplay between the two mechanisms described here. Above all, the work shows that ripple losses of thermal ions seem to be a highly important effect 14

25 influencing the performance of tokamak plasmas in different and often counter-intuitive ways. These results may have profound implications for the design of future tokamaks and in particular for ITER, which is planned to operate with a level of toroidal magnetic field ripple higher than at JET and for which large divertor heat loads are a concern. A particularly interesting result is that ripple losses need not necessarily have a detrimental influence on plasma performance. On the contrary, better overall confinement than in the absence of ripple can probably be obtained by carefully choosing a suitable ripple amplitude profile. Predictive transport modelling of plasmas with mixed type I-II ELMs: Type II ELMs, a form of small, quasi-continuous ELMs, are interesting from a technical point of view, because they do not cause severe divertor loads, an issue that may be of critical importance in the design of future tokamaks. Mixed type I-II ELMy H-mode, an ELM regime with small, frequent, almost continuous type II ELMs interrupted by occasional, large type I ELMs, which has been observed at JET and other tokamaks is of interest, since it combines good confinement properties with slightly smaller divertor loads than a conventional type I ELMy H-mode. These regimes of operation have been studied in predictive transport simulations with the 1.5D core transport code JETTO using simple ELM models based on the often expressed idea that type I and type II ELMs are assiciated with a kind of magnetohydrodynamic (MHD) instabilities referred to as medium and high n ballooning modes, respectively. Type I ELMs are assumed to be caused by violations of the so-called finite n ballooning stability limit, whereas type II ELMs are assumed to occur when high n ballooning stability is violated at the very edge of the plasma. The models for type I and type II ELMs have been combined into an improved scheme for modelling of mixed type I-II ELMy H-mode, which has been implemented in JETTO. In predictive transport simulations with JETTO, the approach qualitatively reproduces the experimental dynamics of mixed type I-II ELMy H-mode, as illustrated in Figure 1, which shows the characteristic signature with small and frequent type II ELMs interrupted by occasional large type I ELMs. The approach has been used in studies on why some special effects and situations such as strong gas puffing, a quasi double null magnetic configuration (a configuration close to having a second singularity), high poloidal beta (ratio of total pressure to the magnetic pressure) and high edge safety factor (a measure depending on the ratio of the toroidal field to the poloidal field) and high triangularity (plasma shaping) can be favourable for mixed type I-II and pure type II ELMy H-mode. By performing MHD stability analysis on interpretative and predictive JETTO simulations, it has been shown that these situations lead to a strong increase in magnetic shear at the very edge of the plasma, which can cause this outermost region to become high n ballooning unstable, thereby effectively returning the operational point back to the first ballooning stability region. The result is important, because it improves the understanding of how to achieve operational regimes with more benign ELMs. Figure 1. Ion thermal conductivity as a function of time in a typical simulation with the model for mixed type I-II ELMy H-mode. 15

26 Predictive transport modelling of type I ELMy H-mode with theory-motivated ELM models: One problem with many ELM modelling approaches used in predictive transport simulations (such as the approach for modelling of mixed type I-II ELMy H-mode described above) is that they involve a large number of ad hoc parameters, which have to be specified. Usually it is difficult to set these parameters in a self-consistent manner. The best one can do is often to try to match the energy loss per ELM with experiments. In a more self-consistent approach designed to overcome this problem in part, type I ELM dynamics have been studied in predictive transport simulations with theory-motivated models based on linear ballooning and peeling stability theory. The model has been coupled into the JETTO transport code. When modelling ballooning-type ELMs, only the first equation is solved and when modelling peeling-type ELMs only the second one. In a combined ballooning-peeling approach, both equations are solved and the resulting mode amplitudes are added up according to the third equation. Transport in the ETB is in each approach enhanced with Gaussian-shaped perturbations, the amplitudes of which scale linearly with the calculated mode amplitude. Simulations with the ballooning, peeling and combined modelling schemes qualitatively reproduce the experimental dynamics of type I Figure 2. Ion thermal conductivity as a function of time at the magnetic surface ρ = 0.95 just outside the top of the ETB in a typical JETTO simulation with the theory-motivated ballooning model j [10 Am ] 5 2 z Peeling unstable Ballooning unstable α Figure 3. The path traced in the αs operational space during one ELM cycle in a typical simulation with the combined ballooning-peeling model. Consecutive points in the trace are separated by 0.1 ms. 16

27 ELMy H-mode, as shown in Figure 2 for the pure ballooning approach. In particular, the simulations reproduce a type I ELM frequency that increases with the external heating power, as in experiments. It has been demonstrated that in the first place the onset of discrete oscillations is related to how the radial profiles of the transport coefficients are perturbed in the transport simulations and to how the pressure gradient evolves as a result of this. With the combined ballooning-peeling model, it has been demonstrated that the individual ELMs are usually driven by a combination of ballooning and peeling modes. Due to the fact that the current generally evolves more slowly than the pressure gradient, the combined ballooning-peeling mode ELMs are triggered by a violation of the ballooning stability criterion. The collapse of the pressure gradient induced by the ballooning phase of the ELM then leads to a violation of the peeling mode stability criterion and the ELM continues in a generally quite long peeling mode phase until the edge current density has been depleted to a stable level. The typical ballooning-peeling mode ELM cycle is illustrated in Figure 3. The significance of the results is that they improve the understanding of type I ELM dynamics, which is of critical importance, given that ELMs both limit confinement and cause large divertor heat loads. Integrated predictive transport modelling of ELM heat pulse propagation: An ELM occurring at the outer midplane of a tokamak results in first an electron heat pulse and later an ion heat pulse arriving first at the outer target and then at the inner target. In experiments at the JET tokamak, the propagation times of the ion heat pulse to the outer and inner targets have been measured to be about 100 μs and 300 μs, respectively. It is of interest to develop a model of the scrape-off layer (SOL) that can accurately describe this propagation, because this would among other things make it easier to predict divertor heat loads on future tokamaks such as ITER. The propagation of a heat pulse induced by an ELM localised at the outer midplane has been studied with the integrated core-edge transport code COCONUT, which is a self-consistent coupling of the 1.5D core transport code JETTO and the 2D edge transport code EDGE2D / NIMBUS. A heat pulse has been induced by increasing the perpendicular transport coefficients in the edge transport barrier on the 1D core grid radially uniformly and in the 2D SOL radially and poloidally non-uniformly for the duration of the ELM. Poloidally, the perpendicular transport enhancement is distributed as a narrow Gaussian function peaking at the outer midplane. Parallel transport is given by the 21-moment approximation and adjusted by kinetic calculations. It has been studied what assumptions about perpendicular and parallel transport in the SOL have to be made in order to reproduce the experimentally observed propagation times of the electron and ion heat pulses to the outer and inner targets as well the magnitude and distribution of the heat fluxes. Initial results indicate that reasonable propagation times can be obtained in simulations with relatively simple assumptions of the transport model. It has been shown that the relative amounts of ion heat going to the wall and the targets depend sensitively on the radial enhancement profiles of perpendicular transport, the parallel flux limiting factors and the density. It has also been demonstrated that ion-electron equipartition increases strongly with collisionality. It has been concluded that because of the strong sensitivity of the heat fluxes on the heat transmission coefficients, the fluid approach assuming temporally and spatially constant transmission factors is insufficient during the transient and has to be complemented by a kinetic approach. For this reason, the parallel heat flux limiting factors and sheath heat transmission coefficients have been determined in particle-in-cell simulations for relevant transient scenarios. Such first-principle simulations show that these parameters vary strongly as a function of time during the transient. The kinetic results have been parameterised and included in the fluid simulations, which makes it possible to obtain more accurate predictions. 17

28 Figure 4. Time-integrals of electron heat flux (blue curves) to both targets, ion heat flux (red curves) to both targets and volume-integrated ion-electron equipartition energy (greenish-yellowish curve) in three transient scenarios: low density (left-hand figure), intermediate density (centre figure) and high density (right-hand figure). An ELM of 100 μs duration is induced at the start of the 5 ms period under inspection Interpretive transport analysis of toroidal momentum transport in the ELMy H-mode scenario Momentum transport is the least known and least studied transport channel (as compared with the ion and electron and particle transport channels) in tokamak physics. However, as the toroidal rotation has a paramount importance in suppressing turbulence, affecting the instability thresholds, stabilising resistive wall modes etc., and therefore, understanding of momentum transport is crucial. The first goal is to understand the toroidal momentum transport on JET and compare it with other transport channels, in particular that of the ion heat transport. After these studies, one should able to better extrapolate how large toroidal rotation velocities are expected to be in the ITER core plasma. For the time being, it is usually assumed the ratio of the effective momentum diffusivity to the effective ion heat diffusivity, i.e. Prandtl number, is close to unity. In addition, the quasi-linear theory of ITG turbulence predicts that ratio. The Prandtl number can be calculated performing interpretive transport simulations with JETTO transport code with self-consistent power deposition profiles and torque densities calculated by the NBI code PENCIL. This ratio is shown in Figure 5 for 9 very high density H-mode discharges (red triangles, same as in Figure 1), 8 other H-mode pulses with T i =T e (black circles) and 25 other H-mode pulses with T i T e (blue circles). The effective diffusivities are averaged over the gradient region between r/a=0.4 and r/a=0.7. It is to note that the ratio χ φ /χ i = for all analysed H-mode discharges, and rather independent of the density. In addition, similar ratio χ φ /χ i = is found for L-mode and hybrid, eff (m /s) = i = 0.2 i i, eff (m /s) Figure 5. Effective momentum diffusivity versus effective ion heat diffusivity for 9 very high density H-mode discharges (red triangles), 8 H-mode pulses with T i =T e (black circles) and 25 H-mode pulses with T i T e (blue circles). 18

29 scenario discharges. This number is smaller than χ φ /χ i = 1, commonly used in ITER predictions and often appearing in the ITG theory. The database shows that while the core transport is smaller for v φ, the pedestal momentum confinement is worse than that of ion heat and in fact, the edge pedestal is weaker for momentum than for the ion temperature. In addition, as low density discharges tend to have stronger T i pedestal, the global energy confinement becomes better with decreasing density while the global momentum confinement does not Transport simulations of toroidal momentum transport An extensive transport modelling study has been performed for high and low density ELMy H-mode discharges, hybrid scenario discharges and L-mode discharges. A comparison of the toroidal velocity, ion and electron temperatures and ion heat diffusion coefficients between the experiment and predictions with the new version of the Weiland model as well as with GLF23 transport model is illustrated in Figure 6 for one high density high performance H-mode JET discharge no This level of agreement between the predictions of both models and experiment is rather good, and this is generally true for other discharges existing in the database as well. Both models also predict that the ratio of χ φ /χ i is around in the gradient region, and thus, the agreement in toroidal velocity is of the same order as the one of the ion temperature Predictive transport modelling of transient transport experiments Rapid increase of the heat transport with respect to the inverse temperature gradient length above some critical threshold is a phenomenon observed in several tokamaks. This so called profile stiffness or resilience strives to keep the profile in a marginally stable area. The implications are degradation of confinement and fusion perfor- c (m /s) T (kev) T (kev) V (km/s) 2 e i f r/a Figure 6. Experimental (black curves) and predicted by GLF23 (blue curves) and Weiland model (red curves) of v φ, (upper frame), T i (second frame) and T e (third frame) for pulse no Fourth frame shows χ φ and χ i predicted by the GLF23 (blue dashed and blue) and by the Weiland (red dashed and red), respectively. JG c 19

30 mance, issues that are critical for the next generation tokamaks like ITER. Therefore, it is important to determine which factors affect stiffness and how the detrimental effects can be circumvented. Since the propagation of a temperature perturbation depends on the transport properties, and consequently the stiffness of the plasma, transient transport studies offer an applicable method for studying profile stiffness. Both in the experiments and in the modelling work, transient transport has been approached with two different methods, one using modulated heating power wave form and Fourier transform, and the other one using cold pulses generated by e.g. laser ablation or shallow pellet injection. A series of JET discharges with modulated heating power and cold pulses were analysed. The aim of the experiments was to explore the dependence of electron stiffness, i.e. the growth rate of the electron heat transport with respect to the temperature gradient, on the repartition of heating between the electron and ion channels. The major difference between the discharges was the power of applied neutral beam (NBI) heating while the small amount of modulated ion cyclotron (ICRH) heating was the same. Consequently, the ratio of the electron to ion temperature T e /T i and the inverse ion temperature gradient scale length R/L Ti, both important factors for turbulent fluctuations and turbulence driven transport, were varied. The analyses were performed by means of predictive simulations using the semiempirical Bohm/gyroBohm and the fluid theorybased Weiland transport models, implemented in the JETTO transport code. The Weiland model was found to underestimate the steady state transport. Especially the ion temperature and the peakedness of the density profile exceeded the experimental values. The Fourier analysis results, though indicating underestimated perturbative transport, were competitive with the Critical Gradient Model (CGM) and slightly better than those of the Bohm/gyroBohm and the Gyro-Landau-Fluid (GLF32) models. However, the model was unable to predict the propagation of fast, moderate to high amplitude, cold pulses, which suggests that the response of the diffusion coefficient to a perturbation of the temperature gradient weakens with respect to the perturbation. The Bohm/gyroBohm model worked reasonably well in steady state simulations and for low amplitude cold pulses such as the one induced by laser ablation. The results of the Fourier analysis were quite similar for both studied transport models. Both underestimated transient transport by a factor of 2 3 under all circumstances. However, the change in the transient transport between different discharges was equal or larger than experimentally observed, even though T e /T i, which was earlier assumed to be the key factor determining the level of the stiffness, remained the same. Encouraged by the results it was suggested that the level of stiffness might be more sensitive to R/L Ti. In order to investigate further the role of R/L Ti on electron stiffness in the Weiland model, the NBI power was varied arbitrarily. As the NBI power was increased, the perturbative diffusion coefficient increased approximately by factor of 2 while the electron temperature changed only about 15%. The phase difference between the modulated heating power and electron temperature response and the amplitude of the response are presented in Figure 7. The slopes of the curves are inversely proportional to the perturbative Figure 7. Two simulations having different R/L Ti made with the Weiland model. The black curve corresponds to the original simulation of the pulse no and the cyan to a simulation with increased NBI heating. 20

31 transport coefficient. In addition, collaborative analyses done with the CGM showed that the R/L Ti dependence of the perturbative diffusion coefficient is indeed nearly linear. Verifying theoretical explanations for the observation and finding ways to utilise them are ongoing tasks Predictive transport modelling of hybrid scenario plasmas The hybrid scenario is the newest scenario of the three base tokamak scenarios (the other two being the ELMy H-mode and steady-state advanced scenario with ITBs). The main difference between the hybrid scenario and baseline ELMy H-mode scenario is that hybrid scenario plasmas are free from large sawteeth because of the different q-profile and thus, aims to avoid large NTMs triggered by sawteeth. This gives rise to pressure profile peaking and further higher fusion performance. On the other hand, it does not exhibit ITBs nor aim towards full steady-state operation, thus it is not called advanced tokamak scenario either. Being the newest scenario it is the least developed scenario among the three and thus, its physics and transport are least known and least investigated. However, the experimental results of the hybrid scenario are very promising, better in many aspects than those of ELMy H-mode and steadystate plasma scenarios. So far it is not clear whether the improvement in the performance in the hybrid scenario originates from the core or the edge, whether the confinement is improved and what causes the improvements. Experimentally, it has been verified that the role of magnetic shear or the q-profile is significant, and as a consequence, real-time control of the q-profile is highly desirable. The simulations of the hybrid scenario discharges with the Weiland and GLF23 transport models have shown limited agreement on a scan over several JET hybrid scenario discharges with varying density. The predicted electron and ion temperature profiles are compared with the experimental ones in Figure 8. For most of the discharges, either electron or ion temperature was well reproduced (roughly within measurement accuracy) while the other heat transport channel showed only limited agreement with the experiment. No systematic reason for this has been found so far. Moreover, the simulations have indicated some difficulties in predicting the experimental q-profile. JETTO transport code coupled with neoclassical transport code NCLASS predicts often central q below one while the q-profile in hybrid scenario experiments stay above one. This is most probably due to the lack of MHD phenomena, such as fishbones and NTMs redistributing the current in the JETTO transport code. However, the predicted radial region with q<1 is less than r/a<0.2. Therefore, it does not cause any sig- Figure 8. The predicted ion temperature profiles (left frame) and electron temperature profiles (right frame) by GLF23 (solid blue) and by the Weiland model (dashed red) compared with the experimental ones (dotted green). 21

32 nificant problems in the interpretation of the simulation results, as the very central region is outside the so-called gradient region (0.2<r/a< 0.8) that is the region of most interest and importance from the transport point of view. The predictions of the hybrid scenario are similar to those in the ELMy H-mode scenarios and as a consequence, the transport models predict the experimentally observed fact that the core confinement is rather similar in hybrid and ELMy H-mode plasmas Predictive transport modelling of advanced tokamak scenarios with internal transport barriers (ITBs) The accuracy of predictive transport modelling in plasmas with Internal Transport Barriers (ITBs) is not as good as that in the standard ELMy H-mode scenario. That is particularly evident in time-dependent transport simulations where the discharge evolves from the early preheating phase towards the onset of the ITB while the q-profile is changing all the time. In addition, the other plasma profiles (density, ion and electron temperature and rotation) can change dramatically when turbulence becomes suppressed and/or heating power and momentum injection is increased. The main aim in this study has been to test semi-empirical and first-principle transport models in plasmas with internal transport barriers. So far, only empirical models have been able to reproduce the time dynamics of ITBs satisfactorily. In order to obtain a consistent picture over a large plasma parameter and geometry regimes, a multi-machine experimental tokamak ITB database, called the International Tokamak Physics Activities (ITPA) ITB database, has been employed. The emphasis has been on the ITB formation and dynamics, in particular investigating the timing of the onset and the radial location of the ITB. As a consequence, the time-dependent transport simulations carried out in this work are very different from those where only the steady-state phase of the discharges are simulated. The question of the ITB formation and dynamics is assessed with fully predictive transport modelling. By fully predictive transport modelling we mean that five transport equations (electron and ion heat, q, density and toroidal rotation) are solved. In order to obtain the most realistic and consistent understanding of the ITB behaviour, it is very important to predict also the density and toroidal rotation which are often taken from the experiments. A range of plasma conditions is examined for JET, JT-60U and DIII-D discharges with ITBs. Three pairs of high performance discharges from JET, JT-60U and DIII-D are simulated with the Bohm/GyroBohm, Weiland and GLF23 transport models using the JETTO transport code. One of the discharges in each pair has a low positive or zero magnetic shear (monotonic or flat q) whereas the other one has a negative magnetic shear (reversed q). Although the ITB formation mechanisms are not known precisely, it is clear that the role of magnetic shear is significant, justifying the selection of discharges with different shear. The ITB formation in the original semi-empirical Bohm/GyroBohm transport model is based on turbulence suppression by the combined effects of the magnetic shear and ω E B flow shear. In this study, the effect of Shafranov shift, so-called (α-stabilisation), has been added in the ITB formation threshold condition. The Weiland transport model includes, in addition to those mechanisms mentioned above, also turbulence suppression by the dilution effects, i.e. plasma impurities or Z eff, density peaking and the effects of plasma geometry, such as elongation. GLF23 includes besides those also the effect of the dilution due to fast particles from NBI. In general with the Bohm/GyroBohm model, the agreement in all transport channels with respect to the onset and radial location of the ITB between the experiments and transport simulations was good in JET and JT-60U, but not as good in DIII-D. This suggests that the mechanisms that govern the physics of the ITB may be different in DIII-D from those in JET and JT-60U, where the combined effect of the magnetic shear and E B flow shear seemed to be enough to explain the behaviour of ITBs. The simulation predictions are shown in Figure 9 where the experimental and simulated ion and electron temperatures, electron density, toroidal rotation and the q-profile for the two of the six simulated discharges are illustrated 22

33 (two JET discharges shown). In order to reproduce ITBs within the same good accuracy also in DIII-D, the α-stabilisation has to be included into the model and indeed, the role of α-stabilisation was verified to be significant in the simulations. In conclusion, having modelled tokamaks with different sizes demonstrated the significant role played by the α-stabilisation in governing the physics of the ITBs. The Weiland model does not predict the time evolution of the ITB plasmas satisfactorily in the multi-tokamak database. It did not predict a clear ITB for any of the simulated discharges. One of the main reasons for the unsatisfactory performance of the Weiland model seemed to be the oversize growth rates of the unstable modes (ITG, TEM,...) in these ITB plasmas, calculated by the model. It seems plausible that using the Waltz rule with the Weiland model with a constant multiplier is too simplified an approach when performing fully predictive, time-dependent transport simulations. Fixing the density, toroidal rotation and the q-profile to their experimental values did not change the predictions for the temperatures as compared to the predictions with the standard version, at any instant of the simulation between the preheating phase (low temperatures) and the high performance phase (high temperatures). As a consequence, the Figure 9. Profiles of the ion temperature (a) and (b), electron temperature (c) and (d), electron density (e) and (f), toroidal rotation (g) and (h) and q (i) and (j) for JET discharges no at t=6.0s (left-hand side) and at t=12.0s (right-hand side). The solid lines correspond to the experimental data and the dashed, dash-dotted and dotted ones to the predictions with the Bohm/ GyroBohm, Weiland and GLF23 transport models, respectively. 23

34 Weiland model seems to have only a weak dependence on the magnetic shear and q-profile and thus, the present versions of the model cannot be regarded as a satisfactory tool in investigating or developing advanced tokamak scenarios with ITBs, for example in view of ITER. The GLF23 model predicted the existence of an ITB (in five of the six simulated discharges from the ITPA database), but the footpoint of the ITB was located too far inside the plasma. In some cases it predicted the ITB in the ion channel only, even though an ITB exists also in the electron channel in the experiment (DIII-D 95989). However, the E B shearing rate is usually close to quenching the turbulence at the correct radial location with respect to the ITB in the experiment. This shows the sensitivity of the ITB prediction in this model. In spite of the GLF23 model having been shown to be rather accurate in single time slice stationary ITB simulations, our results show that its predictions are not robust enough in time-dependent simulations reproducing the ITB dynamics. However, GLF23 features in general better predictions for the temperature profiles outside the ITB than the other models used here. The key question to be raised is how reliably we can predict the behaviour of the ITB plasmas in future devices, for example in ITER. The semi-empirical Bohm/GyroBohm model with its ITB formation threshold condition was derived empirically from JET ITB plasmas. Although it works very well in JET and in a similar size tokamak JT-60U, and also in a smaller size tokamak DIII-D when including the α-stabilisation, it does not prove that the same modelling capability and accuracy can be extrapolated to much larger size tokamaks. On the other hand, the predictions with the first-principle transport models, the Weiland and GLF23 models, are not in a satisfactory agreement even with the experimental results from the present tokamaks. In all these simulations, like in most other transport simulations, the poloidal velocity is assumed to be neo-classical. However, as reported in section , the poloidal velocity is not neo-classical within the ITB and consequently, the estimation for the E B shearing rate, which is affected by the poloidal velocity, may not be a correct one Predicting the ITB dynamics with the measured values of the poloidal rotation velocity on JET Recent results from the measurements of carbon poloidal rotation velocity v θ across internal transport barriers on JET show that the velocities are typically an order of magnitude higher than the neo-classical predictions. As a consequence, the radial electric field can be very different from that calculated using the neo-classical value of v θ. This gives further rise to different E B shearing rates, than normally used in transport simulations to predict the ITB dynamics, location and strength. The 1D first-principle transport models, such as the Weiland model or GLF23, have so far had difficulties in reproducing satisfactorily the time dynamics, location and strength of the ITBs. The Weiland model does not typically predict a clear ITB at all while GLF23 often predicts the ITB at the wrong radial location or too weak an ITB. One of the obvious reasons is that the growth rates of the ITG/ TEM modes significantly exceed the E B shearing rates calculated from the radial electric field E r.in the present calculation of E r in transport codes, the neo-classical value for the poloidal rotation velocity is assumed. However, after the recent measurements of v φ on JET, the question to be addressed is whether the failure to predict ITBs could actually be caused by the incorrectly estimated E B shearing rates, rather than the oversize growth rates. Two predictive simulations with the Weiland transport model are compared in Figure 10a before the ITB formation (left frame) and after the ITB formation (right frame) Figure 10b. The only difference between the two simulations is that the first one (red curves) uses the neo-classical poloidal velocity from NCLASS whereas the second one (blue curves) takes the experimentally measured v θ in the calculation of E r and E B flow shearing rate. In the case when the experimental poloidal rotation is used, the Weiland model predicts the ion ITB just at the right radial location and right instant with roughly the same ITB strength as measured in the experiments. On the other hand, otherwise an identical simulation except with v θ from NCLASS does not exhibit any sign of an ITB. 24

35 5 5 exb (s ) E (kv/m) T (kev) T (kev) -1 r e i (10 ) JET Pulse No: Experimental Weiland with experimental Vpol Weiland with Vpol from NCLASS t = 5.7s R (m) exb (s ) E (kv/m) T (kev) T (kev) -1 r e i (10 ) JET Pulse No: Experimental Weiland with experimental Vpol Weiland with Vpol from NCLASS t = 6.2s R (m) Figure 10. Predictions for the ion and electron temperatures, radial electric field and E B shearing rate before the ion ITB formation (a, left frame) and after it (b, right frame) using the Weiland transport model. If the experimental poloidal velocity is used instead of the neo-classical one to calculate the radial electric field and the E B shearing rate, E r and E B are found to be even qualitatively significantly different. This is most pronounced within the ITB layer. As a consequence, the simulation predictions for the dynamics, strength and location of ITBs change and may improve significantly, as shown here in the case of the Weiland model. In addition to changing the predictive transport simulation results, the non-neo-classical anomalous poloidal rotation velocities might raise the need for further assessment of the neo-classical transport theory in the presence of turbulence. Furthermore, understanding the causality between the onset of the ITB and large v θ as well as the source for this large v θ remain extremely interesting future challenges Effect of toroidal ripple on ion transport The orbit-following Monte Carlo code ASCOT has been used to study the effect of toroidal ripple on the transport of thermal ions in JET and JT-60U magnetic configurations. The method uses ion seeding on one specific magnetic surface and allows analysis of the heat and particle pulse across the selected magnetic surface. Figure 11 shows a characteristic result of this approach. In the figure the ion thermal transport coefficients are plotted as a function of the normalised toroidal flux. The same background density, temperature and current profiles were used in the simulation, but two different levels of magnetic ripple were assumed for JET TF coil geometry, and the JT-60U TF coil geometry was simulated with and without recently installed ferritic insets (FST). One can conclude that by increasing the ripple amplitude the ion heat transport can significantly exceed the level of the neo-classical transport. It is also seen that the same level of magnetic ripple at the outer midplane generates more transport for the JT-60U magnetic configurations than for the JET one. ASCOT simulations also reveal that, depending on the plasma collisionality, thermal ions can have both diffusive and convective losses. ASCOT also allows calculating, for varying levels of the toroidal ripple, the radial profile of the radial electric field generated by ion orbit losses. Figure 12 shows an example of such a distribution for the JT-60U configuration. Clearly the ripple losses can increase the radial electric field amplitude significantly above the level generated 25

36 0.3 Solid: (m /s), Dash (1/300 1/sec) no ripple 2.2T E s withfsts 2.2T E s wofsts 2.2T jet 2.2T JT-60U ripple jet 2.2T Full ripple Rho Figure 11. The ion heat diffusion coefficient and collisionality as a function of minor radius as evaluated from ASCOT simulations. The simulations were carried out for JET and JT-60U geometries and ripple strengths, and the effect of ferritic inserts on JT-60U was also isolated. x Er (V/m) no ripple JT-60U with FST JT-60U w/out FST RHO Figure 12. The ASCOT-calculated radial electric field profile for JT-60U taking into account the effect of toroidal ripple and ferritic inserts. 26

37 by the first orbit losses. It is worth noting that both the overall level of the radial electric field and its dependence on the ripple amplitude depend sensitively on the assumption of how strong is the ion viscosity that balances the radial current induced by orbit and ripple losses. We conclude that with FST the JT-60U coils do not enhance the electric field. This translates into reduction of toroidal rotation into counter-current direction, which indeed was recently reported by JT-60U. 2.2 MHD Stability and Plasma Control Edge localised modes (ELMs) Stability analysis of JET ELMs: The high-confinement-mode (H-mode) in present day tokamaks is regularly accompanied by short bursts of plasma energy and particles. These edge localized modes (ELMs) create high heat loads on the divertors and can cause significant erosion. For an estimate of this erosion in a future tokamak reactor operating in H-mode, it is necessary to understand the ELM phenomenon. This would also allow to analyse the necessity and usability of actual techniques of ELM mitigation and control in such a device. The main issues in the control of the ELMs are the size of a single ELM crash and the confinement degradation caused by the ELMs. In JET the Type I ELMy plasmas have good confinement, but produce large ELM crashes, while Type III ELMy plasmas have small crashes, but the confinement is degraded. The triggering mechanism of the different ELM types is studied using MHD stability analysis with the JETTO transport code, HELENA equilibrium code and MISHKA-1 stability code. JETTO is used to calculate the current profile in the plasma using the experimental density and temperature profiles. The current profile as well as the experimental pressure profile is used in shear #55936 =1.0x10 1/s 22 0.#55937 =1.7x10 1/s 22 0.#55953 =10.7x10 1/s Figure 1. Stability analysis result of a gas scan in JET. The x-axis is the normalised pressure gradient at the edge and y-axis the magnetic shear (~1/current density). The numbers represent the toroidal mode number of the most unstable mode and the red crosses show the unstable region of n= ballooning mode. The coloured markers indicate where the experimental plasmas are located in α-shear space. 27

38 Figure 2. The radial mode structure of n=10 peeling-ballooning mode. HELENA to solve the equilibrium. MISHKA-1 code calculates the stability of a given equilibrium. In the stability analysis, first the experimental equilibrium is solved. Then the parameters in the edge region, current density and pressure gradient, are varied to find the stability boundaries around the experimental point. The stability boundaries can explain which modes are responsible for triggering the ELMs. Figure 1 shows the analysis of a gas scan in JET. There are three plasma discharges with identical shapes and global parameters (plasma current, magnetic field, heating), but varying gas fuelling. Two of the experimental plasmas (#55935, #55937) have Type I ELMs while one (#55953) has Type III ELMs. The Type I ELMy plasmas are at the intermediate-n peeling-ballooning mode (radial mode structure shown in Figure 2) stability boundary while the Type III ELMy plasma is deep in the stable region. We can conclude that the Type I ELMs are triggered by peeling-ballooning modes, but triggering mechanism of Type III ELMs is not an ideal instability Real-time control techniques to control q and the strength and location of the ITB in predictive transport simulations An important experimental programme is in progress on JET to investigate plasma control schemes which could eventually enable ITER to sustain steady-state burning plasmas in an advanced tokamak operation scenario. The triggering and subsequent controllability of ITBs are major issues for fulfilling this goal, and their study is therefore an essential part of this programme. Uncontrolled ITBs are generally not stationary, as often observed on JET, and the coupled evolution of the plasma parameter profiles in high performance non-inductive discharges often leads to the premature loss of the good confinement ITB regime, or alternatively to too large pressure profiles, with major MHD events, sudden barrier collapse and/or abnormal plasma termination. Recently, a multi-variable model-based technique was developed for the simultaneous control of the current, temperature and/or pressure profiles in 28

39 JET ITB discharges, using lower hybrid current drive (LHCD) together with NBI and ICRH. The Real-Time Control (RTC) algorithms have been implemented in the JET control system, allowing the use of three actuators that are the power levels of NBI, ICRH and LHCD systems. Identical algorithms to those used in the JET RTC experiments have been also implemented in the JETTO transport code. This work covers primarily the progress achieved in fully predictive transport modelling of ITB plasmas when applying the real-time control algorithms in the closed-loop transport simulations to control the q-profile and the strength and location of the ITB. This is the first time when predictive transport simulations with a non-linear plasma model (Bohm/GyroBohm transport model) have been used in closed-loop simulations to control the q-profile and the strength and location of the ITB characterised by Te * or Ti * (defined as the Larmor radius divided by the temperature gradient length). Five transport equations (for predicting T e, T i, q, n e,v φ ) are solved and the power levels of LHCD, NBI and ICRH are controlled by the feedback controller matrix and the difference between the set-point (target) and simulated values of q and Te*. In the closed-loop simulations with the real-time control, RTC can be applied as long as desired in the JETTO code, typically for more than one resistive current diffusion time. The power levels vary in the closed-loop simulations, as requested by the controller. The whole procedure of carrying out the open-loop power step-up simulations, determination of the controller matrix from the open-loop simulations and finally performing the closed-loop simulations with RTC is identical to that performed in JET experiments when applying the RTC technique. The benefits in the transport simulations with respect to experiments are that transport simulations are free from unpredictable events, such as MHD events, diagnostics problems or power systems failures occurring often in the experiments and polluting the data. In addition, RTC techniques can be tested for several current diffusion times which is impossible with the present capabilities of JET heating systems. Therefore, the simulations serve as a simplified platform to test, validate and develop the real-time control algorithms techniques, with increasing degrees of complexity and completeness. The modelling results obtained in the closed-loop simulations when applying the real-time control to the q-profile and electron ITB ( Te* ) are illustrated in Figure 3. In each frame, one open-loop and one closed-loop (i.e. with RTC) simulations are compared. In left frame, the set-point q-profile is monotonic and the set-point Te * -profile has small values, indicative of no ITB, and shown by the dashed green lines. In the right frame, the set-point q-profile is strongly reversed and strong ITB is requested (large value of Te * ). The corresponding closed-loop simulations are denoted by the red and blue solid curves in each frame, respectively. For a comparison, a reference openloop simulation with constant power levels is illustrated by the solid black curves in each frame. The fully predictive closed-loop simulations have demonstrated that the real-time control of q can be carried out successfully within the resistive current diffusion time which is beyond the experimental capabilities of JET tokamak. Within the limits of the present transport model, the simulation with the reversed q-profile and strong ITB as the set-point Te * -profile also showed the way to achieve strong ion ITBs, desirable for high fusion performance. As the ion temperature is a more relevant quantity for the fusion performance point of view than the electron temperature, a similar control algorithm for the strength and location of the ion ITB ( Ti* ) was implemented in JETTO. The closed-loop JETTO simulations with the simultaneous RTC of q and Ti* showed that it is significantly more challenging to control Ti * than Te*. Thus, RTC of Ti * is more difficult to achieve than that of Te *, indicating that the ion ITBs are more challenging to control than the electron ITBs. At least the following two reasons could be identified: firstly, the strength of the ion ITB (magnitude of Ti * )is controlled by NBI and ICRH while the location of the ion ITB (radial profile of Ti * ) is mainly controlled by LHCD via the LH driven current changing the q-profile. And secondly, the values 29

40 5 5 4 t=10s t=20s t=30s 4 t=10s t=20s t=30s q 3 q * * Figure 3. Left frame: q and ρ Te* for the closed-loop simulation (red) together with its set-point profiles (dashed green) at three instants. Right frame: q and ρ Te* for the closed-loop simulation (blue) together with its set-point profiles (dashed green.) The same open-loop reference simulation is shown by the black curves at three instants in both frames. of Ti * vary typically a factor 2-5 just outside and inside the ITB while those ones of Te * vary only a factor of 1.5. As a consequence, a larger non-linearity in the controlled T profile is created in the ion heat transport channel. These simulations are real predictions in a sense that the RTC of Ti * has never been done experimentally on JET so far Stochastization of magnetic fields and magnetic reconnection Stochastization of magnetic field lines in fusion machines is known to play a very important role in fast energy loss events from magnetically confined fusion plasmas due to MHD modes. Classical examples are sawtooth crashes and disruptions. Possibly, stochastization leaves its mark also on neoclassical tearing modes (NTM), which degrade the performance of tokamak plasma. Another prominent application is the ergodization of the edge region to influence the exhaust properties. The recent work was motivated by the important finding of a regime of NTMs in which three wave coupling between an (m,n) NTM, an (m+1,n+1) ideal mode and a central (1,1) mode lead to a regime in which NTMs only modestly degrade confinement and allow to reach high beta values at good confinement in the presence of NTMs. This so-called FIR regime, first found on the ASDEX Upgrade tokamak, thus has a high value for future reactor-class devices such as ITER. It has been conjectured that the three wave interaction leads to intermittent stochastization of the region around the NTM. This would lead to fast reconnection, thus preventing the NTM from growing to its saturated size. The mapping technique was developed and applied to trace the field lines of toroidally confined plasma where perturbation parameters are expressed in terms of experimental perturbation amplitudes determined from the ASDEX Upgrade tokamak. (Figure 4). It is found that fast reconnection observed during amplitude drops of the neoclassical tearing mode instability in the frequently interrupted regime indeed can be related to stochastization (Figure 5). It is also shown that stochastization can explain the fast loss of confinement during the minor disruption. By means of the mapping technique the diffusion coefficient is determined of stochastic field lines arising in fast reconnection phenomena in magnetized fusion plasma during the FIR-NTM in ASDEX Upgrade tokamak. The maximum values of the local field line diffusion coefficient are found of the order of 10-5 to 10-6 m 2 /m (Figure 6) which correspond to the electron 30

41 thermal diffusivity of the order of 100 to 200 m 2 /s. Solution of the non-stationary diffusion equation with variable diffusion coefficient predicts that the temperature profile during the FIR-NTM event is shifted towards the plasma boundary within 500 s which agrees well with experimental observations and explains the contradiction between the expected rather slow resistive MHD reconnection rate (few tens of milliseconds in the ASDEX Upgrade). Also non-complete sawtooth reconnection in ASDEX Upgrade tokamak was investigated. Such reconnection phenomena are associated with internal m/n=1/1 kink mode which does not vanish after the crash phase (as would be the case for complete reconnection). It is shown that this sawtooth can not be fully described by pure m/n=1/1 mode and that higher harmonics play an important role during the sawtooth crash phase. The Hamiltonian formalism was applied to reconstruct perturbations to model incomplete Sawtooth reconnection. It is demonstrated that stochastization appears due to excitation of low-order resonances which are present in the corresponding q-profiles inside the q = 1 surface which reflects the key role of the q 0 value. Depending on this value two completely different situations are possible for one and the same mode perturbations: (i) the resonant surfaces are present in q-profile leading to stochasticity and sawtooth crash (q ± 0.1); (ii) the resonant surfaces are not present which means no stochasticity in the system and no crash event (q ± 0.05). Accordingly central safety factor value is always less than unity in case of noncomplete sawtooth reconnection. Our investigations show that stochastic model agrees well with experimental observations and can be proposed as a promising candidate for explanation of the sawtooth reconnection. All this demonstrates that stochastization can be regarded as a possible cause for different MHD events in ASDEX Upgrade. Figure 4. Experimental perturbations for the (1,1), (3,2), (4,3) and (5,4) modes as a function of magnetic flux for the ASDEX Upgrade discharge #11681, t=2.98s. 31

42 Figure 5. Poincare plot corresponding to interaction of the (3,2) and (4,3) modes during the FIR-NTM regime. Shapes of the perturbations are shown in Figure 4. Here is the minor radius of the tokamak. The ASDEX Upgrade discharge No 11681, t=2.98 s. Figure 6. Local diffusion coefficient corresponding to the Poincare plot shown in Figure 4. 32

43 2.3 Plasma Heating and Current Drive Ion cyclotron heating experiments in JET Heating with waves in the ion cyclotron range of frequencies (ICRF) is one of the auxiliary heating methods planned for ITER. In addition to heating and burn control, ion cyclotron waves will be used to control plasma instabilities. In the following, ICRF heating physics experiments carried out in preparation of ITER on the JET tokamak are discussed. Physics of second harmonic ICRF heating: One of the principal ICRF heating schemes foreseen for ITER is second harmonic heating of tritium ω 2ω ct. For this heating scheme, the absorption of wave power at the ion cyclotron resonance is a finite Larmor radius (FLR) effect. Consequently, the wave absorption by the resonating ions is weak at low energies but increases strongly as the ratio of the ion Larmor radius, ρ=v /ω ci,to the perpendicular wavelength of the fast wave increases. However, when ion Larmor radius further increases (corresponding typically megaelectronvolt range) the absorption weakens again and effectively prevents particles from reaching higher velocities. To predict with confidence the performance of second harmonic heating of in ITER it is important to have a good understanding of the resonant ion energy distribution. Experiments have been performed on the JET tokamak with 2nd harmonic ICRF heating of hydrogen in deuterium plasmas to assess the role of FLR effects on the resonant ion distribution function. More specifically, the clamping of high-energy resonant particle distribution due to weak wave-particle interaction at high energy is studied. The distributions of ICRF heated hydrogen ions have been measured with a high-energy neutral particle analyser in the range of MeV. By changing the electron density the energy E*, around which the wave-particle interaction becomes weak, is varied. The dependence of the ion distribution on E* is experimentally observed for a number of discharges and FLR effects are clearly seen to affect the high energy tail shape. Experiments have been analysed with the combination of ICRF modelling codes PION and FIDO, including FLR effects, and good agreement with measurements have been found. Figure 1 shows the final results including error estimates for the simulated distributions Fh (kev -1 st m ) Energy, MeV Energy, MeV Energy, MeV Figure 1. High-energy parts of the resonant hydrogen distributions. Points with error bars are the neutral particle analyser measurements, full lines are from the FIDO simulations and dashed lines are the estimated errors for the simulations (due to uncertainty in plasma parameters). 33

44 Experiments on ICRF coupling with different phasings: High power density ICRF launchers will be needed for ITER and, after the end of the 2004 and 2005 campaigns on JET, an ITER-like ICRF launcher will be installed and tested in JET. The reference design of the ITER launcher assumes toroidal dipole phasing while better coupling performance could be obtained in toroidal monopole phasing. This conservative assumption is related to the earlier measurements of plasma energy content vs. input power at JET, where practically no hydrogen minority heating was observed for monopole with the present A2 antennae although the coupling resistance was better in monopole than in dipole. To further investigate the phasing dependence of heating efficiency with the JET A2 antennas, L-mode coupling experiments were conducted varying input power and its modulation, antenna-plasma distance, plasma configuration, number of active antennas, and phasing. Figure 2 shows a comparison between different antenna phasings. Monopole phasing is seen to heat up the plasma at only about half of the efficiency of dipole. The above results indicate that the parallel wave number spectrum radiated by the antenna plays an important role in the plasma core heating efficiency. This may support the model of parasitic ICRF absorption by the near or far rf sheath voltage rectification. This is further supported by the fact that the part of the ICRF coupled power, which is not seen to be absorbed in the core, is not detected in the measured radiation or divertor heat loss channels. In spite of large measurement inaccuracies the missing power seems to be systematically large for low ICRF heating efficiency and vice versa, in accordance with earlier findings from thermocouple measurements at JET. Physics of polychromatic ICRF heating: The distribution function of ions resonant with ICRF waves is often driven towards a non-maxwellian state with an anisotropic tail forming in the high-energy part of the distribution function. Fast tail ions play an important role for the resulting plasma heating and affect the stability properties of the plasma. Therefore, it is important to have a number of ways to modify the ICRF-driven fast ion distribution function. The most commonly 8.0 a) b) ICRF coupled power ICRF coupled power MJ MW NBI Monopole Diamagnetic plasma energy Dipole NBI Diamagnetic plasma energy Dipole 1.5 Monopole time, s time, s Figure 2. Evolution of coupled power and plasma energy for (a) discharges and and (b) discharges and in a power ramp-up to 8 MW. 34

45 used techniques include varying the resonance position and resonant ion density, as well as inducing radial transport of the resonant ions with toroidally directed waves. In recent JET experiments, further possibilities to influence fast ICRF-driven populations have been investigated using multi-frequency (polychromatic) heating, as opposed to the more common single-frequency (monochromatic) heating. Experimental information on the fast ion populations is provided by two-dimensional gamma-ray emission tomography and the measurements are compared with numerical modelling (Figure 3). Due to its lower peak power density and thereby lower fast ion energy, polychromatic heating has potential for the key task of controlling and optimizing the ion heating profile in present-day and future tokamak plasmas, with the radial spread being controlled by the frequency shifts of the different ICRF antennae. ICRF heating of trace tritium on JET: The JET Trace Tritium experimental (TTE) campaign in 2003 provided a rare opportunity to study ICRF heating of tritium (T) at low concentrations in deuterium plasmas. Heating of T minority ions at their fundamental cyclotron frequency is an attractive though technically challenging heating scenario, which is currently outside the ITER ICRF system frequency range. On JET, it requires the highest equilibrium magnetic fields (3.9 to 4T) and the lowest available generator fre- Figure 3. Gamma-ray emission for polychromatic and monochromatic ICRF heating of 3 He. Polychromatic heating in discharge with resonances in the plasma centre and on the low magnetic-field side (LFS) produces predominantly high-energy standard trapped ions, while monochromatic heating in discharge with a central resonance produces stronger gamma-ray emission with the maximum emission in the midplane close to, and on the LFS of, the resonance, in agreement with the calculated radial distribution of fast ion orbits. The gamma-ray emission is mainly due to nuclear reactions between fast 3 He ions with energies above a threshold energy of about 0.9MeV, and 9 Be, which is one of the impurity ion species in JET plasmas. 35

46 quency (23 MHz), at which only modest levels of ICRF power are available. In TTE, tritium was introduced either by gas puffs of about 5 mg per discharge, or by beam injection (about 0.2 mg in 300 ms). Although tritium increments per shot were small, after a sequence of discharges tritium concentration could be built up to levels of about 1 %. ICRF powers of 1 to 1.5 MW were coupled, producing energetic tails in the triton distribution with effective temperatures between 80 and 120 kev, as derived from the neutron emission spectroscopy data. Such energies are close to the maximum of the deuterium-tritium fusion reaction rate. Increases in the suprathermal neutron emission by about three orders of magnitude were accordingly observed during the ICRF pulses (up to /s with gas puff, and /s with beam injection). The neutron emission profiles show an emission peak a few centimeters on the low field side of the T cyclotron layer, consistent with fast trapped or non-standard triton orbits grazing the latter. Comparison was made between non-directive and directive phasing (i.e., dipole, +90º and 90º phasing) of the antenna arrays, which exhibited differences in neutron emission and evidence of opposite fast ion toroidal rotation. Discharges were also devoted to accelerating tritium at its second cyclotron harmonic, yielding fast tritons above 700 kev (deduced from gamma ray spectra). Investigation of neutron production in hydrogen minority ICRF heated trace tritium plasmas: A neutron-producing nuclear reaction can take place between tritons and energetic protons with a large cross-section above 1 MeV. A systematic study of neutron production due to this reaction was carried out with purely ICRF-heated plasmas in the JET TTE campaign. Proton-tritium neutrons are important for interpreting neutron diagnostics data properly in purely ICRFheated plasmas containing tritium, as it causes discrepancy between broad-energy total neutron yield measurement and narrow energy-band deuterium-deuterium (DD) and deuterium-tritium (DT) neutron measurements. It is also interesting, because the pt neutron yield depends strongly on the effective proton tail temperature above 1 MeV. However, measuring pt neutron yield is challenging in JET because a broad, predominantly low-energy neutron spectrum is produced. It can only be carried out through measur- Excess neutron fraction [%] NBI only DD shots NBI only T blips NBI only T puff PT before NBI pt whole ICRH E+14 1.E+15 1.E+16 1.E+17 Total neutron yield Y [1/pulse] n Figure 4. Excess neutrons in a number of discharges in the JET Trace Tritium experimental campaign. The points indicate excess as percentage of total measured neutron production. The excess is highest in the experiments aimed at producing pt neutrons. 36

47 ing the neutron excess, i.e., total yield minus the contribution from DD and DT fusions (Figure 4). In the experiment, the neutron excess was measured while varying tritium input and ICRF power deposition in plasma. In otherwise identical pulses, the excess was found to increase monotonically with tritium input, which is strong evidence of inducing pt neutrons in plasma. Less excess was observed with broader ICRF power deposition profile leading to less energetic proton tail. Some neutron excess was also observed with no tritium input, suggesting that other, not tritium-related processes may also be causing excess neutron production. This was studied further in a recent experiment, where reactions with beryllium impurity ions were studied. Highly energetic proton tail was created, escaping protons up to 5 MeV were detected by lost ion measurements. Preliminary analysis displays an neutron excess during high power RF heating but further analysis is needed to establish if this can be correlated with beryllium density Lower hybrid current drive experiments at JET Lower hybrid (LH) waves in the frequency range 1 10 GHz are used to heat and drive the non-inductive current in tokamak plasmas. A non-inductive current drive is necessary for steady-state operation in tokamaks. LH waves are the most efficient method of driving off-axis current and thus modifying the current profile for improved plasma confinement.. Waveguide grills are used for wave excitation in the plasma. LH coupling studies: A critical issue for the use of LH waves is coupling of waves from the grill to the plasma. Lower hybrid waves have a cut-off density below which they do not propagate. In ITER, the plasma-wall distance will be about 15 cm and the density at the edge drops very fast. Therefore, it has been essential to demonstrate already on current machines that it is possible to couple LH waves over a long distance. The coupling of LH waves was studied at JET in two different plasma scenarios, an advanced tokamak scenario and a hybrid scenario. Both of these were ITER-like scenarios and at high plasma triangularity. In order to overcome the low-density region in front of the LH grill at large plasma-wall distances, gas puffing from a nearby gas injection module was used. The gas flows from the gas pipe and propagates to the region in front of the grill. In the meantime it is ionised and consequently increases the density in front of the grill to a level at which the LH waves can propagate into the plasma. In these experiments, very good coupling was obtained. In the hybrid scenario, up to 3 MW of LH power was coupled over a plasma launcher distance of almost 14 cm. in the ITER-like advanced scenario, slightly over 3 MW was coupled with actually less gas than in the hybrid scenario. Moreover, in this case the plasma launcher distance was larger, about 15 cm like it will be in ITER. Without gas from the nearby gas pipe the coupling was lost. Hot or bright spots induced by LH power: Part of the coupled LH power is lost in the very edge of the plasma, just in front of the LH grill. A fraction of the power is absorbed parasitically by electrons. This fast electron generation at the grill mouth may limit the lower hybrid (LH) power at high power densities, since the fast particle beam may cause impurity influx from hot spots on the wall structures. Such hot spots have been observed in several current drive experiments on components that are magnetically connected to the grill region. This is especially inconvenient in long-pulse discharges at high power since the hot spots and the related impurity influx limit the power level of the grill. In JET experiments, series of bright spots were detected with the CCD camera on the inner and outer divertor apron, which are magnetically connected to the LH grill region. The LH power level in these experiments was between P LH =1 and 3 MW and strong gas injection near the grill was used to improve the coupling. However, in some cases excess of gas in front of the grill may increase the heat flux to the magnetically connected components. The Infra Red Movie Analyser, IRMA software was used to analyse the CCD videos of the pulses that showed hot spots on the 37

48 Safety factor q Q 95 Pulse #58668 inner apron 3 2 z1 1 z2 z Time, t [s] Brightness of spots, [a.u.] Figure 5. The left-hand side image shows the measuring points in the IRMA analysis, which is shown on the right-hand side for shot The colours of the lines indicate the measuring points denoted by the coloured circles on the left-hand side. The right-hand side image also shows the safety factor versus time. During the first phase of the pulse, from t=44 to 48 s, the brightness is higher because of the ELMy phase. divertor apron. Figure 5 shows the measuring points and the result of the analysis for a JET pulse. The analysis shows clear increases in the brightness of the measuring points in the second phase of the shots, denoted by the arrows. The brightness was analysed as a function of various parameters. In addition to the LH power, an strong inverse dependence on the plasma-wall distance was observed. Recently an infra-red camera has been installed on JET, which enables a more detailed study of the hot spots and also gives information on the wall temperature. A clear increase of the temperature with LH power is seen on the limiter. Moreover, the increase is faster with higher LH powers Gyrotron theory High-power high-frequency gyrotrons play an essential role as microwave sources for plasma heating and current drive in modern magnetic fusion machines. As the size and performance of experimental devices is increased towards a commercial reactor, also the requirements for frequency and unit output power of gyrotrons tighten. It is therefore important to know the operational limits of such gyrotrons. One significant group of limitations may arise from chaos, whose onset becomes more likely with increasing output power. At TKK, chaos in gyrotrons has been investigated for several years. Recently, these studies were completed by including in the analysis the effects of electron beam misalignment and microwave reflections. Also hysteresis effects have been investigated using the gyrotron theory. The two-dimensional self-consistent time-dependent theory of beam-wave interaction in gyrotron resonators was modified to account for eccentricity of the electron beam. It was known from earlier studies that there exists a critical value of the azimuthal index of the mode (which depends on the resonator dimensions) above which stationary single-mode operation becomes impossible. Numerical analysis shows that in- 38

49 dcrit m Figure 6. Maximum tolerable electron beam displacement relative to the wavelength as a function of the azimuthal index of the mode. The operating parameters of the gyrotron are such that the output efficiency reaches its optimal value. creasing misalignment tends to lower this threshold. Figure 6 shows the maximum tolerable beam misalignment as a function of the azimuthal mode number. It is seen that for mode number exceeding 40, the gyrotron is very sensitive to misalignment. Together with high-order mode operation it should therefore always be considered how precisely the beam can be directed into its correct position in the resonator. To describe reflections of microwave power from plasma, transmission lines, or other components back to the resonator, we have developed an extension of the one-dimensional self-consistent time-dependent theory of nonstationary processes in gyrotrons. Different mathematical descriptions of partial reflection of the output signal were compared (see Figure 7), and numerical algorithms for analyzing them were given. Using a novel description, we computed a map of gyrotron oscillations, which identifies the regimes of stationary, periodically modulated and chaotic oscillations in the plane of generalized gyrotron variables when reflection is present. Our findings can therefore be exploited in the development of high-power gyrotrons, which should provide a stationary signal even in the case of accidental reflections. By identifying in the operating parameter plane those regions where chaotic oscillations may be obtained, the results also ease the design of gyrotrons for applications which require broad bandwidth. Figure 7. Illustration of different ways to describe reflections in gyrotrons. (a) In the reflectionless case the whole electromagnetic wave exits the resonator. (b) Reflections have been modeled in a previous study with the help of an iris located at a distance comparable to the length of the interaction space. (c) A reflective boundary condition can be applied at the resonator end, using the phase of the reflection coefficient to simulate delay. (d) The iris can be replaced by the reflective boundary condition at the same location. (e) The reflected wave can be actively launched after a delay. 39

50 2.3.4 Gyrotron development for ITER The development of high-power high-frequency gyrotrons is strongly driven by the needs of fusion technology. Gyrotrons are superior to other radio-frequency (rf) sources in the frequency range relevant for electron cyclotron resonance heating (ECRH), or about 170 GHz for ITER. To make an ECRH system cost-effective, the output power of a single gyrotron should be around 2 MW continuous power. Coaxial cavity gyrotrons have the potential to fulfil this requirement as has been experimentally demonstrated within the development program performed as an ITER task at Forschungszentrum Karlsruhe (FZK). In proof of principle experiments carried out at FZK Karlsruhe on a 165 GHz coaxial cavity gyrotron during the last years, the feasibility of manufacturing a 2 MW, CW coaxial gyrotron at 170 GHz has been demonstrated and information necessary for a technical design has been obtained. Based on these results and on the experience acquired during the development of the 1MW, CW, 140 GHz gyrotron for W7-X, the technical feasibility of a 2 MW, CW, 170 GHz coaxial cavity gyrotron has been studied before EFDA has placed a contract with Thales Electron Devices (TED) for procurement of a first industrial prototype of such a coaxial gyrotron tube. The development work is done in cooperation between European research centers together with TED, the main European tube manufacturer. Within this cooperation the physical specifications and the design of gyrotron components have been done by the research institutions whereas TED is responsible for the technological aspects and manufacturing of the tube. Figure 8. Schematic view and a photography of the completed prototype gyrotron at FZK. 40

51 In the meantime the fabrication of the prototype gyrotron has been finished and the tube will be delivered to CRPP Lausanne where the tests will be performed (Figure 8). A suitable superconducting magnet is expected to be available in spring Therefore experimental operation of the gyrotron experiments could start in summer The Helsinki University of Technology has participated in this development. In parallel many theoretical investigations have been performed. Most recently, the effect of microwave reflections in gyrotrons with radial output and consequences for the ITER coaxial gyrotron was studied, azimuthal instability in gyrotrons with overmoded resonators was investigated, feasibility of coaxial super power (4 MW) was examined, Hamiltonian map description of electron dynamics in gyrotrons was proposed, eigenvalues and ohmic losses in coaxial gyrotron cavity were reexamined by means of a novel method. The past, present and future of coaxial gyrotrons has been reviewed in a special invited paper. 2.4 Energetic Particle Physics Confinement of and heating by fusion alphas in standard H-mode and advanced scenario discharges in JET The successful operation of a fusion reactor relies on sufficient confinement of the fusion-born alpha particles. In order for the fusion reactor to be economically feasible, the alpha particles must be confined well enough to heat the plasma by transferring nearly all of their energy to the plasma before escaping from the reactor. The widths of the 3.5 MeV fusion alpha orbits can be of the order of the size of the plasma. Furthermore, they are inversely proportional to the strength of the poloidal magnetic field. Since the poloidal magnetic field is created by the toroidal plasma current, the alpha particle confinement becomes of particular concern in advanced scenario plasmas, where the so-called internal transport barrier (ITB) is typically created by a low or vanishing toroidal current in the plasma centre. Thus, while the ITB improves the confinement of the thermal plasma, it might be deleterious on alpha particle confinement. The purpose of this work was to compare the behaviour of fusion-born alpha particles in H-mode and advanced scenario plasmas, by simulating the trajectories with the Monte Carlo guiding-centre-orbit-following code ASCOT. The emphasis of this work was on the alpha particles ability to heat the plasma and on the alpha power lost to the walls of the device. Initial simulations in a circularly symmetric geometry and with analytic temperature and density profiles gave promising results indicating that, in ITB plasmas, the alpha confinement is not significantly compromised. The overall alpha heating was found to be of the same order of magnitude in both an ITB and an H-mode plasma, but in the former it was spread over a larger volume. Subsequently simulations were performed with authentic magnetic geometries, and temperature and density profiles of JET shots. The results are shown in Figure 1. In advanced scenario plasma (#51976), the heating by alpha particles was again found to cover a larger volume than in an H-mode plasma (#52009). In addition, the non-standard flux surface structure of the advanced scenario plasma led to interesting fine structure in the power deposition pattern with hot spots and a cooler region close to the plasma core. In the H-mode plasma, on the other hand, the power deposition pattern was symmetric and concentrated close to the magnetic axis. Studying the alpha power lost to the walls of the vessel showed that alpha particle confinement is substantially worse in advanced scenario plasma than in an H-mode plasma. Alpha confinement in the advanced scenario plasma could probably be improved by increasing the plasma current, but it will, on the other hand, be further deteriorated if the effect of the magnetic ripple is taken into account. 41

52 Figure 1. The alpha power deposition pattern of the a) ITB shot #51976 with a current hole in it and b) normal H-mode shot # The deposited powers are given in watts Fast ion distribution in H- and QH -mode operation of AUG The standard H-mode operation is characterized by ELMs, violent bursts of energy and particles, which can lead to unacceptable loads onto plasma-facing components. On the other hand, ELMs play a crucial role in controlling density and exhausting impurities. Therefore finding means to moderate or eliminate ELMs, while keeping stationary density and impurity content, is of paramount importance, especially for fusion reactors such as ITER. In the so-called Quiescent H-mode (QH-mode) ELMs are replaced by more continuous, benign MHD behaviour called Edge Harmonic Oscillations (EHO) enabling stationary plasma density and radiation level. So far, the QH-mode has been obtained only with counter-injection of the neutral beams. The EHO-operation of plasma edge would be ideal in a fusion reactor: it facilitates density control and impurity exhaust while leaving the good core confinement intact. It is not clear what is the trigger for EHO, nor is it understood what suppresses the ELMs, but since counter neutral beam injection (NBI) is a prerequisite, fast ions probably play a role by affecting the edge stability properties. Therefore it is important to understand differences in the fast ion distribution in the plasma edge under regular H-mode and QHM conditions. In the QH-mode, the drift orbits of counter-injected neutral beam ions open outwards causing prompt losses to the walls but, at the same time, also feeding high energy particles into the edge region. With ASCOT simulations we were able to show that with counter-injection, a significant fast ion population exists in the edge pedestal region. A comparison of the fast ion distribution with co- and counter-injection is shown in Figure 2. In the very edge the counter-injection produces more particles at high velocities. To get more experimentally relevant distributions we also simulated co- and counter-injected neutral beam ions 42

53 Figure 2. The perpendicular velocity distribution of fast ions for a) counter-injection and b) co-injection. The thick lines show the maxim points of the distributions. The contour levels in a) and b) are the same for easier comparison. taking into account the effects of finite toroidal ripple and radial electric field E r typical of a QH-mode discharge. The ripple was found to remove particles with small parallel velocity from the distribution, as expected. The radial electric field was found to somewhat modify the distribution in the counter-injection case, but for co-injection the changes were very small Simulation of the high energy NPA signal in AUG One way to infer plasma properties in a tokamak is to measure the energy spectrum of the neutral particles escaping the plasma. The plasma ions can acquire their missing electrons from passing by neutrals or by recombining with the electrons in the plasma. When the ion is neutralised, the neutral particle leaves the magnetically confined plasma to the direction of the ion velocity. The neutrals can be detected, and information about the energy distribution of the ions can be extracted from the resulting fluxes. The diagnostic device with which this can be carried out is called a Neutral Particle Analyser (NPA). The ASCOT code has a model for the NPA. The model was originally designed for a qualitative comparison of energetic ion fluxes in different kinds of plasmas. However, since it is becoming obvious that energetic ions play a crucial role for the edge stability, it would be very important to be able to carry out realistic and quantitative simulations of, e.g., neutral beam generated fast ions. For this purpose, the ASCOT-calculated distributions have to be quantitatively benchmarked against experimentally accessible data, which consists of the neutral fluxes measured by the NPA. In February 2005 six ASDEX Upgrade discharges (shots ) were dedicated to benchmarking ASCOT against NPA measurements. In each shot the NPA had a different lineof-sight, thus sampling a different region of the phase space. In all the discharges the beams were stepped in a sequence half a second long, supposedly long enough period for a slowing down distribution to build up, and the neutral fluxes were collected towards the end of each beam step. The beams differed in energy (60 kev or 93 kev) and in the injection direction, thus populating different parts of the phase space. Three most successful discharges were simulated with ASCOT using the experimental plasma background. The experimental neutral spectra, 43

54 1e+14 AUG shot movable detector orientation 13/10 Simulated spectrum Simulated spectrum with old model *10 18 Measured spectrum 1e+13 Neutral flux* (m sr s ev) 2 1e+12 1e+11 1e+10 1e+09 1e kev 20 kev 30 kev 40 kev 50 kev 60 kev 70 kev 80 kev Energy Figure 3. The high energy neutral fluxes for a radial sightline. The fluxes are provided by very radial injection of 60 kev neutral beams. 1e+14 AUG shot movable detector orientation 13/10 Simulated spectrum Simulated spectrum with old model *10 18 Measured spectrum 1e+13 Neutral flux* (m sr s ev) 2 1e+12 1e+11 1e+10 1e+09 1e kev 20 kev 30 kev 40 kev 50 kev 60 kev 70 kev 80 kev Energy Figure 4. The high energy neutral fluxes for a radial sightline. The fluxes are provided by tangential injection of 60 kev neutral beams. together with the simulated spectra using two different numerical models, are shown in Figures 3, 4 and 5. Clearly, the results obtained with the old model, indicated by the green line, exhibit significant enough deviations from the measured spectra to warrant a complete review of the numerical NPA-model. In particular, it was found that in the model the effective detector aperture depends on particles distance from the detector. This leads to overemphasizing signal from deep inside the plasma and even from the high field side. 44

55 1e+14 1e+13 AUG shot movable detector orientation 13/10 Simulated spectrum Simulated spectrum with old model *10 18 Measured spectrum Neutral flux* (m sr s ev) 2 1e+12 1e+11 1e+10 1e+09 1e kev 20 kev 30 kev 40 kev 50 kev 60 kev 70 kev 80 kev Energy Figure 5. The high energy neutral fluxes for a radial sightline. The fluxes are provided by fairly radial injection of 60 kev neutral beams. The signal accumulation algorithm was completely rewritten so that it now gives the signal in SI units. Also, the geometrical over-simplifications were remedied. The resulting new high energy fluxes are indicated by a red line in Figures 3, 4 and 5, and they clearly match the experimental fluxes much better. The remaining differences are likely due to simplifications in the neutrals model Particle fluxes and power loads on plasma-facing components due to co- and counter-nbi injection in ASDEX Upgrade In high performance plasmas, fast ions are expected to contribute significantly to the particle and power loads on material surfaces. The fast ions created by neutral beam injection can, under some conditions, cause significant damage on the material surfaces. The fast ion flux to the first wall is particularly high with counter-injection of neutral beams, obtained by reversing the plasma current. The fluxes are further enhanced and spatially modified by the magnetic ripple, always present in tokamaks with finite number of toroidal coils. The distribution of fast ions on the wall can also be affected by the strong edge radial electric fields, characteristic of high performance H-modes. The effect of injection direction, toroidal ripple, and radial electric field on fast particle losses was evaluated using the orbit-following Monte Carlo code ASCOT. A realistic 3-dimensional ripple obtained from vacuum field calculations was used in these simulations. With counter-injected neutral beams, the wall load is substantial even without the toroidal ripple, whereas with co-injected neutral beams, significant wall load is obtained only with the ripple. In all cases the wall load is increased by the ripple, but the divertor load is either decreased (counter- injection) or unchanged (co-injection). Additionally, the ripple causes a peak in the toroidal distribution of the wall load approximately halfway between toroidal field coils. The fraction of particles hitting the divertor and wall as a function of the toroidal angle are shown in Figure 6 for counter-injected neutral beams. 45

56 Figure 6. The fraction of particles hitting the divertor and wall as a function of the toroidal angle. The simulations were carried out for AUG magnetic geometry, including toroidal ripple. Separate simulations were carried out in the absence and in the presence of an edge radial electric field. The effect of the radial electric field E r alone is small, but together with ripple it creates a peak in the wall load at R 2.1m, which on the horizontal mid-plane corresponds to the minimum of the E r profile. The particles forming the peak become trapped near the horizontal midplane due to the combined effect of E r and ripple and, subsequently, drift up or down in the direction of the B-drift. Because in AUG the toroidal field is always reversed when the plasma current is reversed, the drift is in opposite directions with coand counter-injection. Therefore the location of the peak along the wall is different. With co-injection the trapped particles form a secondary peak in the toroidal distribution, which does not happen with counter-injection. A positive radial electric field in the SOL is likely to affect the wall load, but measurements of such a field were not available for the simulations. With co-injection, the peak divertor and wall power loads are 0.2 MW/m 2 and 0.1 MW/m 2,respectively. With counter-injection the numbers are 0.5 MW/m 2 and 0.6 MW/m 2. The wall load can be considered as a lower limit, because in the simulations its numerical value is artificially brought down by the 2D wall. The simulated loads can be compared to experimental H-mode steady-state loads, which are between 1 MW/m 2 and 2 MW/m 2 on the divertor, depending on the NBI input power, and up to 1 MW/m 2 on the leading edge of the limiters Fusion-born tritium in the plasma-facing components of a fusion device The long-term retention of fuel tritium will be a critical issue in future fusion reactors. However, already in the present fusion devices, operating with pure deuterium plasmas, a detectable amount of tritium is created in DD-fusion reactions. The fuel tritium is mainly co-deposited with eroded first wall material, but in pure deuterium machines the tritium distribution on the plasma-facing components has been found to be similar to the distribution of high-energy triton implantation. This is understandable, because the fusion tritons are born with very high energy, 1 MeV. To confirm these findings, the surface distribution of 46

57 Figure 7. The triton flux to the divertor and wall, calculated by ASCOT, both in the presence and absence of the axisymmetry-breaking toroidal ripple. The position coordinates are illustrated on the left. tritons born from beam-plasma and thermal plasma DD reactions was simulated with ASCOT. Since the amount of implanted tritium per unit area on the material surfaces corresponds to the time-integrated particle flux, neglecting the decay, it suffices to study the incoming triton flux. The simulated fluxes to the divertor and wall are shown in Figure 7. The simulation results were also compared to Photo-Stimulated Luminescence (PSL) measured from the divertor and wall tiles. The simulation results are in qualitative agreement with the experimental results. The divertor results are especially good, although in the simulation there is almost no flux at the strike points. This is probably due to the fact that the simulated flux consists almost exclusively of high energy tritons, whereas in the experimental results also the thermal load is included. The ELMs, for instance, cause a flux of particles onto the strike points which, at present, can not be reproduced in ASCOT simulations. For a better match between simulations and measurement, a more detailed study comprised of a large number of different discharges may be necessary, since the single simulated discharge, an improved H-mode shot, is unlikely to accurately represent the whole experimental campaign Edge poloidal rotation and radial electric field due to NBI-generated ions in AUG The most striking characteristic of an H-mode tokamak plasma is the steep gradient region in the plasma periphery. This edge pedestal is most evident in the electron density profile, but visible also in the temperature profiles. Simultaneously with the profile steepening, a strongly sheared radial electric field and corresponding poloidal rotation are observed across the pedestal region. According to present-day consensus, the steepening of the plasma profiles is a result of the suppression of turbulent transport by the sheared radial electric field. Unfortunately, the mechanism that generates the radial electric field and, thus, is responsible for the L-H transition, remains elusive. However, there are experimental indications that radial current might play a significant role. The most direct evidence is brought by experiments where an L-H transition has been induced by external biasing. Also theoretical work by Shaing and Crume was able to yield bifurcation-type of solutions for the plasma equilibrium, when the outward radial current due to hot pedestal-top ions was assumed to be balanced by neoclassical return current. 47

58 Because it is known that the L-H transition can be controlled by external heating, it is suggestive that direct ion orbit losses of fast ions, e.g. those generated by neutral beam injection, might play a role in generating a significant part of the radial current. Indeed, in QH-mode, obtained with counter-injected beams, the experimentally measured radial electric field is about twice the magnitude of that observed in H-modes with co-injection. In ASCOT-simulations of NBI-generated ions with co- and counter-injection in AUG, the direct ion-orbit-loss current was found to be twice as large for counter-injection than for co-injection. These simulations included the ripple effects. To obtain the poloidal rotation driven by such a radial current, an attempt was made to solve the poloidal torque and the resulting poloidal rotation according to the theory by Shaing and Crume. However, in all interesting cases, i.e. those with the effects of ripple taken into account, the ASCOT-evaluated orbit-loss current turned out to be so large that no steady-state solution exists. It was concluded that also the effect of ion-neutral friction should be included in the calculation. Once an appropriate expression for the ion-neutral collision frequency is found, these calculations will be carried out for various plasma configurations and heating powers Modelling of ITER divertor load distributions using ASCOT The ASCOT code has been applied to studies of ITER divertor target loads arising from orbit-loss ions escaping from the plasma. The effect of temperature and density pedestal width and height has been studied to find out whether the peaked load distributions as observed e.g. on JET give cause to concern in ITER. The unprecedented size and parameters of ITER make predictions of target loads important. The robust particle-following approach applied in ASCOT, including the self-consistent q [MW/m ] 2 target Inner target n pedestal width 0 14 mm 21 mm q [MW/m ] 2 target Outer target Outer midplane R R sep [mm] Outer midplane R R sep [mm] q [MW/m ] 2 target Inner target T pedestal width 0 14 mm 21 mm q [MW/m ] 2 target Outer target Outer midplane R R sep [mm] Outer midplane R R sep [mm] Figure 8. (top) The effect of density pedestal width on divertor target loads in ITER case #585. With realistic finite pedestal widths, the peak loads remain below 30 MW/m 2. (bottom) The effect of temperature pedestal width on divertor target loads. The black curve is the same as in the top row of figures. 48

59 modelling of edge radial electric field, makes the code well suited for such predictions. For the simulations, a realistic ITER magnetic background and B2/EIRENE scrape-off layer (SOL) data were imported from work conducted by the ITER team for ITER reference case #585. While changes in n and T pedestal heights produced an expected up- or downscaling of the target loads, the effect of pedestal width was found to be much more dramatic. With zero-width density and temperature pedestals, target load peaks of the order of 90 MW/m 2 were recorded. Assuming a constant temperature of 4 kev up to the separatrix and introducing a finite density pedestal width of just 14 mm lowered the target load peaks below 30 MW/m 2. Assuming a constant density of m -3 up to the separatrix and introducing the same finite width to the temperature profile lowered the target load distribution peaks below 3 MW/m 2. The temperature pedestal width, which has a strong effect on the Coulomb and CX collisionality of the escaping ions in the SOL and divertor regions, was thus identified as a key parameter affecting the target load peaks. 2.5 Theory and Code Development First principles turbulence simulations with ELMFIRE code In 2002, the fusion and plasma physics group at the Technical University of Helsinki and at VTT started the coding of a kinetic plasma turbulence simulation programme ELMFIRE, which allows large perturbations from equilibrium. This code can be regarded as a true transport code, as it can follow the evolution of the plasma equilibrium with turbulent transport. ELMFIRE is a full f particle-in-cell code where an enormous number of test electron and ion clouds, each representing billions of true particles, are followed in the 5-dimensional phase space according to their gyrokinetic equations of motion. The Coulomb collisions among the particles are treated with Monte Carlo methods, whilst the long-range collective interactions are accounted for by solving the gyrokinetic field equation. Macroscopic quantities like plasma density, temperature, and plasma flow velocities are obtained from statistical samples of the particle ensembles. Typically, a time step of ns is required for accurate sampling and for solving the field. For the small tokamak FT-2 (Fisichiskii Tokamak-2, Ioffe Institute, St Petersburg), about 100,000 grid cells are needed to resolve the field and plasma parameters to a satisfactory resolution in configuration space. As for good statistics, roughly 1,000 electrons and 1,000 ions are needed in each cell, the total particle number amounts to 100 millions. ELMFIRE code is carefully benchmarked against other (delta f) codes in linear growth rate and frequencies of unstable modes for the so called Cyclone Base configuration. Moreover, the code has passed numerous successful tests of neoclassical physics including the radial electric field evolution in various pressure gradient and collisionality regions. Turbulence saturation benchmarking among other codes is ongoing within the ITM IMP4 project. In FT-2 ELMFIRE runs, the radial ion flux from the hot core plasma to the colder edge plasma caused by turbulence is suppressed by sheared plasma flows for sufficiently strong heating. This shearing of plasma poloidal rotation is caused both by turbulence itself and by neoclassical effects. The result is important as it first demonstrates the self-regulation of turbulence in firstprinciple simulations of tokamak transport. This self-regulation is believed to explain socalled transport barriers observed in present-day fusion experiments. At the transport barriers, the anomalous transport is reduced close to the ion neoclassical level ( times reduction). If such barriers can be realized for ITER, either inside the core plasma or at the edge plasma, a significantly improved confinement results. The possibility of operation with transport barriers allows the direct reduction of the size of the fusion reactor and thus reduction of its costs. 49

60 d) Figure 1. Normalized radial particle flux (green inwards, red outwards) in a poloidal cross section (a) well before, (b) just before and (c) after the transport barrier generation. The occurred vortex size reduction prevents loses outwards. d) Density fluctuations of an unstable structure as calculated in a cross section of tokamak plasma. As in other physical applications, the large-scale problems under study using gyrokinetic model will require hundreds or even thousands of processors during many days of computation. These high-demand requirements have lead to initiatives like NORDUnet in the European Nordic countries (e.g. Finland) and the Distributed European Infrastructure for Supercomputing Applications (DEISA) that are joining supercomputing centres by means of high speed network connections to provide joint parallel computation. ELMFIRE will be run within the DEISA network in 2007 whereby it will be upgraded to calculation of electromagnetic fluctuations and the scrape-off-layer plasma. Keeping the current trend, the 10 PFlops key figure that makes possible the numerical simulation of ITER will be a reality by the time ITER operation starts. 50

61 2.5.2 Development of the ASCOT code The ASCOT code (Accelerated Simulation of Charged Particle Orbits in a Tokamak) is a versatile guiding-centre orbit following code for studies of charged particle behaviour in tokamaks. It has been developed in collaboration between Helsinki University of Technology and VTT since the early 1990 s. During the period , several improvements to ASCOT have been made or initiated. The neutral beam injection (NBI) initialization models of ASCOT have been improved. A more realistic beam attenuation model has been added to the generic NBI model. The new model allows initializing test ions even on the high field side, which is crucial when modelling the global fast ion distribution. ASCOT can now also read neutral beam test particle data directly from specialist NBI codes such as FAFNER. Building the magnetic background for ASDEX Upgrade simulations has been completely restructured to achieve better quality by using a more accurate interpolation algorithm to avoid numerical inaccuracies. ASCOT is now also directly interfaced to the ASDEX Upgrade experimental profiles with a new preprocessor that utilizes the pedestal profile database structure based on MDSPlus. Using MDSplus is in accordance with the proposals made by the ITPA Pedestal and Edge Group. Implementation of the MDSplus data tree structure also for the results of ASCOT simulations has been initiated. This will involve establishing an MDSplus server also at Helsinki University of Technology to provide research results openly and efficiently within the fusion community. The most important output distributions produced by ASCOT are now in physical units. The toroidal magnetic field ripple model has been improved. The old analytical model has been replaced by a more accurate but more memory-consuming model based on magnetic measurements on the three components of the magnetic field. The new model also ensures that the ripple is divergence-free. To achieve better code portability, all mathematical subroutine library dependencies have been removed from ASCOT. The code has been installed on the JET Analysis Cluster (JAC) in Culham, UK, making it available to numerous new users. Further refinements to the self-consistent edge radial field model have improved the agreement between simulated and experimentally measured divertor target load profiles in JET. The load asymmetry between the inner and the outer target now agrees with measured results without the need to apply substantial external radial electric fields in the scrape-off layer (SOL). ASCOT handling of background density and temperature data has also been substantially improved. Background density and temperature information can now be imported in a 2D mesh independently for all plasma and neutral species. This facilitates more realistic studies of e.g. impurities at the plasma edge. In addition, it is now possible to take charge-exchange collisions into account in the core plasma as well as in the scrape-off layer. A version of ASCOT suitable for stellarator simulations is in the testing phase. In this version, a three-dimensional background magnetic grid is imported instead of the usual two-dimensional grid. The capability to import three-dimensional magnetic grids from experiment databases is being developed also for tokamaks and will allow the use of the most realistic magnetic equilibria. Related to the fully 3D magnetic background model, a new model for taking into account 3D wall structures is being developed. These improvements will be of importance in studies of SOL, wall and divertor physics Particle-in-cell modelling of plasmas A 2d3v particle-in-cell (PIC) code, XOOPIC, was used to study the coupling of LH waves from the grill to the plasma. In this study the parameters of the Italian FTU tokamak were used. The frequency is higher, i.e. 8GHz versus 3.7 GHz used in JET and Tore Supra. In the simulations, the reflection coefficient can be obtained for each waveguide. Since the reflection coefficients at FTU can be measured in each waveguide, this 51

62 would give a good opportunity to compare the code with and benchmark against the experimental results. Moreover, the FTU grill has only 12 waveguides while JET and Tore Supra have 32, which makes it slightly easier to perform simulations of the whole grill. The effect of the edge density as well as the density gradient in front of the launcher was studied. First homogeneous density profiles were used. The behaviour was expected as the coupling was improved when the density increased above the cut-off density below which the wave is not supposed to propagate. The optimum for the coupling was obtained at about 4 to 5 times the cut-off density and about twice the one found analytically. This was similar to previous studies for JET and Tore Supra. When a linear density gradient was added, a remarkable improvement in the coupling at edge densities around the cut-off density was observed. As in earlier studies, well above the cut-off density the effect of the density gradient was very small As a spin-off from fusion, a 1d3v PIC code XPDC1 has been applied to plasma chemical processing of materials including oxidisation and decomposition of waste. A library of reactions of oxygen has been added to the original code. Thus oxygen can be used as a background gas instead of or together with argon. This enabled us to find a better understanding of the behaviour of low temperature plasmas used for resin incineration ERO plasma-wall code development Background: ERO is a Monte Carlo based impurity transport code developed at IPP Garching and Forschungszentrum Jülich (FZJ, Germany). The code models the motion of impurity particles in the scrape-off layer of tokamak plasma. ERO has a detailed description of plasma-wall interactions and relevant chemistry: Ionisation rates for atomic species, molecular processes of methane and silane molecules as well as sputtering and reflection yields are included through external databases. In ERO simulations, test particles originating from a surface like a limiter or divertor plate are followed inside a local tracing box. The limiter version (mainly used for the small TEXTOR tokamak at FZJ) is locally 3-dimensional, i.e., the tracing box is 3D but the curvature of the magnetic field is neglected, limiting the size of the simulation volume. The divertor version, which is required for divertor tokamaks like ITER, JET and ASDEX Upgrade, has been previously used only in 2D, but during , 3D simulations of ASDEX Upgrade and JET have been run at TKK. TKK has participated in ERO development since The code is presently running on a workstation at TKK, which was found to be a convenient environment for code maintenance and simulations. The TKK team has so far implemented some significant improvements in ERO, which are described below. So far all our simulation projects have been accompanied by customization of ERO and are still ongoing work. Numerical enhancements: ERO requires either a parameterised or plasma-code-simulated background plasma for test particle tracing. The divertor version always uses a background simulated by SOLPS or EDGE2D plasma codes. The calculation grids of these codes are incompatible with ERO. We have improved the background plasma routine of ERO and written a preprocessor for accurate and efficient interpolation of the parameters in ERO simulations. The savings in the CPU usage make it possible to do much more demanding simulations. Merging of limiter and divertor versions: Over the years, several subversions of ERO have been born as a result of minor modifications for different purposes. For example, there have been separate codes for each simulated divertor tokamak: JET, ASDEX Upgrade and ITER. Moreover, limiter tokamaks and linear plasma machines have their own code branches. This poses a problem of development, since all branches have in common a large part of the code, which undergoes continuous development and is currently stored as several copies that are not linked to each other. To facilitate more efficient maintenance, a long-term goal is to have a single code capable of simulating all different cases. In summer 2005 implement- 52

63 ing all divertor-relevant functionality as an option in the up-to-date limiter version started the work towards such unified ERO. At the moment this project is in testing phase. Molecular Dynamics based surface model for ERO: To properly describe material mixing in plasma-surface interactions, a new surface model is required in ERO. A Molecular Dynamics based model is being developed as an EFDA task by TKK in collaboration with the Accelerator Laboratory of University of Helsinki and Forschungszentrum Jülich. The HCPARCAS Molecular Dynamics simulation code will be coupled directly to ERO instead of generating a pre-calculated database for interpolation during ERO simulations. HCPARCAS uses the classical Brenner-Beardmore potential model, which is simple enough to guarantee fast computation and also provides an adequate description of the interatomic potential. Periodic boundaries are enforced and the system is treated as bulk material. At the beginning, the major task is to generate an ensemble of MD simulation samples. For instance, for different relative concentrations of C and H, about 10 representative samples will be needed. Moreover, typically six samples are generated for each concentration to represent some statistics. Within the framework of the present task we aim at demonstrating the concept with a set of samples of varying relative C/H/W concentrations. Later, some refinement of the model by adding material compositions will most probably be necessary. In order to create samples with correct material and chemical properties numerous simulation runs are needed: to achieve the minimum energy configuration the system is annealed at high temperatures and quenched down to room temperatures. This process is repeated several times. In addition, the system is simulated at high pressure to make sure that the concentration of sp 2 and sp 3 bonded carbon atoms agrees with experimental data. Thereafter, the system is simulated at room temperature and zero pressure to form a stable cell. Finally, one of the surfaces is opened by removing the periodic boundary and thus letting any nonbonded atoms to escape the sample. In total, a single sample requires a couple of days of computation time. The generation of MD samples has been started, and the first sample now shows a correct bond ratio. As the next step, we are moving to mass production of necessary samples. At a later stage the samples will be used as targets for bombardment simulations, which replace the current particle-wall interaction model of ERO. It is imperative that the MD simulations of the interactions be as swift as possible, and therefore the number of atoms in each sample is kept to the minimum: a little under one thousand for pure carbon and gradually increasing with the addition of other elements. At the expected collision energies this volume is adequate Development of JET codes Traditionally, the code management group at JET has been responsible for maintaining and supporting the 1.5D core transport code JETTO, the 2D edge transport code EDGE2D / NIMBUS and the coupling of these two codes known as COCO- NUT. Association Euratom-Tekes has been closely involved in this work by taking part in the upgrading of JETTO with a range of new features including ad hoc ELM models for type I, type II, mixed type I-II and type III ELMy H-mode, theory-motivated ballooning and peeling ELM models (see preceding sections in this report), an extension of the ELM modelling schemes making use of a database of MHD stability results, an implementation of the Porcelli sawtooth reconnection model, a model for neo-classical tearing modes, a model for convective thermal ion ripple losses implemented as an energy sink term in the continuity equation for the ion pressure and an implementation of the ITER neutral beam geometry, to mention but a few of the developments. In 2004, a major code integration project was launched at JET with the aim of providing graphical user interfaces and common standards for input and output for a number of mainly transport and MHD stability codes used on site, including also codes originating from outside JET. This integrated suite of codes is known as JAMS (JET Application Management System). The idea is to make the codes more user-friendly to work with, to make it easier to compare different codes with each other by providing common standards for 53

64 input and output and eventually to allow the best codes to survive and evolve. Currently, JAMS features graphical user interfaces for the core transport codes JETTO, CRONOS and ASTRA, the edge transport code EDGE2D, the coupling of JETTO and EDGE2D known as COCONUT, the MHD stability codes HELENA, MISHKA and ELITE and the gyrokinetic microturbulence code KINEZERO as well as a number of tools for data preparation, manipulation and storage. Association Euratom- Tekes has been heavily involved in developing the graphical user interfaces for the HELENA, MISHKA and ELITE MHD stability codes. A currently ongoing part of the project is to link the Monte Carlo guiding centre orbit following ASCOT to the JAMS suite of codes. 2.6 Diagnostics Upgrading of AUG neutral particle analyser Helsinki University of Technology S. Jämsä The design work for upgrading the Neutral Particle Analyser (NPA) of the ASDEX Upgrade (AUG) -tokamak in Garching, Germany, was carried out. The upgrade will enable time-resolving of individual counting events, thus making it possible to resolve the NPA signal during sporadic events, such as ELMs, from the signal during the steady-state phase. The current data acquisition system (DAQ) of the NPA detector, shown in Figure 1, relies on multichannel scalers to collect the particle rates. The new DAQ will be based on the S-Link, a CERN specification for an easy-to-use FIFO-like data-link. In the present system the particle rates are given as particles per time-bin. The time bin can be of different length. The number of bins in the system is limited to approximately 8000 bins. This means that only part of a shot can be detected with high temporal resolution. To further aggravate the problem, the bins must be set before the plasma is created. Therefore the high resolution Figure 1. The current data acquisition system of the NPA in AUG control room. part of the data can be in the wrong part of the shot, or be too short. The new DAQ will keeps count of each detected particle separately, with at least the temporal resolution of the detector. For example the detector can be sampled for detected particles every 50 ns and all detections are saved. The electrical signal from the detector is converted into a format understood by the electronics readily available at AUG. The conversion is done with dedicated electronic device which is an in-kind contribution from AES-laboratory to AUG. The binning will be performed later in software. With the new system AUG will be first to have better-than-microsecond resolution NPA. This will make possible to analyse rapid phenomena, such as ELMs. Tungsten sputtering has been found to occur strongly during ELMs. One possible source of the tungsten erosion is high energy ions, which can be detected with the NPA. The time resolution of the new system may make it possible to detect passing filaments. The physical upgrade of the NPA system will take place in

65 2.6.2 Upgrading JET NPA system Helsinki University of Technology Marko Santala VTT Micro- and Nanoelectronics Simo Eränen, Juha Kalliopuska The neutral particle analysers (NPAs) at JET are used for measuring atom flux, ie., neutralised ions, that escape the plasma. These measurements are particularly useful because the origin of the atom flux is the ion distribution deep in the plasma. Ions can be neutralised through several processes, most commonly through charge-exchange with hydrogenous atoms or partially ionised impurity ions. Generally, the ions retain their original velocity and direction upon neutralisation. Unless they are re-ionised before escaping the plasma, they can be detected if they were moving towards the NPA. Re-ionisation, however, is very likely for thermal ions neutralised in plasma core but energetic ions, like fusion products or RF accelerated ions, are likely to escape. Hence, NPAs give information predominantly from plasma edge at low energies and from plasma core at high energies. There are two NPAs installed at JET. The high energy NPA (GEMMA-2M) is installed on top of the JET machine and has a vertical line-of-sight. It can measure one ion species on eight energy channels with energy of kev for hydrogen isotopes and up to 3500 kev for He. The low energy NPA (ISEP) has a horizontal, radial line-of-sight through the plasma core. It measures simultaneously all three hydrogen isotopes on a total of 32 channels. The energy range can be configured from 5 kev to 750 kev (for H). The diagnostic hardware as well as all data collection electronics has been supplied to JET by Ioffe Institute, St. Petersburg. In the NPAs, a very thin (<50 nm) carbon foil is first used to re-ionise the atoms. The ions are then deflected by parallel magnetic and electric fields, which separates different isotopes and different energies spatially. Thin (<10 μm) CsI(Tl) scintillators coupled to photomultipliers are used for detecting the ions. Each detector detects a narrow energy band of the selected ion species. Every ion produces a tiny current impulse, and these impulses are counted. The scintillator-based detectors have a number of issues: They have poor energy resolution and they are relatively sensitive to gamma and neutron background. This causes spurious counts that cannot be completely discriminated. The CsI(Tl) is fairly slow scintillator, which limits the count rate. Furthermore, ions with equal charge to mass ratio (like d and 4 He) are not separated spatially by the electro-magnetic fields, and the CsI(Tl) scintillator cannot distinguish them either. To address the problems of the scintillator detectors, an EFDA-funded project has been started for Tekes to develop new detectors for the NPAs. The new detectors will be made of silicon with thin active layer to keep sensitivity to background radiation low but to maintain high sensitivity to ions. The new detectors will be installed during the next major shutdown at JET. NPA Detector Design and Fabrication: In the present task the activities consist of the R&D work for the new NPA prototype realized with a silicon detector. The success of the silicon detectors in the particle detection application is clearly evidenced e.g. by the activities at CERN and other high energy physics experiments. In most of the cases the silicon particle detectors are employed as position sensitive devices in order to find out the particle tracks. However, in the NPA the particle track is not of extreme importance, but the detector is employed in the spectroscopic mode for the detection of the particle energy. The essential requirements for the silicon detector in the NPA are: radiation hardness and background reduction of the heavy neutron and gamma irradiation. The both requirements are tackled by designing the silicon detectors with the active thickness of 5 to 35 microns. The other point is, that at such a thickness level silicon wafers are not anymore mechanically self-supporting, but a carrier wafers are needed. Thus the NPA proto types will be realized on the SOI (silicon on insulator) wafers, where the thin active detector layer is separated from the mechanical support wafer by a insulator (silicon dioxide) 55

66 Figure 2. Potential distribution in a NPA prototype (see text for details) layer with the thickness of approximately 1 micron. Further, in order to boost the radiation hardness, the active layer is made of p-type CZsilicon. Figure 2 shows the potential distribution in a NPA proto type detector at the bias voltage of 20 V. Here the active layer has the thickness of 30 microns and the resistivity of 1 kωcm. Due to the positive oxide charge on the front surface the p-type doping level is enhanced (p-stop) between the n-type read-out electrodes. The support wafer has usually the thickness of 300 to 500 microns, and is only partly displayed Optimization of W7-X interferometer Helsinki University of Technology Markus Airila and Olgierd Dumbrajs A multichannel CO 2 laser interferometer is planned for electron density profile measurements in the Wendelstein 7-X stellarator under construction in Greifswald, Germany. The interferometer can continuously provide local electron density information after applying mathematical inversion procedures to the line integrated measurement data. It is therefore one of the most important diagnostics on W7-X. The spatial resolution obtained depends strongly on the choice of the sightlines. During , in collaboration between TKK and IPP, new mathematical methods were developed for finding the optimum orientation of the various sightlines. In particular, it was taken into account that the system will be installed in groups of four beams. The problem of reconstructing local densities from line integrated density measurements can be formulated as a matrix inversion. The quality of the reconstruction is largely determined by the condition number of the matrix describing the problem, which has to be minimized, i.e., the sightlines have to be chosen such that the information content of different channels is maximally distinct. For the W7-X interferometer four- and eight-sightline arrangements were investigated and their orientations are optimized 56

67 Figure 3. (a) and (b) The optimized sets of four and eight sightlines traversing the W7-X plasma cross section. (c) A simulated reconstruction of an extremely broad density profile with eight and four sightlines. In our method each beam is assigned the task of providing maximum information about the density at one particular depth indicated by the arrows, separately for the four- and eight-beam configurations. for standard conditions of magnetic configuration and density profile (see Figure 3). The optimized arrangements were tested by simulating the inversion of hypothetical reference density profiles in a number of different magnetic configurations. In the ideal case the error of a reconstruction using four sightlines is typically a few times larger than that with eight sightlines. The robustness of the optimisation was demonstrated by a variation of the position of the whole interferometer, and the influence of noisy phase data on the reconstructed profiles is investigated. These factors narrow significantly the difference between the four- and eight-beam setups. 57

68 2.6.4 Smart tiles for first wall diagnostics VTT Materials Performance J. Likonen, Elizaveta Vainonen-Ahlgren, Tommi Renvall Diarc Technology Oy J. Kolehmainen, S. Tervakangas In all present devices, the tokamak inner divertor is a region of net deposition and the outer divertor being a zone of net erosion or erosion/redeposition. This is due to partly the prevalence for a colder, denser inner divertor and a hotter outer divertor. Wall erosion, material migration and associated deposition preferentially at the inner divertor are among the most critical issues to be solved for ITER and for future fusion devices because of long term tritium retention in deposited layers. Heavy deposition at the inner divertor has led to flaking on the water-cooled louvres and after the DTE1 tritium experiment at JET it was observed that the majority of the retained tritium is in the flakes that have spallen from the louvres. Studying these processes constitutes one of the main research avenues at JET and ASDEX-Upgrade. JET is a device dominated presently by carbon walls but also unique both in its use of beryllium and for its tritium capability. The ASDEX-Upgrade tokamak has also been steadily acquiring experience with operation in a carbon and tungsten environment as a precursor to future experiments with a full W wall and divertor. The principal existing diagnostic method for measuring hydrogen isotope retention and erosion of first wall components is the analysis of tiles removed from the tokamak. Quantification of erosion in present tokamaks is based on the analysis of so-called smart tiles which have a marker layer with known thickness. Marker film is typically a carbon or metal layer. Smart tiles coated by DIARC-Technology Inc have successfully been used both at JET and AUG. The thickness of the marker layer is measured prior to the installation into the tokamak with surface analysis techniques. After the experimental campaign the smart tile is removed from the tokamak, the thickness of the marker film is measured again and the erosion rate can be determined. Smart tiles allow also investigation of deposition because co-deposited layer on top of the marker layer can be identified more easily. Figure 4. JET outer divertor tile with a tungsten marker layer. Figure 5. Same tile as in Figure 4 after exposure at JET in Figure 4 shows an example of a JET outer divertor tile coated with a tungsten stripe (thickness 3 µm). The tile was installed in 2001 and removed in After removal from JET visual inspection shows a complex erosion and deposition pattern (see Figure 5) on the tile. The W marker layer has been eroded very strongly at the bottom part of the tile due to the outer strike point. At the centre (yellowish band) and at the top (dark band) there is some deposition. For AUG several sets of divertor and limiter tiles have been coated both with carbon and various metals. Figure 6 shows an example of an AUG probe coated with carbon, aluminium, nickel and tungsten. The probe can be inserted into the plasma for a short time during a discharge allowing real time erosion studies. The smart tiles shown in Figures 4 6 are passive in a sense that the marker layers do not 58

69 Figure 6. AUG time-resolved probe with carbon, aluminium, nickel and tungsten marker layer. Figure 7. Smart tile at JET for in-situ monitoring of erosion. have any functional properties. In Figure 7 is shown a smart tile which allows in-situ monitoring of erosion. The tile has thin buried titanium and chromium layers under a carbon layer. Aim is to analyse emitted visible light from titanium and chromium layers when the topmost carbon layer has been eroded. This tile is right now under investigation at JET Micromechanical magnetometer VTT Sensors A. Kärkkäinen, J. Kyynäräinen, J. Saarilahti, A. Oja, and H. Seppä VTT s feasibility study evaluates the possibility to use a micromechanical (MEMS) magnetometer to measure steady state local directional magnetic field in ITER environment. Introduction: ITER requires an extensive set of diagnostic systems to provide several key functions in support of the design goal: protection of the device, input plasma control systems, evaluation, and analysis of plasma performance. One of the key areas in ITER diagnostic is the characterisation of the magnetic field. This will be mainly based on induction coils. A magnetic flux going through a coil creates a voltage proportional to the field strength. The magnetic flux can be measured by integrating the voltage as function of time. This is a valid method for plasma reactors where the length of a test pulse is order of seconds. The pulse length in ITER will be 3600 seconds, which implies that the measurement result is prone to various integration errors. These include disturbances arising from strong background magnetic field, nuclear heating, thermal gradients, and EM forces during disruption. Hence magnetometers that measure the absolute value of the magnetic flux density of about 2 T are needed to correct the steady-state drift of the coils. After 10 years of discussion, some design and R&D, ITER still lags a steady-state magnetic sensor. Recently the major research effort has focused on development of a Hall sensor. However, this principle of operation is fundamentally radiation dependent since the Hall constant depends on the number of charge carriers in the semiconductor which is increased by neutron radiation. A new solution is needed. MEMS: The excellent properties of silicon micromechanical (MEMS) sensors have been demonstrated in many commercial applications such as pressure sensors, air bag inflators, gyroscopes, etc. The sensors are characterized by small size, low power consumption, and robustness. Silicon itself can withstand high temperatures, although careful selection of electrode material needed as well as a temperature compensation for the sensor. Silicon semiconductor detectors have successfully been used in high radiation environments for decades. All these features make MEMS sensors attractive candidates to monitor the steady state magnetic field at ITER. Small and affordable sensors can be placed dense and redundantly. Hence local changes in the magnetic flux can be detected and even some loss of 59

70 sensors can be tolerated. These properties are very valuable in ITER environment where maintenance or replacement of sensor is difficult. VTT Magnetometer: VTT has designed, manufactured and characterized a 3D micromechanical magnetometers intended to detect Earth s magnetic field (about 50 µt). The magnetometer is based on Lorentz force. The same principle of operation can be used at high magnetic fields as well after complete rescaling of the dimensions. In the devices shown in Figures 8 and 9 the Lorentz force is produced by a current-carrying coil, processed on a single crystal silicon resonator. Mechanical resonance is employed to lower power consumption and to enhance the signal level, thus reducing the effect of electronics and 1/f noise. Sensors for all Cartesian components of the magnetic field vector can be processed on the same chip. The vibration amplitude is detected capacitively and the resonance is tracked by a phase-locked-loop circuit. An optional force feedback can be used to increase the dynamic range and to shorten the response time. MEMS at ITER: The high magnetic field in ITER induces orders of magnitude larger Lorentz force compared to Earth s field simplifying the sensor design, fabrication and readout electronics. Neither resonance operation nor vacuum encapsulation is required. As a drawback, capacitive readout requires that low-level signals are transmitted over quite long cables which reduces signal-to-noise ratio. Figure 8. Principle of operation of the magnetometer sensitive to the magnetic field component B along the chip surface. Electric current is denoted by I and the Lorentz force by F. Figure 9. Principle of operation of the magnetometer sensitive to the magnetic field component B normal to the chip surface. Radiation caused damage mechanism in silicon semiconductor detectors are well known. However, these studies are mainly concentrated on electric properties of silicon material like change of effective doping concentration, creation of generation/recombination centres, and charge carrier trapping in silicon. In MEMS applications the silicon mechanical properties have to be considered as well. The effect of radiation on the Young s modulus of silicon, which is not known, is crucial in this application since it would change the spring constant of the device. Problems caused by dielectric charging of MEMS sensors due to nuclear radiation have been reporter in scientific literature. However, these problems can be eliminated with proper sensor design. Conclusion: VTT 3D MEMS magnetometer is a viable candidate for a steady state magnetic field sensor at ITER after rescaling of the dimensions. Optimization of the magnetometer design for this purpose is ongoing, as well as further characterisation of MEMS components radiation response. 60

71 3 Fusion Reactor Materials Research VTT Materials Performance L. Heikinheimo (Technology Manager), S. Tähtinen (Project Manager), P. Moilanen, S. Saarela, H. Jeskanen, U. Ehrnstén, P. Karjalainen- Roikonen, P. Kauppinen, K. Lahdenperä, A. Laukkanen, T. Planman, L. Taivalaho, A. Toivonen, W. Karlsen, P. Aaltonen, M. Valo and K. Wallin, R. Rintamaa Metso Powdermet Oy, J. Liimatainen Luvata Oy, T. Parviainen 3.1 Characterisation of Copper Alloy to Stainless Steel HIP Joints Tensile and fracture behaviour The current design of ITER utilises CuCrZr alloys in the first wall and divertor structures. The function of copper alloy is mainly to dissipate heat produced by plasma steady state operation and disruptions. The copper alloy is not designed to provide structural support for the first wall, however, both heat dissipation and structural support is expected for the divertor cassette. The anticipated temperatures range for copper alloys in the first wall and divertor is from 100 C to about 350 C. Mechanical and fracture behaviour of CuCrZr to 316LN joints are dominated by the properties of the copper alloy, and particularly, by the strength mismatch and mismatch in strain hardening capacities between copper alloy and stainless steel. The mismatch promotes localisation of plastic strain in the copper alloy side of the joint and near the joint interface. The test temperature, neutron irradiation and thermal cycles related to component manufacturing or operational cycles primarily affects the CuCrZr to 316LN joint properties through changing the strength mismatch between the base alloys. Mechanical properties of precipitation hardened CuCrZr alloy and corresponding 316LN joints are dominated by applied thermal cycle during manufacturing process. Table 2 summarises the tensile properties of the CuCrZr alloy and corresponding joints after different manufacturing heat treatments expected for component manufacturing. It is noted that the specimens T10-4 and T13-2 show very similar tensile properties for both CuCrZr base alloy and corresponding joints indicating that CuCrZr is notably softer than stainless steel and that tensile properties of joints are dominated by those of copper alloy. In the case of T12-3 the CuCrZr base alloy shows higher yield strength than that of corresponding joint which indicate that stainless steel or joint interface is probably softer than CuCrZr base alloy. In this case yield strength of joint specimen is described by yield strength of stainless steel although tensile strength of joint is dominated by that of CuCrZr base alloy. Joining of CuCrZr alloy with stainless steel by applying Hot Isostatic Pressing (HIP) method in temperature range of C in actual fact means solution annealing heat treatment for both alloys. The obtained strength level of the CuCrZr alloy, however, is dominated by subsequent cooling rate and aging heat temperature where as that of stainless steel is expected to be relatively insensitive to those heat treatments. The yield strength of the CuCrZr alloy may vary more than about 150 MPa depending on applied cooling rate and aging heat treatment. The yield strength of the CuCrZr alloy in specimen T12-3 is very close that of prime aged CuCrZr alloy whereas those of specimens T13-2 and T10-4 are about 60 MPa lower. This means that at room temperature the CuCrZr alloy have higher yield strength in specimen T12-3 and lower in specimens T13-2 and T10-4 than that of stainless steel which is about 253 MPa. It is also noted that at 350 C the yield strength of the CuCrZr alloy is close to or higher 61

72 Table 1. Summary of manufacturing heat treatments of CuCrZr/316LN joints carried out by CEA, France. Code HIP Heat treatment Ageing T C/2hrs/140MPa 980 C/0.5hrs/MCR* 560oC/2hrs T C/2hrs/140MPa 980 C/0.5hrs/MCR* 480 C/2hrs T C/2hrs/140MPa 980 C/0.5hrs/MCR* 560 C/2hrs * MCR: medium cooling rate C/min. Table 2. Summary of average tensile properties of CuCrZr alloy and corresponding CuCrZr/316LN joint specimens after different manufacturing heat treatments at room temperature determined by CEA, France. Code Material Yield stress (MPa) Ultimate stress (MPa) Uniform elongation (%) Total elongation (%) T10-4 CuCrZr CuCrZr/316LN T12-3 CuCrZr CuCrZr/316LN T13-2 CuCrZr CuCrZr/316LN than that of stainless steel in all specimens T12-3, T13-2 and T10-4. The lowest yield strength values for the CuCrZr alloy are associated with HIP joint specimens with very slow cooling rate from HIP temperature followed by subsequent aging heat treatment. Fracture toughness behaviour of HIP joint specimens seems to be dominated by the strength mismatch of the CuCrZr alloy and stainless steel. Those HIP joint specimens which have experienced relatively high cooling rate and optimum aging heat treatment have subsequently low strength mismatch (CuCrZr have either higher or lower yield strength that stainless steel) and show relatively low fracture toughness values. On the other hand, those HIP joint specimens which have experienced slow cooling rate or over aging heat treatment have larger strength mismatch (CuCrZr have substantially lower yield strength that stainless steel) and show relatively high fracture toughness values. The HIP joint specimen T12-3 have lower strength mismatch and fracture toughness when compared with specimens T13-2 and T10-4 which have higher strength mismatch and fracture toughness. The fracture toughness of the HIP joint specimens seems to decrease with increasing temperature as can be seen in Figure 1. It is also noted that joint interface type of fracture mode is favoured at elevated temperatures. Temperature dependence of both mismatch in yield strength (yield strength of CuCrZr alloy is less temperature dependent that that of stainless steel) and mismatch in work hardening capacities are expected to favour low fracture toughness and interface type of fracture mode at elevated temperatures. On the other hand also the metallurgy of the HIP joint interface between CuCrZr alloy and 316L(N) stainless steel is relatively complicated (precipitates, phase transformations, composition and hardness gradients) and the role of interface structure in fracture behaviour is not clear. It is worth noting that at temperatures above 300 C relatively low fracture toughness values have been observed independent on mismatch in yield strength. 62

73 500-2 ) 400 CuCrZr / 316L(N) ycu / y SS << 1 Cu / SS ~> 1 y y J initation J (kjm Q TEMPERATURE ( C) Figure 1. Initiation fracture toughness of HIP joint specimens between CuCrZr alloy and 316L(N) stainless steel Stress corrosion cracking susceptibility of tube joints The copper alloy to stainless steel tube joints will be applied in connections between water cooling tubes of divertor components and main divertor cassette. The test specimens were prepared by electric discharge machining from the tubes delivered to VTT by Ansaldo, Genova. The tube samples were produced by electrodepositing Ni layer on CuCrZr alloy followed by TIG welding between electrodeposited Ni layer and 316L stainless steel. Two slow strain rate test runs with four specimens were conducted in an autoclave testing system with recirculating high temperature water that contained controlled amounts of hydrogen peroxide, sulphates and chlorides. Nominal concentrations of sulphate, chloride and hydrogen perox- 200 STRESS (MPa) A 5B 6A 6B STRAIN (%) Figure 2. Stress-strain curves of CuCrZr/316L joint specimens in SSRT run no

74 ide were 10 mgl -1, 4 mgl -1 and 5-10 mgl -1, respectively. The applied strain rate during the tests was 10-7 s -1. The nominal temperature and pressure were 150ºC and 60 bar respectively. Redox potential, conductivity, temperature, pressure, inlet and outlet oxygen contents, stresses and electrochemical corrosion potential of each test specimen were continuously measured during the tests. The stress-strain curves of the SSRT run no. 2 is presented in Figure 2. All specimens fractured in the gauge section along the interface between Ni-insert and stainless steel TIG weld. The interface between CuCrZr alloy and Ni-insert was deformed but no cracking was observed along the interface. Fracture mode was mostly ductile and partly interface type of fracture close to or along interface between Ni-insert and stainless steel TIG weld. The facture path is close to TIG weld due to both heavy corrosion and large grain size in heat affected zone of Ni-insert Thermal fatigue behaviour of small scale FW mock-ups In order to provide input for the design and fabrication of the full-scale components the fabrication and testing of both small and full-scale mock-ups of ITER primary first wall modules has been carried out. The main objectives are to compare the performance of armour materials e.g., mainly Be alloys and heat sink materials such as CuAl25, CuCrZr and powder CuCrZr alloys after the different manufacturing processes such as Hot Isostatic Pressing (HIP) and brazing. An important aspect is that the joint properties, Cu/316L(N) and particular Be/Cu, complement the results of the mechanical characterisation of these joints. The full, non-destructive examination of all the Be/Cu, Cu/Cu and Cu/SS joint interfaces have been carried out by ultrasonic methods before and after the thermal fatigue tests in order to check the performance of the mock-ups and to monitor possible crack propagation due to thermal fatigue tests. The general conclusion of the obtained results is that the small scale mock-ups with beryllium tiles survive heat loads of about 0.75 MWm -2 for cycles and up to 1.5 MWm -2 for 1000 cycles. Those mock-ups without beryllium tiles survive heat loads of about 5 MWm -2 for 1000 cycles and first defect indications on stainless steel tube to copper alloy interfaces appear after several hundreds of cycles at 7 MWm -2. These heat loads are well above the expected nominal heat loads of the ITER primary first wall panels. It is also evident that the CuCrZr mock-ups show a higher margin against high heat loads than CuAl25 mock-ups do. It is also noted that the mechanical and fracture properties of copper alloys and their HIP joints also indicate a better performance for CuCrZr when compared to the CuAl25 alloy. An important observation is also that manufacturing quality of those components which have copper alloy to stainless steel joints is generally relatively good. On the other hand, manufacturing of components with beryllium to copper alloy joints is more challenging and requires more precise quality assurance procedures. 3.2 In-Reactor Material Testing under Neutron Flux In the past, it has been a common and wide spread practice to carry out post-irradiation testing in order to assess the effects of neutron irradiation on the mechanical properties of irradiated materials. A large number of such experiments have been carried out to determine the degradation of mechanical properties as a function of irradiation dose and temperature. These results are then used to evaluate the performance and lifetime of structural components in the dynamic loading conditions of a nuclear or thermonuclear reactor. The utilization of the post-irradiation results in this evaluation implicitly assumes that the kinetics, as well as the level of damage accumulation in the specimens irradiated in the unstressed condition, is the same as in the case of the structural components of a reactor. It is therefore of vital importance to recognize that in the case of deformation of reactor components, the damage accumulation and its impact on mechanical response will be controlled by two interactive kinetics operating simultaneously. The displacement damage rate will drive the kinetics of defect accumulation and the kinetics of dislocation build up will be driven by the applied strain rate. The effects of the dam- 64

75 age rate on defect accumulation and its spatial distribution will be continuously modified by the rate of dislocation generation and their motion by the applied strain rate and the corresponding stress level. Clearly, this is a vastly different and far more complicated situation than that existing during a post-irradiation test. It is therefore relevant to consider as to whether or not the post-irradiation tests can be taken to represent the deformation behaviour of structural materials under the service conditions of a fission or fusion reactor. In order to carry out in-reactor mechanical testing experiments, special test facilities consisting of pneumatic loading modules and a servo-controlled pressure adjusting loop were designed, constructed and calibrated. The pressure adjusting loop operates on a continuous flow of helium gas through an electrically controlled servo valve. The irradiation rig was designed to accommodate two complete test modules such that two independent mechanical tests can be performed at the same time. In addition, reference specimens were attached close to each test module which were irradiated in unstressed condition but with the same neutron flux and at the same temperature as for the in-reactor mechanical test specimen. These reference specimens were used for the post-irradiation (out of reactor) mechanical tests and microstructural investigations in the as-irradiated and post-irradiation deformed conditions. Before presenting the actual results of the in-reactor mechanical tests it is only appropriate to clarify the experimental conditions under which the materials response has been determined. Furthermore, this clarification is crucially important in order to understand and appreciate fundamental differences between the traditional post-irradiation tests and the present in-reactor mechanical tests. The post-irradiation mechanical test is performed, for instance, on a specimen who has been already irradiated (in the absence of any applied stress) at a certain temperature and to a certain displacement dose level but in the absence of applied stress. In other words, the specimen, prior to the commencement of the mechanical test has a given defect microstructure which has evolved in response to the irradiation temperature and the dose level but in the absence of applied stress. The stress-strain response measured during a post-irradiation test must reflect, therefore, the deformation behaviour of the frozen-in irradiation-induced microstructure. In the case of the in-reactor mechanical test, on the other hand, the situation is entirely different. Here, the test specimen is loaded in the fully annealed condition and begins to deform when it has a relatively low density of irradiation-induced defect clusters. Hence, the material will begin to deform in a manner similar to the unirradiated virgin material with a characteristic low yield stress. However, as test continues with a constant damage rate and strain rate, the density of both the irradiation-induced clusters (interstitial loops and clusters and vacancy SFTs) and deformation-induced dislocations will increase with increasing displacement dose (and strain). In other words, the measured stress response in these experiments includes the time (i.e. dose) dependence of both work hardening and radiation hardening Tensile tests The basic principle of the tensile test module is based on the use of a pneumatic bellows to introduce stress and a linear variable differential transformer (LVDT) sensor to measure the resulting displacement produced in the tensile specimen. The movement of the bellows is controlled by LVDT sensor which also gives the feedback signal for the servo controller. The outside diameter of the module is 25 mm and the total length of the module together with the LVDT is 150 mm as shown in Figure 1. A special calibration device was used to calibrate the applied gas pressure in the bellows with the actual load acting on the tensile specimen. A two-step calibration procedure was implemented where in the first step the characteristic stiffness of the bellows together with friction forces of the moving parts of the module were determined and in the second step the load induced on the tensile specimen by the applied gas pressure was measured directly by a load cell. The accuracy of the load calibration is approximately±1%oftheactual value of the stress resulting from the applied pressure causing the displacement in the specimen up to 1.3 mm. The tensile test was performed under a constant displace- 65

76 ment rate where the servo controller compares the actual LVDT signal to the set value and close/open the servo valve to induce the movement of the bellows by increasing/decreasing the bellows pressure. To accommodate the test modules and the necessary instrumentation to perform the uniaxial tensile test in the reactor, special irradiation rig were designed and constructed. The irradiation rig was manually inserted into the open tube at position G60 of the BR-2 reactor core during the steady state operation of the reactor. As the rig was lowered in the open tube, the temperature of the test modules increased rapidly due to a gamma heating power of 4.4 Wg -1 and the stagnant reactor pool water close to test specimens reached an equilibrium temperature of about 363K within a couple of minutes. Note that the test temperature of each test module and the specimen was measured and recorded separately. The test specimens were not loaded during this temperature transient. Both test specimens (and the corresponding reference specimens attached to each module) received a neutron flux of 1.7 x n/m 2 s (E > 1 MeV) corresponding to a displacement damage rate of 4.2 x 10-8 dpa/s (NRT displacement per atom). For comparison purposes, the tensile stress-strain curves for the unirradiated as well as post-irradiated copper specimens tested at the irradiation temperatures are also shown in Figure 2. It is worth noting here that the specimens used in the post-irradiation tests were irradiated in the same test module used for the in-reactor Test No. 1, 2 and 3. The results illustrate the following significant differences in the effect of irradiation on the mechanical response between the conventional post-irradiation tensile tests and the dynamic in-reactor tensile tests: (i) The post-irradiation test of specimens irradiated at 363K even to a dose level of as low as dpa exhibits the occurrence of a yield drop showing a clear transient between the elastic and plastic deformation regime. In the case of the in-reactor Test Nos. 1 and 3, on the other hand, there is no indication of such a transient. It is interesting to note though that in the case of the Test No. 2 which was given a pre-loading irradiation to a dose level of only 8.3 x 10-3 dpa, the stress- Pressure Inlet Bellow Tensile Specimen Linear Variable Differential Transformer Thermocouple Thermocouple Dosimeter Thermocouple Figure 1. Simplified layout and operational features of the test module used in the in-reactor tensile tests and the final assembly of the complete test module in the instrumented irradiation rig. 66

77 strain curve does, in fact, show a tendency for the occurrence of such a transient. (ii) The post-yield hardening behaviour in the post-irradiation tensile test is also distinctly different from that in the case of in-reactor tests. Although the rate of hardening is clearly higher in the in-reactor tests, yet the level of maximum hardening (i.e. ultimate tensile strength) remains significantly lower than that in the post-irradiation tests. (iii) In the case of the in-reactor tensile tests, both the yield stress and the level of the maximum hardening are dependent on the pre-loading displacement dose level. The lower this dose level, the lower is the yield stress as well as the maximum hardening. (iv) The uniform elongation measured in the in-reactor tests is dependent on the pre-loading dose level. The lower the pre-loading dose level, the larger is the uniform elongation. Figure 3 shows a representative example of the microstructure observed in the TEM discs taken from the region close to the crack in the tensile specimen used in the Test No. 1. The visible microstructure shows fairly homogeneous distribution of dislocation segments, black dots which are most probably SIA clusters and some loop like structure (which is expected to be of interstitial type). The presence of the high density of small SFTs in this specimen is not visible in this relatively low magnification micrograph. It is interesting to note that practically all dislocation segments are curved and show some limited amount of movement. However, there is no indication of any extensive dislocation motion and dislocation-dislocation interactions, causing segregation of dislocations in the form of dislocation walls. This is in a dramatic contrast to dislocation behaviour observed in the specimens deformed in the unirradiated condition. Figure 2. Engineering stress-strain curves of OFHC Cu for the in-reactor tensile tests carried out at 363K and 393K. The tensile curves for the post-irradiation and unirradiated tests carried out outside of the reactor are also shown. Note that in these in-reactor tests, the stress-strain curves represent the combined effects of strain hardening due to the applied stress and radiation hardening due to increasing level of displacement dose. The pre-deformation doses were 8.1x10-4 dpa, 8.3x10-3 dpa and 1.9x10-5 dpa for in-reactor tests No1, 2 and 3, respectively. 67

78 Figure 3. An example of the dislocation microstructure in the in-reactor deformed specimen in the Test No. 1 at 363K. (a) The homogeneous nature of spatial distribution of deformation induced dislocations and an almost complete lack of dislocation segregation in the form of dislocation walls and (b) examples of cleared channels. The microstructural evolution during the in-reactor tests is also very different from that observed in the post-irradiation tested specimens. The most significant difference is that in the specimens used in the in-reactor tests, the dislocation generation is found to occur throughout the whole specimen in a relatively homogeneous fashion. In the case of post-irradiation tested specimens, on the other hand, practically no dislocations are generated in the volume between cleared channels. In other words, dislocations are generated only at the sites of stress/strain concentration and are responsible for the localization of the plastic flow in the form of cleared channels. A close analysis of the mechanical response and the post-deformation microstructure of the in-reactor tested specimens reveals an interesting (from a scientific point of view) and an important (from a technological point of view) effect of the pre-loading dose level not only on the deformation behaviour but also on the lifetime (i.e. the total plastic strain prior to the initiation of failure) of specimens exposed concurrently to cascade damage and applied stress. The yield stress and the maximum flow stress of pure copper decrease with decreasing the pre-loading dose level whereas the rate of strain hardening increases with decreasing pre-loading dose level as can be seen in Figure 4. The results also illustrates that the irradiation-induced increase in the flow stress comes to saturate at a lower level and at a lower displacement dose with decreasing pre-yield dose level. In other words, when the plastic deformation is induced during very early stages of irradiation, (i.e. at low levels of pre-yield dose), the effect of irradiation on hardening is considerably reduced. Consequently, the specimen under these conditions continues to deform homogeneously but without any further increase in hardening and without initiating plastic flow localization and yields a considerably higher uniform elongation (i.e. longer lifetime). Similar in-reactor tensile test have been performed also for CuCrZr alloy. Although the deformation behaviour of the prime aged CuCrZr during in-reactor tensile test is, in many respects, very similar to that of pure copper, two significant differences should be noted. First, in the case of CuCrZr alloy, the irradiation induced increase in the flow stress reaches a maximum at a very low strain (dose) level and then begins to decrease as a function of strain. Second, for a given pre-loading dose level, cracks are formed in the CuCrZr specimen at a much lower strain (dose) than that 68

79 OFHC COPPER In-reactor Test =T test T irr f = (irr.) (unirr.) (Mpa) f f Test No.1, 363K Test No.2, 363K Test No.3, 393K?? /?t = 1.3x10-7 s -1 (Test No.1 and 2)?? /?t = 1.0x10-7 s -1 (Test No.3) TRUE PLASTIC STRAIN (%) Figure 4. Increase in flow stress, Δσ f, as a function of true plastic strain for in-reactor tests at 363 and 393K. Note that while the increase in the flow stress, Δσ f, decreases with strain, the magnitude of the increase in the flow stress (at a given strain (dose) level) increases with increasing level of pre-yield displacement dose. Furthermore, the increase in the flow stress beyond a certain strain level (i.e. dose) decreases with increasing strain (dose) and then tends to saturate (e.g. Test No. 1 and 3). in the case of pure copper. The reasons for these differences are not at all clear at present. Judging from the cleared channel formation behaviour it seems plausible that both of these effects may have their origin in the factors controlling the initiation of the cleared channels. Most recently in-reactor tensile experiments were also carried out for pure Fe and FeCr alloy. The preliminary results indicate that in case of pure Fe maximum flow stress is reached at relatively low strain and dose levels whereas Cr alloying seem to induce very strong hardening already at low strain and dose levels. The detailed analyses of the experimental results are still under progress Creep fatigue tests The basic principle of the creep fatigue test module is similar to tensile test module, e.g., the use of a pneumatic bellows to introduce stress and a linear variable differential transformer (LVDT) sensor to measure the resulting strain produced in the fatigue specimen and to give the feedback signal for the servo controller. In order to produce both the tensile and compressive loads required in cyclic creep fatigue tests, the loading module had two bellows with two independent pressure adjusting loops. The simplified layout of the creep fatigue module is shown in Figure 5. The outside diameter of the module is 53 mm and the total length of the module together with the LVDT is 250 mm. The LVDT sensor was connected to special flanges which were further attached to the fatigue specimen gauge section. The arrangement for strain measurement consisted of the LVDT sensor, flanges, posts and connection clips. Two of the posts were mechanically connected to the upper flange and to the LVDT sensor s body. The other two posts were mechanically connected to the lower flange and to the LVDT sensor s plunger. The flanges were connected to the posts with spring loaded connection clips in order to avoid free clearance between the connecting parts. 69

80 Figure 5. Simplified layout and operational features of the test module used in the in-reactor creep fatigue tests and the final assembly of the complete test module in the instrumented irradiation rig. Strain controlled in-reactor creep fatigue experiments were carried out for CuCrZr alloy using various strain amplitudes, e.g., 0.25, 0.35 and 0.5% with 10 and 100 seconds hold times at maximum tensile and compressive strain values. The irradiation rig was again manually inserted into the open tube at position of the BR-2 reactor core during the steady state operation of the reactor. In this case the loading modules were cooled by flowing reactor pool water and test specimens reached an equilibrium temperature of about 353K within a couple of minutes. Both test specimens received a neutron flux of about 7x10 17 n/m 2 s (E > 1 MeV) corresponding to a displacement damping rate of about 6.5 x 10-8 dpa/s (NRT displacement per atom). The cyclic loading was started within 10 minutes after the irradiation rig was inserted in the reactor core corresponding to about 10-5 dpa pre-loading dose for both test specimens. Typical example of creep fatigue cycle with controlled strain amplitude of 0.5% with 100 second hold time and frequency of 3.3x10-3 Hz is shown in Figure 6 with corresponding stress-strain curves. During the first hundred cycles some amount of hardening followed by continuous softening induced by simultaneous effect of cyclic loading and irradiation defect accumulation was observed. Although the observed softening was different compared to that in unirradiated condition, the cycle life of CuCrZr alloy in in-reactor creep-fatigue tests was consistently longer in the tested strain amplitude range from 0.25 to 0.5% as indicated in Figure 7. The microstructural examination is still under progress but the preliminary results indicate that there is no indication of any dislocation-dislocation interactions, causing segregation of dislocations in the form of dislocation walls and that the amount irradiation induced defects are drastically reduced due to mechanical cyclic loading. This is in accordance with in-reactor tensile test results of copper and CuCrZr alloy but in clear contrast to dislocation behaviour observed in the specimens cyclically deformed in the unirradiated condition. The in-reactor tests have clearly demonstrated the feasibility of performing well defined and controlled dynamic deformation experiments in a fission reactor. The validity of the experimental results obtained during in-reactor tensile testing has been established so that these results can be compared and contrasted with the results obtained in conventional post-irradiation tests carried out using standard mechanical testing machines outside of a reactor. 70

81 STRAIN (%) STESS (MPa) TIME (s) Figure 6. Typical strain controlled creep fatigue cycle (a) controlled strain and corresponding stress response (b) stress-strain cycles during in-reactor creep fatigue test of CuCrZr alloy showing the simultaneous effects mechanical cyclic loading and irradiation defect accumulation on creep fatigue cycle life. 71

82 CuCrZr Ht1 Creep fatigue Strain amplitude 0.5% Cycle time 300s t/c hold time 100s MAX. STRESS AMPL. (MPa) In-reactor Unirradiated NUMBER OF CYCLES (N) Figure 7. Stress amplitude as a function of number of cycles for in-reactor and unirradiated out-of-reactor creep fatigue tests for CuCrZr alloy. 3.3 Advanced Welding of ITER Vacuum Vessel Sectors VTT Industrial Systems V. Kujanpää (Project Manager), M. Karhu and T. Jokinen VTT has carried out several EU-ITER tasks during the FUSION technology programme in which a suitable welding method has been considered for the manufacturing of vacuum vessel for fusion reactors. The tasks have been involved hybrid Nd:YAG laser and electron beam welding and, although the walls of vacuum vessel are made of 60 mm stainless steel, they has shown great potential of being the joining method, except for some additions. One reason to use Nd:YAG laser is the enormous size and weight of the vacuum vessel, which means that the vessel will be built on-site. This, together with the geometry of the vessel, sets the positional welding requirements for the processes used and the Nd:YAG laser seems to be usable because of the flexible transmit of laser light via optical fibre. In addition, the low total heat input, which is typical for laser welding, is an advantage when joining large austenitic stainless steel structures. Recently, some new potential laser types, fibre and disc lasers have been available. They can be also possible choices for assembly welding. Electron beam welding must use vacuum. Therefore it is not necessarily the potential choice for assembly welding, but when welding vacuum vessel sectors or parts of it, they are very potential. Numerous hybrid Nd:YAG laser (Figure 1) and electron beam welding experiments have been carried out during different tasks, concentrating on the welding parameters on keyhole and conduction-limited welding, multi-pass welding procedure, narrow gap configuration and root control of electron beam welding. In addition, intersector welding robot (IWR) developed by Lappeenranta University of Technology (see Chapter 4.3) was tested by Nd:YAG laser welding. The main highlights of these experiments are reported in this chapter

83 On single pass experiments with a plate thickness of about 6.5 mm the characteristic factors for both processes were studied. The experiments showed the gap bridging ability, which can be used also for multi pass welding. In the experiments keyhole welding was observed in both laser and hybrid welding in spite of the existence of the arc. Due to that the welding speed could be raised from 0.5 m/min to 0.9 m/min. The main point in the multi-pass experiments was to study the applicability of the processes to a narrow groove. It was noticed that applicability is dependent on the accessibility of the energy to the bottom of the groove. From such reasoning the minimum groove angle can be defined. Experiments showed that laser welding with filler wire could be achieved with a groove angle of 8, while stable hybrid welding is possible for the first filling pass with an angle of 10. The welding procedure shown consists of a root pass with the laser and filler wire and welding of the upper sections of the groove with more efficient hybrid welding. Use of this procedure is flexible since it is possible to use the same machine for filler wire feeding, i.e. with the root weld arc is turned off, while for the upper sections is turned on. Figure 1. Experimental set-up of Nd:YAG hybrid laser welding Keyhole laser and hybrid laser welding of thick sections The laser welding with filler wire and the hybrid welding were used in welding of thick section austenitic stainless steel for a very narrow gap and using a multi-pass technique. By the help of the characteristic features of the Nd:YAG laser both processes were able to enter the energy inside the narrow groove. By using the narrow groove with procedures shown in this work the number of passes can be decreased significantly and the efficiency of the welding can be increased. The maximum thickness welded was 20 mm for the laser welding with filler wire and 30 mm for the hybrid welding, Figure 2. The main factor for the thickness is the groove geometry and the groove angle, which together with the thickness determine the air gap at the surface of the material and vice versa. Applicability of the processes used is defined by the maximum air gap in the joint. The procedure introduced is based on the use of a 3 kw Nd:YAG laser with a focusing optic of 200 mm. Use of the procedure with different Nd:YAG, Figure 2. Example of multi-pass Nd:YAG hybrid laser weld (thickness 30 mm) made by keyhole mode. 73

84 disc or fibre lasers is possible, but the groove geometry will vary according to the focusing optic. Continuously the lasers are developed and new types of lasers are introduced. The development is aimed at the higher laser power with better beam quality. This allows in the future the use of even narrower grooves and thus a more efficient process. In the point of view of the industrial feasibility, laser welding has suffered high demands of groove manufacturing and fixing of the components to be welded. By using filler wire in the process the feasibility increases together with the gap bridging ability. The use of hybrid welding extends the joint tolerances to a new level together with the increased welding speed. Increased welding speeds is not the only reason for the interest of industrial use of these processes, but also reduction of the distortions and thus a time consuming work phase after welding. The same applies for multi-pass welding of thick section steels. method in welding of filling passes even a gap width of near 11 mm could be bridged and filled. This improved gap bridging ability concerning welding of filling passes based on the feasibility of conduction limited hybrid method to make filling passes wider than what can be produced by using a keyhole-mode hybrid welding alone. In the multi pass welding experiments approx mm wide, partially chamfered narrow gap grooves could be effectively filled; Using one pass per layer technique, approx. 24 mm deep groove geometries could be filled with 5 to 6 filling passes. That means on average a vertical fill-up of approx. 4 4,8 mm per filling pass Conduction-limited hybrid laser welding of thick sections Series of conduction limited hybrid welding tests for austenitic stainless steel test joints were carried out using a combination of 3 kw Nd:YAG laser and MIG welding process. In this conduction limited hybrid process variation, a power density of a laser beam spot was purposely dispersed by using defocusing. In practice, a power density of laser beam was kept inside ~10 4 W/cm 2 range, which led the hybrid process towards to conduction limited regime. Welding tests started with preliminary study, in which a basic knowledge concerning the influence of parameter variations on process behaviour was studied. Parameter knowledge gained from the preliminary study was applied in the further experiments. In those experiments the feasibility to fill and bridge larger groove gaps than can be welded with keyhole-mode hybrid welding was studied. Above tests included also narrow gap multi pass applicability tests for 30 mm thick test pieces, Figure 3. The experiments showed that with using conduction limited hybrid Figure 3. Example of multi-pass Nd:YAG hybrid laser weld (thickness 30 mm) made by conduction-limited mode. The results of distortion measurements confirmed the pre-assumption that distortions in conduction limited hybrid welding will be elevated compared to the ones generally occur in keyhole hybrid welding method. This arises from the fact that used groove volumes were larger than the ones normally applied in keyhole mode hybrid welding. Larger/wider groove geometries need more melted filler material and more dispersed welding energy which in turn results in increased welding energy input and increased distortions. 74

85 During the multi pass welding experiments of 30 mm thick test pieces, it was realised that hot cracking had been occurred in the certain upper filling passes of every multi pass weld. By large, the hot cracking occurred in the filling passes which depth to width ratio were in the range of 1,3 1,7. It is known that e.g. too large depth to width ratio of produced weld pass combined to high rigidity of welded structure could cause increased hot cracking sensibility. Too large depth to width ratio of a produced filling pass could be avoided by controlling the process parameters, but the effect of high overall rigidity of welded structure on the sensibility of hot cracking is a less known factor. A new task has been started concerning this item Root control of electron beam welding Closed loop feed-back control system was developed and connected to the EB-welding machine, Figure 4. The performance of developed control system was checked with welding tests. The evaluations of the welding tests were made visually from the test welds and from the welding data derived from PC s data collection unit. The test pieces were austenitic stainless steel plates which thickness was whether linearly increased or decreased in ratio of 1:40 in the middle part of the test piece. An increment and a decrement of thickness was chosen to be altered within 2 mm, which will simulate possible thickness alteration of the root face, resulted from manufacturing of very large-size work piece. Welding tests showed that the control system managed to respond quite well as the thickness of the test piece was grown from 10 mm to 12 mm, Figure 5. The system raised the amount of beam current quite a linearly in relation to the growth of the plate thickness. When the thickness of the test piece decreased linearly from 10 mm to 8 mm, the control system kept beam current values in quite a constant level during welding. The explanation to the latter case may be that the level of the pre-cho- Figure 4. Principle of the developed root control of electron beam welding system. Figure 5. Example of root side image of electron beam weld when using the root control system developed (weld thickness varied from 10 to 12 mm). 75

86 sen through-current (3mA) is so high that it widens the parameter window, where a successful welding can be taken place. When the thickness of the test piece is decreasing like in the above case, no remarkable changes in beam current is needed because process stands inside the parameter window. During the additional welding tests and the development work of the control system, it was found that the EB-welding machine was tended to be quite sensitive at the start of the welding sequence. If the beam current request of the control system (ma/second) is too high at the start, the safety system of the welding machine will shut the beam off. A beam current rise ramp was implemented to the control system in order to prevent this instability occurred between the welding machine and the control system. As a whole, the carried welding tests gave promising results on the functionality of the control system, as well as the applicability of the concept itself. For a proposal of further developments it can be mentioned that it may be advisable to improve the reliability of the system with switching from the current PC-based implementation to an embedded, microcontroller based control system. In addition, it would be useful to carry out further welding tests, where the control system is tested with a longer weld length and at different welding positions Testing of a intersector welding robot (IWR) by laser welding Applicability of seam tracking equipped inter sector welding robot (IWR) for Nd:YAG- laser welding was investigated. The applicability study was executed with applied welding experiments. In those experiments, this hydraulic driven 5-DOF parallel robot augmented by seam tracking system was employed in laser welding of pre-engineered test pieces with 1 degree tilted weld seam regarding y- and z-axis. The experiments were carried out such that three test sets were welded with using different methods (reference method, teaching method, on-line method). The results from the experiment of teaching method showed that few weld discontinuities occurred. The on-line method experiment gave the best results, although one section of discontinuity in weld seam occurred. This indicates that at this stage of development IWR together with control system does not fully meet the issued requirements. Moreover, it was realised, that the used seam tracking/control system needs to be improved in order to get out the full potential of IWR. The results of the reference method experiment confirmed that with the used 200 mm focal length optic, the tolerance area of a positional accuracy of a laser beam focal in relation to joint preparation is more severe in y-direction than in z-direction. Further work needs to be carried out in order to completely fulfil requirements set by laser welding process, but the accomplishment achieved so far will certainly work as a strong basis when the concept is further developed. 3.4 Radiation Damage in EUROFER FeCrHe Thermodynamics University of Helsinki Kai Nordlund The long-term commercial use of a fusion reactor requires structural materials that can withstand bombardment from the 14 MeV neutrons produced in the D+T fusion reaction for time scales of tens of years without significant degradation of mechanical properties or activation by long-lived radioactive isotopes. Due to these reasons, special steels are under development that can meet the goals. In these steels, the Mo is replaced with W to reduce the activation, and the Cr content is high since Cr-rich steels have been found to have good radiation tolerance properties. The class of steels under development is known as reducedactivation ferritic-martensitic (RAFM) steels. The European candidate material is known as EUROFER. The mechanism by which the steels are damaged under long-term irradiation are not well under- 76

87 stood, and this can not be directly tested since there are no high-intensity sources of 14 MeV neutrons available (the IFMIF materials testing facility would provide one, but will not be available in many years). Moreover, the 14 MeV neutrons pose an additional problem when compared to fission neutrons with energies of a few MeV: their energy is high enough to initiate the production of nuclear reactions which lead to He production inside the materials. Due to these reasons, there is a significant effort to obtain scientific understanding of radiation damage and He effects inside RAFM steels, which could be used to guide the further development of the steels as well as possibly even the running conditions of the reactors. The starting point for any neutron-irradiation damage modelling effort is the study of the primary state of damage produced by displacement cascades in the relevant material. Molecular dynamics (MD) is well known to be a suitable simulation tool for the study of displacement cascades and the analysis of the mechanisms of formation and motion of interstitial atoms and their clusters, provided that a valid many-body interatomic potential is available for the system of interest. Within the FUSION programme, we have been involved, together with the Belgian and Swedish association, in the development of interatomic potentials for FeCr and FeCrHe with realistic thermodynamic properties and suitable for modeling the primary damage state produced by neutron recoils in reactors. We first developed an interatomic potential that correctly describes the heat of mixing in the Fe-Cr alloy for all Cr concentrations. To achieve this it was necessary to modify the conventional embedded-atom method (EAM) functional form to separate between s and d electron contributions to the embedding energy. This so called two-band model (2BM) potential was then used to examine radiation damage in FeCr alloys. Since experiments have shown that the Cr concentration plays a crucial role on the radiation hardness of steels, FeCr-alloys are the simplest alloys which can be used as model materials when studying the basic mechanisms governing their behavior under irradiation. Analyses of displacement cascades in Figure 1. Typical distribution of interstitials (red/dark spheres) and vacancies (light spheres) produced in a 20 kev cascade in FeCr. Fe 90 Cr 10 and Fe, with recoil energies ranging between kev, have been carried out, paying special attention to the behavior of the produced defects when using the 2BM potential constructed for FeCr alloys. An illustration of a typical damage distribution is given in Figure 1. Comparisons with results of simulations with another available FeCr potential and in pure Fe were also done, revealing that the 2BM potential applied to Fe 90 Cr 10 gives similar results with respect to the total damage production and defect clustering as in Fe. It does, however, estimate a smaller chromium content in interstitial atoms than the earlier potential (see Table 1) This concentration is likely to grow with time due to the attractive energies of the mixed dumbbell predicted by the potential. The quantitative difference between the Cr-SIA enrichment in the two models is attributed to the different description of single interstitial and interstitial cluster mobility predicted by the corresponding Fe potentials. We also developed interatomic pair potentials for He-Cr and He-Fe interactions that describe correctly the energy of a He impurity in different interstitial and substitutional sites in the Cr and Fe lattices. 77

88 Table 1. Production of defects in cascades in FeCr obtained using the earlier Chakarova potential and two different versions of the new two-band model (EMTO and PAW). The total production of Frenkel pairs and fraction of clustered defects is essentially the same in all models. However, the fraction of interstitial defects with Cr atoms in them differs significantly. The statistical uncertainty of the reported quantities is about 5% or less. Chakarova potential Two-band (EMTO) Two-band (PAW) Frenkel pairs Fraction of clustered vacancies Fraction of clustered interstitials Fraction of interstitials with Cr Fraction of clustered interstitials with Cr Fusion Neutronics VTT Nuclear Energy Petri Kotiluoto and Frej Wasastjerna For each fusion reaction, releasing 17.5 MeV, one neutron is produced. This means that a fusion reactor is a more prolific source of neutrons than a fission reactor of the same power, which produces approximately one neutron per 80 MeV. Moreover, in a fusion reactor, the neutrons carry 80 % of the reaction energy, compared with about 2.5 % for fission reactors. The neutrons from D-T fusion start at 14 MeV, so they are able to cause a greater variety of reactions than in a fission reactor. Therefore one definitely cannot neglect neutronics for fusion reactors. Instead, it is essential to calculate the reactions induced by fusion neutrons. One important reaction is the tritium breeding required for the fusion process, but there is a wide variety of other reactions. Neutrons cause heating, damage or embrittlement of reactor materials. Neutrons also activate materials, producing radioactive waste and complicating maintenance. Calculating the different reaction rates is necessary for design of a fusion reactor or a fusion material irradiation facility Radiation transport and shielding calculations Radiation transport calculations can be performed either with deterministic or stochastic simulation methods. Deterministic methods solve the transport equation for the average particle behaviour. By contrast, the stochastic Monte Carlo method does not solve an explicit equation, but rather obtains answers by simulating individual particles and recording some aspects of their average behaviour. Deterministic methods are in general faster, but cannot yet cope with large detailed 3D models, mainly due to large memory requirements encountered in such cases. Deterministic methods are also more or less approximate and are usually ill-adapted to coping with transport through voids. The stochastic Monte Carlo method duplicates a statistical process, such as the interaction of nuclear particles with materials, and can simulate the particle transport with great accuracy. The problem is that particle tracking is time consuming and in order to get answers, such as fluence or dose rates, with proper statistics, generally a huge number of particle histories have to be simulated. This is particularly true in deep penetration problems such as radiation transport through thick shield. Effective variance reduction methods, such as use of spatial importances to multiply some tracks and to terminate others, have to be used for such problems. In fusion neutronics calculations, Monte Carlo code MCNP or its special version McDeLicious have been largely used. An especially critical problem from the shielding point of view is to ensure low enough shutdown dose rates to permit the necessary hands-on maintenance operations, following the ALARA principle (As Low As Reasonably Achievable) of radiation protection and not to exceed dose limits of the personnel. To properly calculate the shutdown dose rate, the following must be done: first, a neutron transport 78

89 calculation to find the neutron flux causing activation; second, a calculation of activation; third, a calculation of the transport of gamma photons from the activated materials and of the dose rate produced by them. The calculation can be done by performing these steps separately, in what is called the rigorous two-step method (R2S), using two separate transport calculations, but it is also possible to combine the steps and use the so-called direct one-step method (D1S), in which the neutron and photon transport calculations are combined into a single n,p calculation in MCNP, using delayed gammas from activation instead of prompt gammas. Each method has its own advantages and disadvantages. Perhaps the most obvious disadvantage of the R2S method is the spatial discretization error. Instead of each delayed gamma being born at the exact location where the neutron producing it was absorbed, as in the D1S method, the neutron flux is averaged over each cell and then the resulting activation is calculated in a separate program, such as FISPACT. In the gamma transport calculation, the gamma source is taken to be constant in each cell, losing the detailed spatial information within the cell. Some work has been done to estimate the spatial discretization error in the R2S method. The D1S method, however, has been used most often in estimation of the shutdown dose rates ITER neutronics In ITER, the shutdown dose rate is the critical quantity from the viewpoint of shielding design for the equatorial and upper ports. The shutdown dose rate is dominated by activation caused by neutrons streaming along the gap between the port plug and the port walls. Calculating this streaming with a Monte Carlo program is a very challenging problem, and has required a lot of investigations. Shielding calculations have been performed for instance for helium-cooled pebble bed test blanket modules (designed for tritium breeding) and rf-heating systems of ITER. In addition, an investigation of the effects of certain design details on the gap streaming for a generic equatorial port plug has been made. In summary, it appears to be possible to keep the shutdown dose rate low Figure 1. Sabrina visualisation of the 3D MCNP model of ITER, showing also the Helium-Cooled Pebble Bed Test Blanket Module (HCPB-TBM) modelled by VTT (indicated with an arrow). enough if careful attention is paid to minimizing the gap streaming IFMIF neutronics The International Fusion Materials Irradiation Facility (IFMIF) will play an essential part in the global fusion research program, providing data about how well materials and components can stand intense neutron irradiation. IFMIF employs two continuous-wave linear accelerators each generating a 125 ma beam of 40 MeV deuterons striking a thick target of flowing liquid lithium. Neutrons are generated in the lithium target by the d-li stripping process and various other nuclear Li(d,xn) reactions. Since producing an intense flux of high-energy neutrons and exposing the materials and components to be tested is the whole purpose of the project, neutronics calculations are obviously an essential part of the design work. A special version of MCNP called McDeLicious has been developed in FZK/IRS. McDeLicious can simulate the neutron generation in the lithium target based on evaluated d + 6,7 Li cross-section data. VTT as a representative of the Association Euratom-Tekes has modelled the whole IFMIF 79

90 Figure 2. Elevation view of the IFMIF MCNP model with the horseshoe shield. test cell. Most of the calculations using this model have been done by FZK/IRS, but VTT has also done some calculations. Among other things we have evaluated the neutron field along the beam tubes, for further activation and dose rate calculations of the raster beam scanner system performed by UKAEA. Special attention has been paid to the dose rate evaluation on top of the IFMIF test cell cover, which represents a very demanding streaming and deep penetration problem though a complicated structure of ducts and shields. In order to solve this taxing problem, different variance reduction methods have been applied and studied. Although the difficulties in evaluating the dose rate on top of the cover have been so severe that the work is not finished yet, it is clear that, when IFMIF is operating, this dose rate will considerably exceed the limit of 100 μsv/h considered acceptable for hands-on work. However, preliminary results suggest that the shutdown dose rate will be acceptable. This is almost certainly the case 10 6 seconds after shutdown. How short a cool-down interval is acceptable remains to be determined Neutronics summary Fusion neutronics work has been performed in close collaboration with FZK/IRS. Different fusion neutronics calculations have been carried out for ITER, including for instance the modelling of the equatorial port plugs and test blanket modules and the corresponding activation calculations for evaluation of the shutdown dose rates. Much of our work has concentrated on optimisation of the dogleg geometry of the gap between the port plug and the port walls from the radiation shielding point of view, as the shutdown dose rate is dominated by activation caused by neutrons streaming along this gap. A large effort has also been the creation and updating of the global geometry model of IFMIF. Dose rate evaluation on top of the IFMIF test cell cover has probably been the most challenging task, demanding the use of sophisticated variance reduction methods for particle transport though the complicated structure of ducts and shields. Further work for the solution of such deep penetration and streaming problems with Monte Carlo 80

91 method would be required, as the obtained results for the dose rate evaluation on top of the test cell cover are still somewhat unsatisfactory. So far neutron fluxes have been calculated only above 0.1 MeV. However, work on implementing what we call the tally source method, based on a proposal by John Hendricks, is under way. The first results suggest that this method can provide flux data for all energies, albeit at the cost of great amounts of work and somewhat reduced accuracy due to discretization errors. In general, VTT has actively taken part in the neutronics modelling of both ITER and IFMIF facilities. Various EFDA tasks have been discharged on the subject matter, and separate development work for improving MCNP methodology and competence, e.g. optimising variance reduction, has been carried out. 3.6 Material Transport and Plasma-Wall Interactions VTT Materials Performance Jari Likonen (Project Manager), Tommi Renvall, Elizaveta Vainonen-Ahlgren Helsinki University of Technology (TKK) Markus Airila, Otto Asunta University of Helsinki (UH) Juhani Keinonen (Head), Tommy Ahlgren, Carolina Björkas, Kalle Heinola, Niklas Juslin, Pia Kåll, Kai Nordlund, Petra Träskelin Diarc Technology Oy Jukka Kolehmainen, Sanna Tervakangas Contributors to the EFDA JET workprogramme The ASDEX Upgrade Team Material transport and erosion/ re-deposition in JET and AUG tokamaks One of the major challenges that ITER has to face is the problem of tritium retention in the vessel wall. This problem is compounded by the fact that tritium is easily co-deposited with carbon, a material needed for some of the main plasma-facing components in ITER. Therefore, understanding the erosion, migration and re-deposition of carbon in tokamak edge plasmas is of crucial importance. Erosion and deposition have been investigated after experimental campaigns both at JET and ASDEX Upgrade with marker tiles and various post mortem surface analysis techniques. The amount of deposition has been quantified at the inner divertor both at JET and AUG, but the amount of erosion at the outer divertor is not yet known. Studying carbon migration is not easy: tokamak experiments consist of a multitude of different plasma configurations and modes of operation, all of which can exhibit very different impurity dynamics. Moreover, the migration patterns that are observed in the divertor are integrated over long experimental campaigns. Therefore, only little about carbon dynamics can be determined from post-mortem analysis of the wall components. Trace element studies provide a means to circumvent this problem: a known amount of the trace element is puffed into the plasma right before the machine is opened, and the plasma is operated only in one specific mode of operation. Several divertor and wall tiles are then removed, and the amount and deposition profile of the trace element measured. In this way, quantitative information about carbon migration and deposition can be obtained. Such experiments have been carried out both at JET, and ASDEX Upgrade using 13 Casthe trace element. However, measurements as such cannot reveal the mechanisms responsible for the carbon migration. Therefore, it is useful and necessary to simulate these experiments with impurity transport codes, such as ERO and DIVIMP. Such simulations help to identify the basic physical mechanisms behind the carbon migration Surface analyses of JET and AUG divertor tiles Several sets of JET and AUG divertor tiles from different experimental campaigns have been analysed using secondary ion mass spectrometry (SIMS) and time-of-flight elastic recoil detection analysis (TOF-ERDA) in Analysed JET tiles were exposed in during the MkIIGB phase and in during the MkIISRP campaign. AUG divertor tiles after ev- 81

92 ery campaign in 2003, 2004 and 2005 were analysed. SIMS analysis was made with a double focussing magnetic sector instrument (VG Ionex IX-70S).A5keVO 2+ primary ion beam was used and the ion currents of secondary ions 1 H +,D +, 9 Be +, 10 B +, 12 C +, 13 C +, 58 Ni + and 183 W + were profiled. Some selected samples that had been analysed by SIMS were also measured with TOF- ERDA to obtain elementary concentrations at the near surface region. In the measurements, the 5 MV tandem accelerator (EGP-10-II device) of the University of Helsinki was used with a 53 MeV beam of 127 I 10+ ions Erosion and deposition During the campaigns in JET was operated with a MkII-SRP divertor shown in Figure 1. A new set of markers were installed in the MkII-SRP divertor for These comprised a poloidal set of tiles coated with stripes of W with a thickness of 3 μm, and have provided new information on erosion and migration of impurities at both the inner and outer divertor. The analysis of the films in the outer divertor has been very important for specifying the W-coated tiles to be used for the ITER-like wall experiment planned for JET. Figure 1. JET MkII-SRP divertor. Under normal field conditions, impurities eroded from the main chamber are transported around the SOL and deposited where this intersects the inner divertor. After the operation a duplex film structure had been found for deposits at the inner divertor tiles 1 and 3, with an inner film of high Be/C and an outer film with much lower Be content. The outer film also contained a much greater D concentration than the inner film. Figure 2 shows SIMS profiles for 12 C, 13 C, Be and Ni through the film on sample from the top of tile 3. The Be/ 12 C ratio is much lower in the outer one-third of the film to that in the inner part of the film, and the behaviours of Be and Ni are similar. Figure 3 shows the SIMS depth profiles from top of tile 3 that had been in the vessel from 1998 to The profiles at the inner part of the film are similar to those seen in Figure 2 and probably correspond to film deposited in the period , implying the outer part corresponds to film deposited The majority of the deposits again contain high concentrations of Be. The Be/C ratios are comparable with the levels seen in the inner layer formed in (see Figure 2) and in earlier campaigns. It is clear that chemical sputtering of C at the inner divertor returned for the majority of the operations from the reduced rate at the end of the campaign to its earlier value, despite the similar JET wall temperature. Outer divertor tiles 7 and 8 removed in 2001 showed no net deposition, and markers had largely been eroded. New divertor tiles with W-markers were installed in 2001, and tile 7 removed in 2004 showed a more complex picture. There appear to be thin deposited layers on the upper part of the tile 7 and some evidence of the 3 micron poloidal W marker stripe remained visible at all points. Surface analyses of the W stripe show that up to 70% of the film has been sputtered away from the area of the tile corresponding to the most common strike point position which is towards the bottom of the tile. Part of the deposition occurred during the 13 C puffing experiments, but deposition also occurred on tile 7 during the reversed field campaign. The appearance by eye of the W stripe on outer divertor tile 8 exposed in was very similar to when it was mounted in However, as for tile 7, detailed analysis of the stripe has shown that the majority of the W marker layer has been eroded on the upper part of the front face of the tile and from the outer, horizontal, surface. Even though the W layer is still continuous over 82

93 Figure 2. SIMS depth profiles from top of tile 3 exposed Figure 3. SIMS depth profiles from top of tile 3 exposed the lower front face of the tile, the layer is thinner than when mounted in JET. Long term erosion/deposition data were also measured on AUG upper divertor tiles (Figure 4) with W and C marker stripes after 2004 campaign. Depth profiles obtained from C and W marker stripes on the limiter tile PSL are presented in Figure 5 (a) and (b), respectively. In the case of C stripe (Figure 5 (a)), layer deposited by plasma during the campaign can be observed near the surface. Then the original C film with following Re interlayer can be seen. For the W stripe (Figure 5 (b)), layer deposited by plasma and 83

94 by the divertor configuration. In support of this conclusion, results obtained in earlier experiments at JET can be referred. The deposition rate was found to be and for MkIIA and MkIIGB divertor configurations, respectively. Both AUG lower and JET MkIIGB divertors have quite closed geometry because of the dome baffle and the septum, respectively. Results for the deposition rate for these divertors agree with each other very well. In the other hand, both AUG upper and JET MkIIA divertors are quite open. Result g/s found for the AUG upper divertor after 2004 campaign is closer to the one determined for JET MkIIA divertor. It seems that obtained carbon deposition rates depend rather on the divertor configuration than the machine size, however, the reason for this is not clear Material transport in scrape of layer (SOL) Figure 4. Cross-section of the ASDEX Upgrade vacuum vessel together with the magnetic surfaces. Numbers correspond to the divertor tile numbers. original W film can be seen. In both cases films deposited by plasma are enriched with hydrogen, deuterium and boron. Furthermore, presence of carbon was observed in the layer deposited by plasma on the W stripe. The deposition rate of carbon by plasma was obtained to be g/s. Density of the deposition layer was assumed to be 1.8 g/cm 3. Total time of plasma in the top divertor configuration during the campaign was 231 s. Similar study on the long term carbon erosion/deposition was performed earlier with marker stripes on the lower divertor tiles after 2003 campaign. The deposition rate of carbon by plasma was obtained to be g/s. Density of the deposition layer was assumed to be 1.8 g/cm 3. Total time of plasma in the bottom divertor configuration during the campaign was 4944 s. It can be suggested that the deposition rate is affected Prior to the 2004 shutdown, an experiment was carried out to provide specific information on material transport and SOL flows observed at JET. 13 CH 4 was injected into the outer SOL from injection points between tiles 7 and 8 in the last day of discharges using one type of discharge only. The purpose of the experiment was to determine how the 13 C is transported around the SOL, and where it is eventually deposited. The amount of 13 C has been measured over a complete poloidal scan of the divertor using SIMS and Rutherford backscattering (RBS) techniques (see Figure 6). RBS measurements were made at the University of Sussex. The largest 13 C amount was found on tile 7 just above the strike point and at the top of the tile and on the apron of tile 8. The deposition on tile 7 is due to local downstream impurity drag. The deposition on the apron of tile 8 could perhaps be due to upstream transport and far SOL downstream flow. There may be a leakage path for the puffed gas from the supply manifold to the top of the outer divertor, so that gas may enter the far SOL close to the deposition site. The amount at the inner divertor is relatively small but there is somewhat higher level on the floor tiles. Integrating the amounts of 13 C through inner divertor wall tiles and extrapolating to the 84

95 H Intensity (s ) HD B C 187 Re a) Depth ( m) H Intensity (s ) HD B C 187 Re b) Depth ( m) Figure 5. Long term depth profiles on C (a) and W (b) marker stripes measured on the upper divertor tiles after 2004 campaign. Presented spectra measured from the limiter tile PSL. 85

96 Figure C distribution in the JET divertor. Z-coordinate is along the divertor tile surfaces and origin is at the horizontal part of tile 1. entire inner divertor wall of JET (assuming uniform concentrations toroidally), gives 26% of the injected amount of 13 C. The total percentage is only a small fraction of the input, and much less than found after the 2001 puffing when 13 Cwas puffed from the top of the vessel. At the end of the 2003, 2004 and 2005 experimental campaign 13 CH 4 was puffed into the AUG torus from the outer mid-plane at one toroidal location during identical discharges in hydrogen. Distribution of 13 C on the lower divertor together with strike point position from magnetic reconstruction during last 13 shots of the 2005 campaign (bottom single null L-mode discharges were used) as a function of V-coordinate is presented in Figure 7. V-coordinate represents the distance from the private flux region along the tiles surface and describes poloidal positions (V=0 corresponds to the top edge of the dome baffle tile 9A). During the 13 C puffing experiments the strike points were at the inner divertor tile 4 and the outer divertor tile 1B. 13 C peaks are observed on tiles 1B (V=0.27, 0.31 and 0.35 m), 6B (V=5.55 m) and 5 (V=5.73 m). The peaks on the tile 1B are shifted toward the roof baffle with respect to the outer strike point. Unfortunately samples from the inner strike point tile 4 were not available. Huge peak of 13 C was observed on the outer divertor tile 3B (V=0.83 m) located horizontally at the chamber. Assuming toroidal symmetry in deposition the amounts of 13 C deposited onto the lower inner and outer divertor was found to correspond to total amounts of and atoms, respectively, which together make 8.9% of the total puffed amount. On the heat shield and upper divertor tiles, and atoms of 13 C were measures, respectively. These numbers represent 4.0 and 5.7% of the total puffed amount. Detailed distribution of 13 C as a function of V-coordinate over the whole torus is shown in Figure 8. A peak was observed at the top outer divertor at V=3.42 m. At the heat shield 13 C deposition increases from the top towards the middle part and then decreases towards the bottom of the chamber. To investigate local deposition of 13 C samples from the limiter tiles located near the puffing valve were measured. Because samples were available only from the limiter tile close to the valve and from the top tile, assumption of linear decrease of the 13 C deposition along the limiter 86

97 6 10 1B 1A 2 3 3B 6A 6B C strike point C (x10 at/cm ) outer divertor inner divertor 26 Discharge time (s) V coordinate (m) Figure C distribution at the lower divertor (dashed line) together with strike point position from magnetic reconstruction during last 13 shots (solid line). 6 low out.div. upper div. HS low in.div. 13 C 5 4 C (x10 at/cm ) V coordinate (m) Figure C distribution over the whole chamber. 87

98 Table 1. Amount of deposited 13 C (per cent of the puffed amount) found in the experiment. Vessel part Lower inner divertor Lower outer divertor Upper divertor Heat shield Limiter Total L-mode, bottom null 2.4 % 6.5 % 4.0 % 5.7 % 21.8 % ~42.9 % surface from the middle to the top and to the bottom was made. Integrating over the limiter surface, it was obtained atoms, i.e. 21.8% of the puffed amount of 13 C was deposited onto the limiter tiles in the vicinity of the puffing location. Summarized data are presented in Table 1. To obtain the total amount of 13 C deposited over the torus some assumptions were made. Unfortunately neither inner lower divertor tile 4 nor dome baffle tiles 9A, 9B and 9C exposed during 2005 campaign, were available for the measurements. However, data obtained on these tiles after 2003 campaign can be scaled using ratio of 13 C amount deposited onto lower divertor tiles available after both 2003 and 2005 year experiments. Results measured on the upper divertor tiles after 2005 campaign were extrapolated to get estimation for the 13 C deposition on PSL tile. Combining data presented in the table 1 and taking into account assumptions above, result in 42.9% of the total puffed amount of 13 C that was detected Simulations of trace methane puffing experiments at AUG In order to study carbon migration and deposition in a tokamak, the above mentioned 13 CH 4 puffing experiments performed in ASDEX Upgrade (AUG) at the end of experimental campaign 2004/ 2005 were simulated using the Monte-Carlo based impurity transport code DIVIMP. The simulated wall deposition profiles were compared to experimental results obtained by post mortem SIMS analysis, presented in Figure 8. While a qualitative agreement was found between the measured and the simulated profiles, Figures 8 and 9, respectively, a quantitative agreement between the modelled results and the experimental observations could not be achieved. Taking the inner and outer lower divertor regions separately from the rest of the vessel wall, there is a reasonable agreement between the model and the experimental results. In Figure 7, at around 0.6 Outer div. Limiter Upper div. Heat shield Inner div. Deposition (arbitrary units) V (m) Figure 9. Simulated 13 C deposition profile assuming 13 CH 4 dissociates into hydrogen and carbon mostly at 4.2 cm away from the wall from which it was puffed in to the device. 88

99 the outer strike point, V = m, one can observe a compound peak structure, and at around V = 0.8 m the experimental results show the greatest level of deposition. In the simulation results, see Figure 9, one can also observe an increased amount of deposition near the outer strike point. Furthermore, at around V = 0.8 m the simulation results, as well as the experimental ones, show the greatest level of deposition, albeit the profile obtained with the simulations is significantly wider than in the measured deposition profile. Also on the inner divertor the experimentally observed peaking in the deposition profile is reproduced. The most prominent peak in the experimental deposition profile, located on the tile 3B (see Figure 7) cannot be explained with the present simulations. The problem stems from the deficiencies in the calculation grid, namely, its distance from the wall. At the puffing location, i.e., the outer midplane, the distance of the calculation grid from the wall is 4.2 cm. That is way too far considering the ionization and dissociation of 13 CH 4. In reality, the puffed methane ionizes and, consequently, starts moving in the direction of the magnetic field lines before reaching the region covered by the grid. As a result, the carbon ends up on field lines intersecting the horizontal part of the vessel wall, exactly where the largest deposition was found in the experiments. Due to the insufficient extent of the grid, the corresponding peak does not exist in the results of the simulations. The highest deposition peak of simulation results (Figure 9), located at the tiles 1B 3, on the other hand, is also an artifact resulting from the lacking grid. If the grid would extend to the wall at the puffing location, majority of the carbon would never get to the flux surfaces intersecting the targets at the location of the peak and would, therefore, be deposited elsewhere. Figure 10. Toroidal (a) and poloidal (b) distribution of the injected carbon at AUG divertor. The vertical axis has been scaled with the observed total deposition. 89

100 In a similar experiment in 2003, 13 CH 4 was puffed at the outer midplane as well as a valve at the outer divertor. The local deposition at the divertor was measured by nuclear reaction analysis (NRA) by R. Pugno et al., (2005). In 2005 the 3D divertor version of ERO was first applied to this experiment, which provides an exceptionally detailed 2D deposition pattern for benchmarking the code. We found that the simulated toroidal decay length of carbon agrees with the measured value, but the cross-field width and shape of the measured distribution can not be reproduced at present (Figure 10). In particular, the observed deviation of the deposition tail from the magnetic field direction does not show up in the simulations. This effect is suspected to be due to the ExB-drift, which only appears in the simulations if a nonzero electric field is imposed in a layer much thicker than the sheath. The simulations will be refined by improving the background plasma in the first phase ERO simulations at JET In 2006, ERO has been applied to two separate problems at JET: the 2004 tracer methane injection experiment and the interaction of ELM filaments with the outboard limiters. In the tracer injection experiment, the dissociation, ionization, local migration, deposition and re-erosion in the vicinity of the injector location can be best modeled with ERO. ERO results can help explaining the observed deposition and re-erosion measured by SIMS and RBS. They are also used to provide information about local migration paths for EDGE2D modeling being performed in parallel in a global scale. The injection geometry is shown in Figure 11. Modelling ELM filaments represents a novel application for ERO due to the rapid evolution of the background plasma. The filamentary plasma structures which constitute an ELM burst can reach the outboard limiters with an appreciable density. The evolution of ELM filament energy and particle densities has been successfully described by simplified kinetic and fluid models by W. Fundamenski and R. Pitts (2006). The fluid model reproduces ELM filament densities and electron energies measured at the outer poloidal limiter on JET and ELM filament ion energies in the JET far SOL, providing thus a reasonable basis for ERO simulations of ELM-limiter interaction. A first attempt to quantify the resulting erosion and material migration from this interaction has been done using the ERO code. Plasma filaments have been implemented in the ERO code as the background plasma for test particles. The filaments have been described by a simple model: Figure 11. Two of the 48 injector locations at the outer divertor of JET. 90

101 uniform along field lines and broadening Gaussian transversal distributions of density and temperatures. The time scale of filament propagation from the separatrix to the limiter is 10 4 s, and it was found that eroded particles migrate over a roughly similar time. However, the background plasma in ERO is time-independent. Therefore, some revision of the code will be needed before actual simulations are possible MD simulation of plasma-wall interaction In ITER the divertor will consist of both C and W. The C parts will be sputtered (eroded) by the H ions coming from the plasma and bombarding the wall. While the sputtering of tungsten by H ions is negligible, there exists another problematic irradiation-induced effect, which needs to be understood. Under intense hydrogen irradiation gas bubbles are formed under the surface. Depending on the implantation energy and dose, this zone can extend to a depth of several microns. When the bubble pressure exceeds a certain limit, the surface is fractured. Large pieces of material are then ejected to the boundary plasma. We have examined the possible blistering of W under H and He bombardment. The results provided an explanation for the vastly different depths of hydrogen and helium bubbles found in experiments on tungsten after non-damaging ion irradiation. The fundamental reason is the different self-trapping behavior of the gas ions in the target. Our density functional theory calculations and molecular dynamics simulations showed that two hydrogen atoms trap each other weakly with a binding energy of less than 0.3 ev, whereas helium atoms form strong pairs with a binding energy of about 1 ev. Due to this, larger helium bubbles can form spontaneously close to the projected range, whereas hydrogen atoms migrate deep down in the tungsten target before becoming trapped by defects, as verified by the present results of kinetic Monte Carlo simulations. These findings are also relevant to blistering, as they help to explain why the lid thicknesses of hydrogen blisters are so much larger than the projected range of the ions. Since the He bubbles are formed close to the projected range, it is possible to simulate the entire bubble formation and rupture process with molecular dynamics for the lowest bombardment energies. An illustration of the rupture of a He nanosized bubble is shown in Figure 12. Analysis of about ten such rupture events showed that they do not lead to any W erosion, which indicates that at least the close-to-surface bubble ruptures may not be a problem for reactor operation. Figure 12. Rupture of a He bubble in W. The blue spheres represent He and the red spheres W atoms. Since ITER will contain both C and W parts, the transport of hydrocarbons in the reactor will lead to the formation of WC thin films in the surface layers of the divertor. Hence it is important to understand how these behave during bombardment by H atoms: do they erode like C, or form blisters like W? To be able to examine this, we first developed a reactive interatomic potential for the ternary WCH system, capable of modeling both bulk C and W, small hydrocarbon and tungsten hydride molecules as well as H dissolved in bulk W. We have then used this potential to examine the erosion of WC materials by H bombardment. We found that prolonged bombardment of W by D leads to the for- 91

102 3.6.8 Flake formation of re-deposited amorphous carbon films Figure 13. The rate of W buildup in WC during prolonged bombardment by D. Since the C is eroded preferentially, the W content of the cell increases with increasing D fluence. Our simulations carried out in initially crystalline WC cells show excellent agreement with the experimentally measured buildup. mation of an amorphous WC surface layer, regardless of the initial structure of the WC sample. Hydrocarbons formed at the surface of these amorphous layers can erode by swift chemical sputtering. Larger sputtering yields are obtained from carbon-terminated surfaces than from tungsten-terminated surfaces. In both cumulative as well as non-cumulative simulations C is observed to sputter preferentially. The results obtained have important implications for mixed material erosion in fusion reactors. Just like C-based materials, WC-like materials can be expected to be subject to chemical erosion down to very low energies of impinging deuterium or tritium particles. This is in contrast to pure W, which does not erode due to chemical sputtering. In agreement with experiment, we observe preferential sputtering of carbon (see Figure 13). This suggests that WC layers formed by C re-deposition will be reduced in C content if subjected to hydrogen/deuterium bombardment. Hence, if a section of the reactor first wall is subject to both re-deposition of hydrocarbons and hydrogen bombardment, a dynamic balance in the C content could be reached under prolonged operation. Carbon fibre composites (CFC), tungsten (W) and molybdenum (Mo) are candidates for the divertor material in the next step fusion device ITER. Carbon based materials are used in areas with high heat load and particle bombardment. The chemical erosion of C due to hydrogen is high. Eroded C particles can redeposit on other regions of the wall surface and form rigid diamond-like carbon (DLC) coatings. These coatings hold high internal stresses, which can lead to flaking of the film. Carbon flakes can contain large amounts of radioactive hydrogen isotope, i.e. tritium. The level of tritium containing flakes has to be low in order to minimise the total amount of hazardous tritium in the fusion device. Several sets of C thin film samples with different deposition parameters were prepared to study the adhesion and flaking mechanism. Samples were manufactured by Diarc Technologies Inc. using the Diarc Method. Carbon film thicknesses were 250 and 500 nm on both W and Mo substrates and in addition 750, 1000 and 1250 nm films on Mo were prepared. Films were grown at three different temperatures T dep (room temperature, 100 and 300 C) and methane partial pressures P CH4 (none, 10-4 and 10-3 mbar). Secondary ion mass spectrometry (SIMS) and Raman spectroscopy were used to determine hydrogen concentrations C H and the sp 3 /sp 2 bonding ratios in the films, respectively. C H was found to be constant throughout the films (with P CH4 = 10-4 mbar C H was at.% and with P CH4 = 10-3 mbar approx. 6 at.%). Films grown at RT in vacuum or in P CH4 =10-4 mbar had sp 3 s between %. Films deposited at 300 C held sp 3 s from 0 to 35 % increasing with the C H. Scanning electron microscopy (SEM) and 3-D stylus profilometer were used in analysing topographies of the samples and dimensions of the stress relief patterns therein. In general, C films had good adhesion, but delamination occurred in some samples. Various 92

103 Figure 14. Sinusoidal and straight-sided stress relief patterns on a 250 nm C film on Mo seen with SEM (left). Deposition temperature was here RT and CH 4 partial pressure 10-4 mbar. Increase in the partial pressure (10-3 mbar) generated flaking to the film (right). Figure 15. SEM picture of sinusoidal stress relief patterns starting at some dislocation in the film with thickness 250 nm on Mo substrate (left). 3D profilometer picture of sinusoidal wrinkle (right). stress relief patterns were seen, e.g. straight-sided and sinusoidal patterns see Figure 14. Increasing film thickness or T dep decreased the film adhesion to the substrate. Flaking was strongly dependent on C H.AtlowT dep the film adhesion decreased with increasing C H. At 300 C increasing the C H decreased the film flaking. This dynamic effect of codeposited H with deposition temperature stands for a competing behaviour of the residual stress evolution in the film Deuterium irradiation induced defect concentrations in tungsten In fusion devices, hydrogen escaping the plasma can cause hydrogen build-up in the subsurface region of the first wall materials. Captured hydrogen can be partly released back to the plasma, and at the same time retention in the wall materials also takes place. Retention is affected by several 93

104 parameters, e.g. incoming hydrogen energy, net damage in the material, presence of trapping impurities. The hydrogen recycling will influence the particle balance in the plasma. Tungsten (W) is a valid candidate as plasma facing material in fusion devices used in extreme conditions in divertor and baffle areas. In the course of ion implantation, different types of hydrogen trapping defects are formed, namely host material interstitials, vacancies and clusters of them. In addition to these lattice damages, the host material holds other trapping defects that exist a priori implantation, i.e. grain boundaries and impurity atoms working as an effective ion trapping sites. Formation of implantation damage is a dynamical process where highly mobile interstitials and almost immobile vacancies interact with each other e.g. recombining, forming clusters, diffusing to the sample surface or deeper into the bulk. Some part of the implanted ions gets backscattered from the sample surface and the rest penetrates into the sample. In the case of hydrogen implantation into W, the free hydrogen can be trapped at defects, out-diffuse from bulk to the surface or diffuse deeper into the bulk far beyond the implantation zone because of its low diffusion activation energy 0.39 ev. Implantation of D into W was carried out at room temperature (RT) using D + 2 ions 5, 15 and 30 kev per deuteron, respectively. The implantation dose was5.8x10 16 D/cm 2. Annealing was carried out x x Temperature (K) QMS (molecules s cm ) st 1 trap nd 2 trap rd 3 trap th 4 trap Annealing time (s) Figure 16. D molecule thermodesorption spectra and annealing temperatures as a function of annealing time for removing D atoms from four different traps in W. D release starts at sample temperatures 455, 560, 663 and 801 K from traps 1 to 4, respectively. 94

105 x As-implanted Ann., max. temp. 528 K Ann., max. temp. 603 K Ann., max. temp. 726 K D concentration (am ) st 1 trap nd 2 trap rd 3 trap th 4 trap Depth (nm) Figure kev as-implanted and annealed D depth profiles measured by secondary ion mass spectrometry and normalized by nuclear reaction analysis. The different striped regions show profiles of each trap type. in a quartz-tube furnace equipped with a quadrupole mass-spectrometer. The temperature was measured with a calibrated thermocouple in direct contact with W surface. Figure 16 shows the initial part of the annealing temperature and QMS signal as a function of annealing time for 30 kev D implantation. Due to the high mobility of D atoms, only trapped D is left in the sample after RT implantation. Each defect type has different D binding energy, which is defined as the energy difference for a D atom in a trap compared to a solute or mobile atom in the crystal. Figure 17 presents the 30 kev depth profiles of D atoms for each sample measured by secondary ion mass spectrometry and normalized by nuclear reaction analysis. The amount of retained D after implantation is rapidly decreasing with implantation energy decrease. This can be evidently seen in Table 2 and Figure 19, where the retained amount of D compared to implantation fluence decreases from 55 to 22 and further to 7 % with decreasing implantation energy 30, 15 and 5 kev, respectively. A part of this decrease is due to increased D backscattering with decreasing implantation energy: 10, 17 and 29 % backscattered D with 30, 15 and 5 kev, respectively. However, the main reason to the rapid decrease in the retained D is the increase of the recombination rate of implantation induced self-interstitials and vacancies with decreasing implantation energy, resulting in a decrease in the number of defects to which D can be trapped. 95

106 Retained D / D/cm kev/d 15 kev/d 5 kev/d SRIM-03 as-impl. 520 K 610 K 725 K 875 K Figure 18. Retained D concentration relative to implantation dose 5.8 x 1016 D cm -2 presented as function of annealing temperature. SRIM-03 results (incl. backscattering) are shown as a reference to the measured as-implanted D concentration. Annealing of the W samples decreases retained D concentration gradually via different recovery mechanisms. Table 2. Retained D concentration after 30, 15 and 5 kev implantation with dose of 5.8 x 1016 D cm -2. In the table are presented the SRIM-03 simulated results which include backscattering and the SIMS experimental results from as-implanted and annealed W samples (in units of D cm -2 ). 30 kev 15 kev 5 kev SRIM x x x10 16 as-implanted 3.2x x x K 1.7x x x K 8.0x x x K 3.3x x x K 3.2x x x

107 4 Remote Handling Systems 4.1 Divertor Test Platform for ITER VTT System Engineering Jorma Järvenpää and Arto Timperi (DTP2 Manager) Introduction The fusion reaction produces high energy neutrons which are absorbed by components lining the inside of the reactor vessel leaving them beta and gamma activated to a level of several hundreds of Grays per hour. A major issue for the successful operation of ITER therefore is the maintenance and/or exchange of in-vessel components by remote handling methods. The ITER machine contains thousands of in-vessel components, but in order to limit the scope and complexity of remote handling activities within the Tokamak itself, the ITER maintenance concept relies on the removal / installation of relatively large modular systems whose maintenance at sub-component level is achieved within a purpose built hot cell. The lower part of the plasma chamber is occupied by the divertor. Due to erosion of the plasma-facing surfaces and the possible need for improving the basic divertor design, its periodic replacement and refurbishment in the hot cell is foreseen a number of times during ITER s 20 year operational lifetime. To meet the requirement for modular replacement, the divertor comprises 54 cassettes independently mounted on toroidal rails fixed to the inside of the reactor vessel (see Figure 1). Each cassette weighs approximately 10 tonnes and their external dimensions are approximately 3.5m long x 2.1m high x 0.8m wide. Cassettes are removed from the vessel using remote handling devices known as cassette movers. One such mover is the Cassette Multi-functional Mover (or CMM) which when fitted with a number of different end-effectors is used to carry out a number of maintenance tasks in the vicinity of the vessel access ports. One such end-effector is the Second Cassette End-Effector (or SCEE) and is used specifically to remove the cassettes immediately to the left of the vessel access port. Having developed the outline engineering design for the CMM and its SCEE, the European Partici- Figure 1. Section through the divertor region. 97

108 pant Team for ITER constructed prototypes. These prototype devices will subsequently be used for the detailed development of ITER divertor remote handling procedures within a purpose built test facility known as the Divertor Test Platform 2, or DTP2 (see Figure 2), to be constructed in the VTT Research Centre in Tampere, Finland. The decision to establish the DTP 2 laboratory in Tampere Finland was made in the autumn The DTP2 laboratory is a joint venture activity between EFDA, Euratom Tekes, Tampere University of Technology and VTT. The main element of the DTP2 facility is the Divertor Region Mock-up (DRM), a large fabricated structure which replicates the geometry of the lower part of the ITER vessel in the region of one radial port. The DRM provides the necessary radial and toroidal rails to support the divertor cassette and the Remote Handling Equipment used to carry out divertor maintenance Design The preliminary design work for the DRM test facility was performed in Italy by ENEA. Finalizing the design work and updating the changes of the ITER model for DRM was started in the beginning of The concept design for the European tendering process was finalized during the spring For the delivery of the basic structures of DRM the international tendering process was organized in the summer The winner of the competition was Tampere based company TP-Konepajat Oy. After selection of the supplier the co-operation to fine tune the constructions started in a good co-operation with the supplier. The design work of the steel constructions was started at TP-Konepajat on December All the manufacturing drawings and instructions were finalized by the summer Figure 2. ITER DRM model. 98

109 4.1.3 Implementation design During 2005, it was realised the assembly of DRM requires the major reinforcement changes for the floor of the DTP2 laboratory. This was due to the heavy steel constructions and the very tight position tolerances of the DRM structures. The work for the floor was started by opening the floor in the spring 2006 (figure 3.). The reinforcement project became ready before the delivery of the first steel structure in the autumn The manufacturing of the DRM parts started in the spring 2006 and it included many large welded constructions that required quite a lot of work. The assembly of DRM structure was started in October 2006 and it was delivered and had its acceptance inspection in December As a conclusion, it can be mentioned that the co-operation between the partners in this phase of the large project worked out very well and in the scheduled time frames (Figure 4) Safety analysis The work performed in this kind of remote handling and remote controlled system requires extra caution of the safety. The analysis and design of the work environment and DTP2 laboratory is included in the project portfolio. This task was performed by utilizing the safety analysis method developed by VTT. This method has been developed and used for the analysis of environment and safety for the conditions where the remote operated mobile work machines are functioning. Typical applications are mines and automated factories. This method is applicable for the other ITER-tasks. Figure 3. Floor reinforcing. Figure 4. Final assembly of DTP2. 99

110 In this development environment, there will be prototype service robots under the wired control of complicated software. The reliability and safety of the control software is an essential requirement. The development of this kind of software requires high quality and reliability of the structure and detailed coding work. The performance environment sets it own demands for this development too. In this work, the method for the verification of the software was developed for the ITER environment. The method was used for CMM/SCEE control software Future The test environment will be finalized during the first quarter of The required computers and cabling will be delivered from Spain and assembled on the site during the first part of the year The Cassette mock-up will be delivered from Luxemburg to Tampere in January CMM/SCEE will be delivered to Tampere from Spain. The first test runs of the DTP2 facility are scheduled to be performed in the autumn After this the actual development work with this DRM tool for ITER will be started PREFIT In relation to DTP2 laboratory a new education programme was started for young engineers in December The programme is called Preparation of Remote Handling Engineers for ITER (PREFIT). The purpose of the programme is to train and educate young professionals for the future fusion and ITER research needs. The funding of the programme is an extension of EU s Marie Curie Programme. The participants of the programme are Oxford Technologies Ltd UK, CEA France, Tampere University of Technology and VTT Finland. The each country has nominated two trainees in the programme and they have all started working for the project. The kick-off meeting was held on December 14th in Tampere. The programme runs for 3 years and after that the trainees should receive the PhD degree specified for Fusion engineering at Tampere University of Technology. Two Finnish trainees are participating to this programme (Figure 5.). Figure 5. PREFIT kick-off meeting participants 14 th December. 100

111 4.2 Development of Water Hydraulic Tools and Manipulators for ITER Divertor Maintenance Tampere University of Technology Institute of Hydraulics and Automation Mattila Jouni, Siuko Mikko (project managers) Aho L, Esqué S, Koivisto H, Kopperoinen A, Kunttu P, Linna O, Luomaranta M, Muhammad A, Mäkelä A, Mäkinen H, Nieminen P, Nykänen T, Pitkäaho M, Poutanen J, Raneda A, Saarinen H, Sainio A, Takalo V, Tammisto J, Toivo M, Tolonen M, Verho S, Vilenius M, Virvalo T. TUT/IHA has worked for ITER project since year 1994 in the field of ITER divertor maintenance. During these years IHA has participated in various projects, including modelling, designing, manufacturing and testing of different prototypes and components. IHA s special expertise areas are water hydraulic robotic devices and virtual modeling / prototyping. Due to the erosion of the plasma facing components and possible need for improving the design of critical items, replacement of the divertor and its refurbishment in the hot-cell is foreseen 4 times during the ITER lifetime. The divertor consists of 54 cassettes (each weighting approximately 8-9 tons), which can be replaced through three RH ports. The cassette multifunctional mover (CMM) with a set of end effectors (e.g. Second Cassette End Effector SCEE) and the cassette toroidal mover (CTM) are needed to transport these cassettes from the reactor to the transportation cask. During the maintenance 6 DOF water hydraulic manipulators will be used on the top CMM and CTM for assisting work such as bolting/unbolting, pipe cutting and welding etc. Divertor maintenance tasks are expected mostly to be teleoperated and often with force feedback (human-in-the-loop) due to complex tasks and limited viewing system. High requirements for accuracy, reach and controllability, as well as typically very high payloads, are the main reasons for choosing hydraulics as the operating principle for the maintenance equipment due to its compactness (high power-to-size ratio) and reliability. However, instead of oil, hydraulic systems will use water as a pressure fluid to eliminate the risk of contaminating the reactor elements. Water is also not activated or effected by the radiation of the environment. During years 2003 to 2006 IHA has participated in several projects, mostly on CMM system design, documentation and virtual modelling, water hydraulic manipulator (WHMAN) development, and water hydraulic components development Development of water hydraulic manipulators Introduction: Manipulators will be used in ITER divertor maintenance tasks for various supportive work. IHA work has concentrated so far mainly on a manipulator arm locating on top of CMM, whose operation relates to divertor cassette refurbishment cycle, where the cassette is picked up from the reactor, and carried to the hot cell. The aim of this work is to start from defining the manipulator tasks and gradually proceed towards creating a prototype manipulator for DTP2-platform (Divertor Test Platform 2), where ITER divertor replacement operations are studied, developed and demonstrated. The development has progressed is logical steps: 1. Defining manipulator tasks. 2. Creation of preliminary manipulator mechanism design, and testing it in virtual environment. 3. Designing and building a general purpose simplified (3 DOF) teleoperated prototype Water Hydraulic Manipulator and study teleoperation with it. 4. Proceeding towards a more complex manipulator design by adding degrees of freedom (6 DOF) to the prototype. 5. Continuing on the path towards creating a prototype manipulator suitable for ITER relevant operational scenarios and environment. 6. Studying and developing control methods, remote handling and force feedback operation. 101

112 In parallel with the manipulator design, a development of robust designs for water hydraulic rotary joints was carried out to support the overall manipulator development. Water hydraulic vane actuator has superior performance when light and compact in volume but still high torque actuator is needed to drive a robotic joint. The downside of the vane actuator is difficulty in finding an optimal balance between actuator seal leakage and seal friction. In order to proceed towards prototype manipulator, a new vane actuator design was needed. Manipulator task defining and preliminary design (2003): Manipulator tasks were defined based on information from EFDA, and a preliminary manipulator mechanism design was created, consisting of 7 joints. A 3D IGRIP virtual model of the manipulator and maintenance tunnel environment was created for verifying that the tasks can be performed with the current design (see Figure 1). By simulating the manipulator arm in its actual working environment the joint configuration, the dimensions and the actuator motion ranges and torque requirements were obtained (see Figure 2). Simulations also assisted in understanding the required operations better and allowed some simplifications of the tasks. Figure 1. MAM unlocking the diagnostic rack. Figure 2. Preliminary manipulator design. 102

113 Figure 3. Teleoperation experimenting with Phantom master arm and 3DOF WHMAN. 3DOF manipulator prototype design (2004): A prototype of 3DOF manipulator was designed and manufactured in order to start teleoperation experimentation with Phantom desktop 6DOF force/torque feedback master arm (see in Figure 3). In mechanical design the structural stiffness of all components and the kinematics of the manipulator were optimized for force control purposes. The first two joints were actuated by cylinders and the third joint by a vane actuator. The load carrying capacity was defined to be 200 kilos where 100 kilos were reserved for future robot 3DOF wrist weight and other 100 kilos for payload. In addition, combined position and force/torque controller was designed to control the manipulator. Measurements were carried out with a purpose built single joint test bench for preliminary tests to provide experience on the designed force controlled water hydraulic manipulator. 5DOF manipulator prototype (2005): The target was to extend existing water 3DOF hydraulic arm design with 3 DOF wrist mechanisms, shown in Figure 5. Designed 3DOF wrist was installed to existing planar arm and master-slave teleoperation methods were developed with water hydraulic manipulator using Phantom desktop 6DOF force/torque feedback master arm. For wrist assembly actuators, improved water hydraulic vane actuator design, manufacturing and testing were carried out. Also different fluid power transmission line (pipes, hoses and borings) designs were studied for wrist mechanism. Virtual prototyping was used extensively for improving wrist designs and for controller development prior to actual real prototype building. WHMAN wrist was build with novel IHA developed robust and cost-effective water hydraulic vane actuator design concept (see Figure 4). Based on design specifications verified by new vane actuator tests, 3 more vane actuator units where manufactured for 3DOF WHMAN wrist. The wrist consists of three water hydraulic vane actuators connected in series. Fluid power transmissions lines were fitted inside of manipulator structural linkages. After last joint 6DOF force/torque sensor was installed for research and controller development purposes. 103

114 Figure 4. New design actuator chamber and vane shaft FEM analysis. Figure 5. 3DOF wrist design installed to existing planar arm. Test and teleoperation development program for WHMAN prototype which consisted of earlier developed WHMAN arm and newly developed 3DOF wrist was carried out. Figure 6. 6-DOF WHMAN design. On the basis of experience acquired from 5DOF WHMAN, a 6DOF WHMAN is under development at IHA/TUT. The arm of the manipulator is composed of three rotational and one linear joint. The rotational joints provide a spherical workspace while linear joint provides the telescopic motion of the arm. A spherical wrist is attached at the end of the arm resulting in improved dexterity of the arm. Manufacturing of the arm is estimated to be completed by spring A 3D model of the arm is shown in Figure 6. WHMAN control and teleoperation development Starting from the first 3 DOF prototype, control and teleoperation experiments have been carried out parallel with the mechanical and hydraulic development. Teleoperation experimentation has benefit greatly from early development and testing that was carried out from the very beginning of the project with a real-time dynamic virtual prototype of WHMAN including 3D visualization. WHMAN virtual prototype is updated with the new and more accurate design data as the WHMAN design is improved. 104

115 Master-slave configuration control software is developed for new multi-joint water hydraulic manipulator. Teleoperation master-slave control methods are studied and tested with multi-joint water hydraulic manipulator with Phantom desktop 6 DOF force/torque feedback master arm. Existing Graphical User Interface (GUI) is modified and improved for new multi-joint manipulator teleoperation tasks. Test environment for testing and training typical teleoperation tasks (e.g. insertion of a peg in a hole or bolting) is designed and manufactured. Virtual modeling of the test environment (ENVISION) is carried out and the integration of virtual world visual feedback into teleoperation GUI is started. The Ethernet has been chosen as the communication medium because of its wide availability and long familiarity in the industrial world. The time critical control data and nondeterministic information is exchanged between the system components using UDP (User Datagram Protocol). Software tools have been developed at application layer to tackle the unreliable and nondeterministic behaviour of UDP. The general architecture of the teleoperation system has been shown in Figure 7. A real time operating system running on a computer performs the low level control and measurement operations for WHMAN. The WHMAN is employed as slave manipulator to carry out the tasks at remote environment. Several cameras are located at the remote site to provide the live video feedback of the environment and task to the operator. The operator site is equipped with digital computers, video monitors, a haptic device, a joystick, light indicators and an emergency stop switch. The operator control stations provide Figure 7. Teleoperation system architecture. 105

116 many levels of control and several aids to control a remote handling task. The operator of the WHMAN host controls and monitors the online status of WHMAN and has the highest authority. In case of emergency the operator can either shut down the slave manipulator with emergency switch or can take over the control of manipulator. One or two operators can use the GUI (Graphical User Interface) and Phantom Premium 3.0/6DOF master manipulator. Operator/s can control the slave manipulator from this station in several possible modes provided the permission is granted by the WHMAN host operator. In addition to live video feedback, the operators have the visual access to the virtual 3D model of the task and environment on IHA3D station. This model is not only helpful during the task execution but also for the task planning, practicing and simulating Development of divertor maintenance equipment (CMM) Introduction: For cassette refurbishment, the divertor cassette is picked up from the reactor, and transported to service area, called Hot Cell. In co-operation with EFDA, IHA defined the cassette replacement procedure. The whole divertor replacement cycle involving all CMM tasks was studied, and based on these studies the second cassette handling was determined as the most demanding task. Therefore IHA has designed the CMM and the second cassette end effector (SCEE) for the second cassette replacement. The aim of this work is to start from defining divertor cassette replacing task and to gradually proceed towards testing cassette replacement scenario with real applications in DTP2 facility. The development has progressed in following order: 1. Studying the CMM tasks, and specify the design requirements. 2. Designing CMM basic mechanical structure and hydraulic system, and selecting preliminary electrical components. 3. Providing design changes for the CMM and the SCEE meeting the latest design of the ITER reactor design. Updating the documents and drawings and re-analyzing mover strength. 4. Updating 3D virtual models of the CMM, SCEE and the reactor environment. 5. Providing a reference design plus technical assistance and expertise to EFDA for call-for tenders of the CMM mock up. 6. Testing the general applicability of an commercial motion controller to a large water-hydraulics actuator, and confirming that such a system can achieve the positional accuracy and speed required for the CMM under a number of realistic load conditions. 7. Producing the low-level software required for driving the CMM hydraulic axes well before the CMM system is available. 8. Developing the high level software, which, interfacing with the low level software, will enable the real CMM/SCEE prototype to carry out its required overall functions. 9. Finally, when the CMM prototype is delivered (during 2007), start testing the cassette replacement scenario with full-size prototype. Divertor maintenance equipment CMM designing: The design started based on the conceptual design made by the Framatom Siemens. Based on the design simulation model including flexibility of the structure and the hydraulics was made in co-operation with LUT. These simulations were made by integrating the mechanical structure models made with Adams and hydraulic models made with Matlab-Simulink simulation programs. The work cycles of the CMM were studied with IGRIP simulations (an example shown in Figure 8) in order to define optimal trajectories and motion ranges against the service tunnel space limitations and tight clearances. These simulations proved to be valuable tool finding out potential difficulties in the cycle as well as for presenting the work cycles in several different occasions. As a result of these simulations some design changes were seen necessary in the reactor and cask design. The sizing and selection of the CMM and the SCEE components was done based on calculations and simulations. Hydraulics and electric components were selected taking into account 106

117 Figure 8. CMM in RH port reaching for the second cassette. the defined environmental and task requirements, and assistance was provided to EFDA in writing technical specifications for the mover call-fortender documentation. Parallel to the design process radiation testing of the hydraulic components were carried out in EFDA RADTOL program. The initial CMM/SCEE design in presented in Figure 9. After the initial design was finished, IHA started evaluation of the effects of recent ITER-reactor design modifications. The reactor radial port and the Cassette were brought up to date to the CMM and the SCEE design. Also design modifications were made to consider the renewed cassetteto-rails locking system. The most changes were made for the SCEE hook-plate and the cantilever arm design, but also side wheel position and lift arm structure was modified. Also the Cask, carrying The Mover and the Cassette to the hot cell, was modified and the mover design was checked against the changes. The Mover strength and the Cask rails strength were re-analyzed. Figure 10 presents the updated CMM design and reactor geometry in virtual environment. Figure 9. CMM+SCEE structure. 107

118 Figure 10. Updated Mover operation in virtual environment (IGRIP). Test models and software for the CMM system: In order to simulate Mover (CMM and SCEE) mechanical properties in a test bench, the Mover mechanism natural frequencies were calculated based on 3D CAD model data. The estimated natural frequencies were used as a design input for the test bench. A Single Axis Model (SAM) representing typical CMM/SCEE water hydraulic axis was designed and built to be used as a general testing system (see Figure 11). SAM is able to emulate CMM/SCEE joints with two different cylinders and four different sets of weights. SAM can be balanced with weighting so that it has purely inertia load only like is the case with CMM/SCEE last two joints (hook-plate and cantilever arm). CMM/SCEE lift and tilt joints, however, are loaded heavily with gravity loads, but SAM is able to emulate them as well. SAM was used to prove the general applicability of a Moog Servo Controller (MSC) M3000, including also testing programming and communication methods. To support development of Higher Level Controller (HLC) and user interface (see Figure 12), a hardware-in-the-loop system was created consisting of a virtual model of the CMM/SCEE (the Multi-axis virtual model, MVM), MSC and SAM. MVM is most useful for CMM/SCEE High Level Control (HLC) system development including its graphical user interface design and testing. HLC development was initialized and key functions like forward and in- Figure 11. Single Axis Model (SAM) test bench design. 108

119 Figure 12. Graphical user interface of the CMM High Level Controller. verse kinematics and trajectory generators were build and implemented for chosen real-time operating system. SAM and MVM were tested with the first prototype of IHA CMM/SCEE HLC. Radiation tolerance testing of water hydraulic cylinder seal: IHA participated in RADTOL testing project, in order to obtain important information about hydraulic cylinder seal radiation tolerance characteristics. The seal samples to be tested were selected (by IHA, CIEMAT and EFDA). The samples were tested for leakage and friction (by IHA) in real hydraulic system prior to irradiation. After that, the seals were irradiated (by CIEMAT) with different doses. After irradiation, the leakage and friction characteristics were measured again. Based on the leakage test results, the seal properties seem to remain at satisfactory level up to 1 MGy cumulative dose Conclusions The development of a teleoperated force feedback 6DOF Water Hydraulic Manipulator arm prototype has been progressing steadily, including new vane actuator designs and concepts. Virtual prototyping has proved to be a valuable tool in this kind of a demanding design process. Teleoperation master-slave control methods and techniques have been studied and developed with multi-joint water hydraulic manipulator with Phantom desktop 6 DOF force/torque feedback master arm. Cassette Multifunctional Mover design has undergone several iteration cycles, and without the aid of virtual modeling the process would have required significantly more time and resources, and probably several prototype phases. The virtual modeling has also enabled the development of high- and low-level-controllers and software well in advance before the actual prototype is built and operational. In parallel with the main ITER related work, water hydraulic component development has been carried out, and valuable knowledge of radiation effects on hydraulic seals and components has been gained. 109

120 4.3 Development of a High Precision Intersector Weld/Cut Robot Lappeenranta University of Technology Institute of Mechatronics and Virtual Engineering H. Handroos (Project Manager), H. Wu, P. Pessi, Y. Liu, H. Yousefi, E. Tenkanen and J. Hopia The vacuum vessel of ITER composes of sectors that are very large-components with some complex geometrical features. Plasma physics and internal components assembly require more stringent tolerances than normally expected for the size of the structure involved. Overall, assembly tolerances are expected to be within 10 mm (± 5 mm) for the whole vacuum vessel. During the first assembly of ITER the sectors need to be joined by leak tight welds. Because of components and systems mounted on the outer surface of VV the assembly joining needs to be carried out from the inside of VV. Because there is no space for machine bed inside VV for mounting the welding robot and the work spaces of commercially available industrial robots are not sufficiently large for welding complete seams special robot has been developed in series of projects. The main objective of IWR (Intersector Weld/Cut Robot) development has been to produce a robotic system, which carries out several processes necessary to assemble or remove sectors of the VV. In the previous program a 5-axis parallel robot was developed and built Figure 1. The robot is designed to be mounted on a trolley that moves along the track mounted on the surfaces of VV sectors. In the robot hydraulic rams and parallel kinematic structure are utilized to maximize its stiffness to weight ratio. In the current program the 5-axis IWR-prototytype was equipped by 6-axis seam tracker and laser welding tests were carried out in VTT Laser Welding Laboratory. Also cutting tests with and without lubricant were successfully carried out. Since the track support system of IWR was significantly modified by EFDA, the robot design had to be upgraded. The complete design of the robot trolley was modified to meet the new requirements. The robot kinematics was converted into 6-DOF Stewart type kinematics and three additional motion axes were introduced to meet the new work envelope requirement. In total the new robot version includes 10-degrees of mobility. The robot has now to be able to operate on the both sides of the track. The design has been finalised with a manufacturing company and the manufactory of the mechanical parts are started. The design of water hydraulic system is been finalized. In addition to the actual design and optimisation of the hydromechanical system of IWR an emergency shutdown program for the robot controller to protect the robot and its parts against failures in cases of loss of control caused by electronic problems was developed. Also a vibration damping method based on piezo-electric actuators for IWR to reduce the vibration amplitude while machining was studied and demonstrated by a simplified mock-up. Furthermore, the applicability of laser measurement in measuring piston position of a water-hydraulic cylinder was studied by theoretical analysis and literature survey. Upgraded IWR construction is shown in Figure 2. The water hydraulic 6-DOF robot 1 is mounted Figure 1. 5-axis prototype of IWR. 110

121 on a steel carriage 2 which is driven by two separate servo motors including cyclo gears 4. The carriage moves along the track by employing rack and pinion drive. A water hydraulic bearing force compensation system 3 is maintaining constant contact force between bearing wheels and rails. Since the radius of the track is varying the distance between the upper and lower wheel must vary in order to prevent damage. To be able to operate in the both sides of the track slewing bearing 8 with rotational drive unit 5 is introduced. To enlarge the work envelope of IWR an additional linear drive unit with bearings 6 will be used. Hydraulically driven tilting mechanism of the hexapod frame 9 is also needed to reach the lower areas of the work space. All ten degrees of mobility cannot be active during the operations of the robot. They are mainly used for moving the hexapod robot in appropriate positions to avoid workspace boundaries. The rotation table 8 is equipped by water hydraulic clamping system 11 in order to improve the stiffness while machining. The manufacturing of the mechanical parts of the robot are finished by Imatran Kone Oy. The component selections for the water hydraulic system of the robot is completed. The water hydraulic servo cylinders and valve blocks are currently manufacturer by Hytar Oy. The prototype controller based on MATLAB/Simulink Software and dspace-hardware that was developed for 5-axis IWR was upgraded into industrial PC-controller based on Beckhoff s EtherCAT-system shown in Figure 3. To demonstrate the idea of suppressing machining vibrations in IWR by piezo-hydraulic actuators was demonstrated by a simple test rig. A 30mm long piezo actuator with 300 micron stroke was purchased. A test rig including the piezo actuator, a single hydraulic position servo and a hydraulic loading force servo was built and instrumented. Load compensation tests were carried out by giving sinusoidal force into the position control system. The error caused by the flexibility of the hydraulic position servo was successfully compensated as shown in Figure 4; Experimental vibration suppression test result. Figure axis version of IWR (under construction). 111

122 Figure 3. Industrial EtherCAT PC-controller for 10-DOF IWR. A novel method for measuring the position of a water hydraulic cylinder was proposed. The method based on laser interferometer measurement was developed by using theoretical analysis and literature survey. According to theoretical study the method is applicable but expensive and it requires precise temperature measurement because the measurement principle is extremely sensitive to temperature variations. Also variations in water pressure must be taken into account. The experimental test system based on commercially available components was care- Figure 4. Vibration suppression test for piezo-hydraulic hybrid actuator. 112

123 fully designed. Because of unexpectedly high equipment costs and requirement for extreme temperature measurement accuracy the experimental testing plan was given up. Despite of this, the proposed method in conjunction with fiber optics might be promising solution in ITER environment because of its insensitivity to radiation. A condition monitoring program for IWR was also developed to ensure safe operation. Before fault situations can be detected, all the possible faults, which could occur conceivably during the execution of task, must be defined. Defined fault situations for the IWR robot are: Loss of sensor signal Defective sensor signal Jam of the valve spool Unexpected contact with environment Oversized machining force. After defining all fault situations, the detection methods were developed. To detect defective encoder signal or jammed spool, the speeds of cylinders are approximated from pressures and valve currents using hydraulic flow equations. To detect the overloading situations a Jacobian matrix based estimation of Cartesian forces from hydraulic forces is carried out in the developed program. The fault detection executes command to the robot to go to pre-defined safe-state. The safestate must be chosen such that the robot can freely be driven into a position in which in can be transported into such track location in which it can be taken out from vacuum vessel for repair. Also when entering the safe-state, the robot is not allowed to collide with any obstacle. A monitoring algorithm was designed and demonstrated in MATLAB/Simulink program successfully. Future work: In 2007 the 10-DOF IWR version will be assembled and tested in the laboratory of IMVE, LUT. The welding and machining capabilities of IWR will be carefully analyzed by carrying our real experiments with limited mockups. In addition to the experimental work, manufacturing simulations of IWR with VV sectors are carried out to develop IWR-assisted assembly scenarios for ITER vacuum vessel. 113

124 5 System Studies 5.1 Socio-Economic Studies VTT, Energy Systems Antti Lehtilä and Riitta Korhonen TIMES modelling of global energy systems Introduction: The project has been carried out under the EFDA programme for Socio-economic Research on Fusion (SERF). The objectives of the project were to analyse the potential of key energy technologies in the global energy system under various scenarios for resources, demands, policies and technologies. As fusion is among the major promising new energy supply options for the future, it is important to evaluate its potential during the 21 st century, with respect to the foreseeable global energy system challenges. A key area of analysis is the assessment of long-term climate change mitigation, which can have great implications to the future competitive position of many energy technologies. The project has been carried out in collaboration with five other research groups, all using a common tool in the analyses, the EFDA Global TIMES model. The six groups participating in the work programme were IPP (Garching), Ciemat (Madrid), OEAW (Graz), UKAEA (Culham), ENEA (Rome) and VTT (Espoo). The modelling tools have been developed under the IEA ETSAP Programme, with VTT s active contributions. The EFDA Global TIMES model is a very large and complex multi-regional partial equilibrium model. The basic version of the model consists of 15 world regions, with an additional distinction of the resources and production of primary energy in OPEC and non-opec countries within each region. International trade of crude oil, natural gas and LNG is included. The model database includes characterizations and projections for over 1000 new energy and process technologies in each region. The maintenance and updating of such a large technology database is a demanding task. Within the project, parts of the database have been reviewed and updated at VTT, under the coordination by EFDA. At VTT the model has been also augmented by an additional region describing the Finnish energy system for global analyses with a national side-perspective. Key base assumptions used in the model: When analysing the future global energy system, among the most important drivers to be considered are the development of the global economy, the development of energy technologies and the availability of energy sources. All the assumptions related to these factors can be very significant for the analyses, in particular concerning the future role of specific technologies. Therefore, very balanced base estimates are needed for credible model studies, and sensitivity analyses should be included whenever possible. In the EFDA model, the base projections for the global economic and demographic developments are close to the Reference scenario of the IEA s World Energy Outlook 2004, as well as to the IPCC B2 marker scenario. Both the technology and resource base data have been reviewed and updated during the work at VTT. Base estimates for fossil fuel resources have been compiled from the most commonly referred sources (e.g. IEA, WEC). Some of the most important estimates for non-fossil energy resources are presented in Table 1. Assumptions concerning uranium resources and the acceptability of new generation fission reactors may have a strong impact on the projected competitive position of fusion. The data for uranium resources are based on estimates made by IEA/NEA and IAEA, but with some conservatism concerning unconventional resources. Uranium extraction from seawater and the break-through of breeder reactors are considered likely too speculative. Global wind power resources have been comprehensively estimated by VTT s experts. 115

125 Data for biofuel resources correspond to the lower mid range of the estimates presented by e.g. IPCC. Further detailed reviews of the all resource base data will be made during 2007 among the collaborating research groups. For the analyses on the possible role of fusion power, estimates about the development of fusion power plant technology until 2100 are needed. The commissioning of the first commercial plants is assumed to take place around 2050, at the earliest. Assumptions concerning the economics of fusion power are roughly based on the cost analysis made in the EFDA PPCS study. Constraints imposed by tritium supply and total power plant construction capacity have been taken into account by applying gradually rising bounds on annual newly installed capacities. In the highest penetration case, the share of fusion power could reach about 20% of total world electricity generation by Potential of fusion under climate change mitigation: According to the decisions made by the EFDA steering group, the global model still needs considerable improvements before to be used for serious scenario analyses. Nonetheless, during the development work, the model has been already used for a number of analyses, which have also given quite illustrative preliminary results for the potential role of fusion under climate change mitigation policies. In the scenarios, three different projections have been used for the investment costs of fusion power, based on the Table 1. Key assumptions related to power generation from non-fossil energy sources. Energy source Assumed resource base limitations Constraints on capacity Uranium (fission power) Lithium (fusion power) Category RAR EAR-1 EAR-2 Speculative Unconventional Total t EJ Translated to annual capacity expansion constraints Regional limits (global total GW) Global and regional limits on annual new installations Wind power None assumed Large regional potentials (global total 12,000 GW) Solar power None assumed Large regional potentials Biofuels (all energy) Regional max. annual yields Global total potentials: Crops ~200 EJ/a Residues ~60 EJ/a Total ~260 EJ/a No definite constraints Constraints on market penetration Endogenous within capacity limits Endogenous within annual growth limits Max. 35% of market by season (Canada: 50%) Max % of market by season Endogenous 116

126 range of estimates presented in the EFDA PPCS study, and with constraints on new capacity as mentioned above. The mitigation target for global warming was set to a maximum of 2 C increase in the global average temperature by the year According to the results, fusion power would become competitive under all the three projections for the costs of fusion. In the highest cost case, the global electricity generation from fusion plants would reach about 9,000 TWh, but using either of the lower projections, it would exceed 20,000 TWh in Using the lowest cost projection, fusion plants would become commercially attractive already around The resulting development of CO 2 emissions is shown in Figure 1 for the Baseline scenario and the 2 C mitigation scenario corresponding to the mid cost projection for fusion power. The contribution of fusion to the emission reductions cannot be definitely calculated, but in this scenario it can be roughly estimated at about Gt CO 2 in Figure 2 presents the overall development of global electricity generation by main technology category in the mitigation scenario, using the mid estimates for the economics of fusion. In this case, the share of fusion power reaches about 20% of total world electricity generation by 2100, corresponding almost to the assumed maximum construction pace after Fusion and wind power would attain approximately equal global market shares. In the most optimistic case, commercial investments in fusion power would start about 15 years earlier. In terms of present value, the total direct economic benefits from the introduction of fusion power could be up to USD 500 billion by the end of the century, using our scenario assumptions. This would well outweigh e.g. the costs from a 50-year R&D programme with annual budgets of USD 20 billion. Conclusions: The preliminary scenario analyses with the model already indicate that fusion power can have a very significant role in the achievement of global climate change mitigation targets. Figure 1. Development of global CO 2 emissions in a scenario with 2 C limit for global warming and mid estimates for the economics of fusion power. 117

127 Electricity generation, PWh Other Wind Solar Fusion Fission Hydro CON+CCS CON-Gas/Oil CON-Coal CON-BioIGC CON-Bio CHP-Gas/Oil CHP-Coal CHP-Bio Figure 2. Development of global electricity generation in a 2 C limit scenario. Using reasonably favourable assumptions concerning renewable energy potentials, fusion power could still reach a 20% share of global electricity markets by The present value of the direct economic benefits alone may well exceed the R&D budgets needed to achieve the competitive position for fusion. Indirect economic effects and spill-over benefits would probably be even much higher. Further analyses with the model will address the sensitivities of the conclusions to all of the key assumptions. Nevertheless, the current version of the model is still in the need of many revisions, updates and improvements, which are currently ongoing. To ensure that the model can be used for credible analyses under the EFDA Socio Economic Studies, the EFDA steering group has organized the further development work into well-defined sub-tasks assigned to each of the collaborating research groups. VTT was assigned the task of updating that data for all of the industrial sectors. A benchmarking exercise will also be carried out with the model in order to validate all the important baseline assumptions. The model is expected to be ready for serious analyses during the latter half of External costs of fusion be socially justified energy production is required to be safe and clean. Actually it is not possible to find any absolutely safe and clean energy alternative. Fusion might be a future long term energy source, because resources of fusion fuel are rather unlimited and environmental impacts are estimated to be relatively small. However, fusion requires much scientific and technical development work. Also production costs (m /kwh) of fusion energy are estimated to be relatively high. Environmental costs (m /kwh) of fusion energy as studied in SERF since 1998 are estimated to be lower than for most other alternatives. Therefore the total costs including environmental costs are estimated to be competitive. Economic valuation might therefore be considered to support fusion. The work of VTT has included the development of models and methods for the estimation of global long-term impact of releases from fusion energy life-cycle. Fusion power plants were assumed to be installed around 2050 and assumed to produce MW electricity. Some most important environmental impacts have been still re-estimated in the SERF4 task. 118

128 5.1.3 Waste disposal or clearing of materials It is important that the environmental questions regarding production of fusion energy are presented so that public has confidence on that what is told about fusion. One problem might be that produced radionuclides have to be disposed and this question should also been openly discussed. Radionuclides are not produced when atoms fuse into helium in the fusion energy production. However, radionuclides produced due to bombardment of materials by neutrons are an important part of the life cycle of fusion energy production. Some other parts of the life cycle also cause other impacts, but these somewhat conventional impacts e.g. due to transportation of materials or building of power plants are not as interesting especially from the point of view of public opinion about the acceptability of fusion. Totally about 1 PBq of C-14 is estimated to be produced by neutron bombardment during a fusion reactor life of about 30 years. New materials are not very much better in that respect. Naturally born C-14 has a production rate about 1 PBq/a in the atmosphere due to cosmic rays. The natural C-14 has cycled in the environment with the carbon flows and the natural C-14 background in the atmosphere has been estimated to be about 140 PBq before the nuclear tests started after about External costs have been studied and also local scale radiation doses due to disposal. Clearing and recycling has been considered to be possible after 100 years of cooling in Power Plant Conceptual Studies (Forrest 2003). This has been questioned in the light of regulations in this SERF task. Activity concentrations (Bq/g) of studied radionuclides are much higher than limits in the present international regulations. Clearing possibilities for fusion waste are shortly studied in the light of Finnish and international regulations. Regulations seem not to be very favourable in that respect. It is therefore necessary that the clearing of materials suggested in the PPCS studies are evaluated further. When disposed the impacts due to radionuclide inventories can be diminished to be very small. Barriers limiting the radionuclide dispersion have then to be effective enough. Local scale considerations indicate that the C-14 retention should last more than 20,000 years so that the limits for individual dose rates will be low enough. In the PPCS studies the clearing has been estimated to be possible after 100 years. Based on external costs the C-14 inventories require about the same retention as in the case of local impacts. (External costs are estimated using long term integration of global impacts.) The local scale impact due to possible Finnish repositories has been studied. Different disposal alternatives were explored in the study. The objective was to study disposal in a rather general way. Local wells and lakes are considered in the scenarios. It might be concluded that the Finnish C-14 limits for annual emissions have been chosen using very pessimistic assumptions, at least for deep repository case if lake scenarios are used. Finnish and Swedish scenarios possibly overestimate accumulation of carbon into fish. Critical individual actually consumes a considerable part of yearly C-14 emission when scenario assumptions are used. Annual emissions due to the normal operation of fission reactors are then allowed to be much higher than annual emissions from waste disposal. The emissions of the C-14 due to a normal operation might be about of the same order (1TBq/a) for fusion and fission, if water cooling has been used. In the disposal the about same amount 1 TBq might be estimated to be high in fission waste. If helium cooling is used then the C-14 emissions in normal operation are avoided in the fusion energy production Time horizon questions In the future environment, carbon dioxide levels will ultimately increase. It seems probable that, in spite of the Kyoto Protocol and other later protocols, the CO 2 concentration level in the atmosphere will close to double by Rather strict 119

129 stabilisation targets have been suggested especially by the European Union, but emission reduction targets in the coming decades will be difficult to reach. The target of restricting the global warming to the two degrees requires the stabilisation of total greenhouse gas concentrations to lower level than 550 ppm equivalent concentration. Thus especially carbon dioxide concentrations have to be stabilised to very low level. In 2050, when fusion might be a possible energy production alternative, the need to restrict carbon dioxide emissions is likely to be greater than today. Actually it is not easy to find out measures to restrict considerably the carbon dioxide concentrations in the coming decades. Fusion economy with thousand 1500 MW fusion plants in the year 2100 might decrease the atmospheric concentrations by only about 10 ppm by 2100, if it replaces coal use. On the other hand, it is not easy to find any single measure to restrict the carbon dioxide concentration. In the long term fusion might help considerably in the restriction of the global warming. In the continuing SERF work fusion will be studied as a part of a energy system using the global TIMES model. Long term global impacts of fusion and fossil alternatives have been studied using developed concepts tonne-years for fossil alternatives and Becquerel-years for fusion. The index for the collective dose commitment is analogous to the index AGWP, Absolute Global Warming Potential, and could therefore be named AGDP, Absolute Global Dose Potential. GWPs are generally used in climate gas emission calculations considered in climate negotiations. The analogy between AGDP and AGWP could be used in many ways, especially in the comparison of energy production alternatives. If global long-term impacts are evaluated, these indices can be used for the comparison of fossil, fission and fusion alternatives. It seems evident that global impacts for fossil alternatives are much higher also when estimated using assumptions which favour fossil alternatives. The concept of Becquerel years in the atmosphere has been used in the evaluation of C-14 transfer. The values of the quantity of Becquerel years in the atmosphere increase when CO 2 concentration increases. This means that the lifetime of C-14 in the atmosphere increases and, in the case of continuing emissions, it accumulates for longer time spans in the atmosphere. The Suess effect or fossil fuel effect is often estimated to cause reduction of the ratio of radiocarbon C-14 to stable carbon, but it will be rather temporary and the impact on collective doses will be rather small, in the long term. The diluting impact or Suess effect will in some decades decrease due to the increased lifetime and the total impact will be that dose impact will decrease relatively little in the case of the double concentration level compared to the preindustrial concentration. Restriction of global warming is one very important constraint which causes that fusion energy might have an important role in the energy production after some decades, when technical problems have been solved. However, it is very important to study also the environmental impacts of fusion and especially production and management of radionuclides, particularly C-14. It might be possible to find optimal solutions for materials in that respect at an early stage and also discuss the possibility to avoid the disposal of radioactive fusion waste. 5.2 Fusion Power Plant Conceptual Studies Helsinki University of Technology R. Salomaa, S. Sipilä, V. Tulkki and G. Zemulis Safety assessment of conceptual fusion power plants EFDA has designed several conceptual fusion power plant models to assess fusion economy and safety and to guide the R&D technology needed to improve fusion reactors. The reactor candidates with successively more advanced features include: Models A and B that use water-cooled Li-Pb and He cooled pebble bed blanket, respectively, and are based on presently anticipated plasma performance and on near-term technol- 120

130 ogy. The more advanced Models C and D assume considerable improvement of plasma performance and involve double coolant loops and long term technology, such as combined He and He-Pb cooling and SiC structural parts. The plant performance, efficiency, re-circulating power, and lifetime of plasma facing components improve successively from Model A to Model D. Subtask 4 TRP-PPCS4 of the Power Plant Conceptual Studies involved thermal-hydraulic analysis of various accident scenarios in Model A. The APROS simulation environment was applied to build a thermal-hydraulic model for such a fusion power plant including the main first-wall and breeding blanket structures, vacuum vessel, the primary and secondary cooling circuits, the pressure suppression systems and the drainage tanks. We simulated several accidents to assess the safety of this kind of reactor concept and to properly dimension its safety systems. As an example, one scenario involved a loss of coolant flow leading to a first wall breach and pressurisation of the vacuum vessel (LOFA + In-vessel LOCA). Calculations revealed strong pressure transients, which called for design changes. One goal was to verify the adequacy of the containment: it remained intact at least 14 hours without any mitigating efforts. Radioactive releases were also estimated. The results were incorporated in the Power Plant Conceptual Study, Final Report on Safety Assessment of PPCS Plant Models A and B and in international conference contributions. As a spin-off, the extension of APROS to fusion reactors further validated the code and introduced needs to extend the modelling to supercritical conditions and to Li and W reactions inside the vacuum vessel Safety and economy requirements of fusion power reactors in co-existing advanced fission plants Nuclear fusion and fission are among the few alternatives which can globally solve long-term energy problems. Fusion energy is abundant and inherently safe but requires extensive technical development to achieve its economic viability. Advanced fission reactor concepts are closer to commercial feasibility, but their political acceptance has to be settled. Decisions concerning ITER and the renaissance of fission reactors have advanced nuclear energy in general, and several concrete evolutionary steps of various genres of fission and fusion reactors are taking place. The new Finnish reactor unit, Olkiluoto 3, is expected to operate from 2010 to In this time range a new fleet of advanced fission reactors of Generation IV and several stages of fusion reactors from ITER to DEMO will emerge. We have discussed the implications of such co-existing reactor genres and their general viability. The main focus has been on gross features of economy and safety, but some cross-cutting issues of technologies were also observed. Comparisons to present EPR reactors were made for fusion and for Generation IV using various pricing methods: mass flow analyses together with engineering, construction and financial margins formed one method, and another one was based on simple scaling relations between components or structures with common technology level. The main issues we have preliminarily addressed are: 1) The newest LWR reactors have a very long lifetime during which ITER, DEMO and even the first commercial fusion power plants and the Generation IV power plants could co-exist. 2) For all reactor types, the performance of present fission power plants will guide the safety and economic reference goals which will vary with competition and changing socio-economic environment. 3) Many of the technology problems share common R&D between fusion concepts and those of Generation IV: for instance, EFDA fusion Models C and D and Generation IV candidates, such as Pb or He cooled fast reactors and He cooled very high temperature reactors, face the same technology problems. 4) In the long run, the alternative development lines could evolve into symbiotic fusion-fission systems. In all these studies, fusion competitiveness has to be improved in terms of plant availability and internal power recirculation. As a comparison one 121

131 ITER G IV DEMO FUSION???? constr, OL3 operation decommissioning G4 DEMO G4 PROTO-1 G4 PROTO-2 ITER DEMO PROTO COMM. FUSION Figure 1. During the lifetime of the newest Generation III+ fission power plants, completely new types of Generation IV advanced fission reactors and fusion reactors will emerge. All types must satisfy the actual socio-economic competitive environment. The actual timing is, of course, only indicative. should note that the best fission plants have an availability of about 95% and an internal power recirculation of the order of 3 4%. The operation and maintenance solutions of Model C and D show the right way to increase their availability to a reasonable range of 75 80%. The internal parts of the reactor are segmented and modularized to speed up maintenance. A huge rise in LWR fuel costs would improve the competitiveness of first the Generation IV breeder options and thereafter also the fusion plants. A further cost reduction could be obtained if safety related LWR systems, such as the containment and equipment for severe accident mitigation e.g. the core catcher, were not needed in Generation IV or fusion power plants. The main conclusions of these studies were reported in the 21 st IAEA Fusion Energy Conference in Chengdu in Improvements in the economy of various new nuclear power plant concepts require: lower direct construction costs (optimal thermal size) minimized construction times improved efficiency (high temperatures) longer economic life time (40 years or more) if compatible with variable costs maximum plant availability (reliability issues) shorter overhaul times (segmentation) smaller re-circulation factor. Fusion R&D is facing huge challenges in these areas. Internal use of current drive and heating power should be cut down to circa 100 MWe for GWe-level plants and new O&M methods have to be developed. Some of the safety systems of present fission power plants can be abandoned and replaced by less expensive designs in fusion reactors provided that their safety case is shown. 122

132 6 Industrial Projects 6.1 Development of ITER superconducting Wires Superconducting wires from Luvata Pori Oy Luvata Pori Oy Ben Karlemo The company Luvata, former Outokumpu Copper Products, is a metal company working mainly with copper products. One of the business units is the Superconductors manufacturing low temperature superconductors. The two main areas of business for the superconductors are the magnetic resonance imaging (MRI) market and the scientific projects market. The spine of ITER will be the magnet system including the superconducting wires. Luvata and Luvata Pori Oy have taken an active part in the development of superconducting advanced Nb3Sn wires and conductors for ITER as a partner with EFDA CSU Garching. The ITER experimental reactor for fusion energy will need four different magnet systems to work. The systems are, the 18 toroidal field (TF) coils surrounding the reactor donut that will confine the plasma, the 6 central solenoid (CS) modules together with the 6 poloidal field (PF) coils that will help to drive the plasma and fusion reaction together and the correction coils (CCs) that will be used to smoothen the field homogeneity flaws existing from manufacturing of the other magnets. Two different superconductor wires will be used for these magnet systems, the NbTi and the Nb3Sn wires. The difference between these two types of wires are that the NbTi wire is much more ductile and is therefore much easier to be used during winding and manufacturing of magnets than the better performing, but very brittle and more expensive Nb3Sn wires. The NbTi wires will therefore be used in the lower field magnets as the PF and the CCs coils and the Nb3Sn wire in the TF coils and the CS modules. The total amount of superconductors that will be needed for the ITER magnet system is 520 tonnes of Nb3Sn and 240 tonnes of NbTi wire. The manufacturing of the Nb3Sn wires has been shared among all the participating teams except India and the NbTi wires between China, Russia and Europe. When squeezing cost in the project the temperature margins of the magnet systems where reduced. The reduction created a concern for the performance and therefore a programme was started to once again raise the temperature margin by improving the wire performance. The European Home Team led by the European Fusion Development Agreement Close Support Unit (EFDA CSU) in Garching Germany has during the last years been the number one contributor to the development work for the ITER project including the field of Nb3Sn wires. An activity to develop the industrial manufacturing abilities of the European superconducting wire industry was started in the year of 2003 with six companies in Europe for the TF coils. Luvata Pori Oy was one of the companies that took part in this exercise and produced in total about 250 kg of strand for this activity. The critical current target of the wire was increased to 1100 A/mm2 from 750 A/mm2 (@ 12 T and 4.2 K) and the AC-losses should still be kept reasonable. The two main industrial production methods to produce Nb3Sn wire are the bronze method and the internal tin method. The main difference in the two methods are that in the bronze method a copper tin bronze with a tin content of typically 15 w-% is used. This limits the formation of Nb3Sn compared to the internal tin where a tin content of 25 w-% can be achieved. The drawback of the internal tin method compared to the bronze method is, that after the insertion of the tin 123

133 inserts the wire cannot exceed about 200 degree of centigrade because a meltdown of the tin has to be avoided. The ductility of the bronze wire can be maintained very well because of the possibility to heat treat the wire during the drawing process. The new set current targets where such that the only way to achieve the necessary current values was to use the internal tin method. Luvata Pori Oy managed to meet the target using the internal tin method. A cross section of the wire can be seen in Figure 1. The wire is externally stabilised with copper and consists of a 37-bundle design with over 6000 filaments. The bundle area is separated from the copper area by a tantalum barrier to maintain the integrity of the copper area. This also reduces the AC-losses. As conclusion it can be pointed out that one major goal of the company Luvata is to aim for a partnership beyond metals with its customers and in the business unit of superconductors this is especially well realized in the scientific project work for ITER where a good working relationship has been established on many levels between the different organisations involved in the ITER preparation Cross-checking of the superconducting strand acceptance tests Tampere University of Technology Institute of Electromagnetics Risto Mikkonen, Iiro Hiltunen Background: The conductor material for the different superconducting magnets in ITER will be either NbTi or Nb 3 Sn. The primary objective of this task is to check and verify if the performance of the delivered Nb 3 Sn strand from the advanced strand procurement action is according to the new strand specification. Institute of Electromagnetics in TUT has facilities to characterize Nb 3 Sn super-conductors designed according to the specifications settled by EFDA. The temperature of the samples can be set with VTI device which enables an operating temperature between 1.6 K 300 K. The external magnetic field is generated with a new hybrid solenoid providing the magnetic flux density up to 16 T. A new sample insert (including the current leads and instrumentation) was made at TUT. During the measurements the ITER type sample holder was used. For the benchmarking EFDA provided samples of Nb 3 Sn strand. The heat treatment of the sample was made in Luvata. The benchmarking tests were made twice in accordance to the specifications made by EFDA. Objectives: According to the original plan TUT had to characterise samples delivered by EFDA. The main objective of the work was to perform measurements of the voltage-current characteristics of Nb 3 Sn superconducting strands in liquid helium (T = 4.2 K) in high magnetic field (B = 12 T). Figure 2 shows a cross-section of such a strand. Figure 1. Cross-section of the Nb3Sn superconducting wire for ITER by Luvata Pori Oy. 124

134 Figure 2. Cross-section of a Nb 3 Sn superconductor. Results: The measuring program included the following items: The thickness of the Cr-plating around the conductor was determined by EDS studies. The Cu/non-Cu ratio was measured with a scanning electron microscope image of the strand cross section with an optical microscope. The twist pitch of the conductor was determined by etching a section of the strand with HNO 3 and the pitch angle was estimated from the optical image of the strand. The residual resistance ratio of the copper in the conductor was measured at room temperature and at 20 K utilising a cryocooler. The critical current of the sample was measured at liquid helium temperature with an external magnetic field of 12 T with a high precision data acquisition system with a nanovolt meter. From the measured voltage current curve the conductor n-value (express the transition behaviour from superconducting to normal state) was also determined. The final report has been delivered to EFDA. Figure 3 shows a measuring set-up. The high field magnet is immersed in liquid helium inside the cryostat. The benchmarking tests indicated that TUT has the facilities and know-how for carrying out the measurements specified in the contract and TUT has now a full readiness for continuing the co-operation which is scheduled to Figure 3. The sample holder (left) with measuring cryostat. 125

135 6.2 Advanced Fabrication Methods for Vacuum Vessel Hollming Works Oy Mika Korhonen Company background Hollming Oy main business divisions are Hollming Works mechanical engineering, Crystal Pool shipping and Auramarine design and manufacture of auxiliary equipment. Hollming Works has a long experience of manufacturing demanding stainless steel vessels and machines for e.g. pulp and paper industry, offshore and shipbuilding industry, power production and mining industry. Our latest deliveries for energy industry are power units of windmills. Typical examples of large and heavy products are components for oil production rigs (e.g. truss connections and fairlead supports), debarking drums for pulp mills and large-scale pressure vessels for power production. Our expertise can be utilized in welding and machining of different materials, assembly in workshops or on site, as well as in design. Hollming Works is a pioneer as developer and applier of new welding technologies. We possess e.g. several years experience in demanding narrow gap and robot welding. Hollming Works is specialized in the use of stainless steel, carbon steel, aluminium and other special materials, such as nickel alloys, titanium, zirconium and tantalum. Machining capacity is modern, very flexible and customer-oriented, thus improving cost-efficiency of operations. Hollming Works has a long experience of manufacturing propeller nozzles, rudders and components for propulsion equipment and has developed special cold forming know-how to produce 3D shapes using press forming. In addition to high quality products and profitable operations, Hollming Works invests in environmental and industrial safety aspects at all stages of the processes. We are certified in compliance with the SFS-EN ISO 9001: 2000 quality system, SFS-EN welding system, ISO environmental system, Pressure vessel licenses (ASME, TÜV, PED, Chinese manufacturing license) and customer specifications. In addition to design and manufacture of components, Hollming Works provides test assemblies of machines and installation or installation supervision on site Hollming Works Finnish technology partner in Fusion energy development Hollming Works has actively participated in domestic and international co-operation around ITER project aiming to demonstrate that building a large-scale fusion reactor is scientifically and technologically possible. National research projects: Several national research projects supported by Tekes funding have been launched to develop technologies needed to support ITER deliveries, e.g. development of advanced welding technologies (e.g. narrow gap welding technologies), 3D cold forming of vacuum vessel inboard walls and 3D metrology for large-scale components. One of the main targets has been to improve the know-how of the narrow gap welding technologies for difficult to weld materials and very large wall thicknesses. Also mechanization and semi-automatic processes have been studied and tested thoroughly during project. In Figure 1 can be seen one of the most efficient narrow gap welding method tested production rates up to 18kg/h with low angular and linear distortion levels. 126

136 New advanced fabrication methods have been presented as a result of the co-operation with Metso Oyj. EFDA has accepted an Issue Card TVV-7/WBS 1.5 Rules and regulations for the fabrication of Vacuum Vessel Sectors where some new ideas for possible fabrication methods were presented. Figure 1. One of the most efficient narrow gap welding method. EFDA Task 3D Cold Forming: The results in EFDA Task Advanced Techniques for Vacuum Vessel Sector Manufacturing, Subtask 2: Assessment of 3D forming of inboard walls has shown the capability to complete demanding research activities needed to support ITER project. This research project was set up for estimating the largest size of the stainless steel plates with the thickness of 60 mm which could be deformed into the required shape by 3D cold pressing and the pressing forces needed for this bending. The first results of the task were presented at the 6th ITER Vacuum Vessel Manufacture Design and Assembly EU R&D Meeting in Magdeburg Figure 2. The simulation test results of the 3D cold forming project. 127

137 By the 3D cold pressing large steel plates are deformed by 3D bending into the required shape and the final vessel can be manufactured by welding together these large steel plates. This route would significantly reduce the amount of welding and the related problems during the fabrication of the vacuum vessel. In figure 2 can be seen some simulation test results of the 3D cold forming project. Networking: Hollming Works has business network internal and external RTD resources and research network with research institutes and universities to serve ITER deliveries. Hollming Oy commits Hollming Works efforts in ITER project. 128

138 ANNEX A Fusion Projects and EFDA Activities Table A1 gives a summary of FUSION projects and their funding during the programme period The total figure of the FUSION research projects is about 14,9 million and the total figure for the FUSION related industrial projects (including industry co-ordination) is about 3,5 million. The most of the work carried out in the FUSION projects consists of the Physics Programme, EFDA JET Orders, EFDA Technology Tasks and Contracts. Table A2 summarises the EFDA Article 5.1a Technology Tasks carried out by the Association Euratom-Tekes during Table A2 shows that the focus area of the FUSION programme is vessel/in-vessel technology. Table A2 shows the preferential support (PS), which is awarded for capital investments, hot cell work and work dealing with tritium and beryllium contaminated material. The total value of preferential support items is 2,216,000 during the four year period. Table A3 shows the EFDA Article 5.1b and Article 7 Contracts during and Table A4 the EFDA JET S/T Orders and JOC (JET Operating Contract) secondees to operator (UKAEA). Table A1. FUSION programme funding including industrial projects, EFDA secondments and staff mobility visits in Activity Projects 2003 (k ) 2004 (k ) 2005 (k ) 2006 (k ) Total (k ) FUSION Co-ordination Fusion Physics Underlying Technology EFDA Physics Integration EFDA Vessel/In-Vessel EFDA T-Breeding & Materials EFDA System Studies EFDA Technology Contracts EFDA JET Technology EFDA JET S/T EFDA Secondments Staff Mobility Visits Industrial Projects Total (keuro)

139 Table A2. EFDA Article 5.1a Technology Tasks by the Association Euratom-Tekes in The total volume and the value of the Preferential Support (PS) items are given in thousand Euros ( 1000). Task Reference Task Area / Task Title Tot/PS Institute TW3-THHE-CCGDS1 Physics Integration / JET Technology k Coaxial cavity gyrotron, 170 GHz, 2 MW, CW, analysis on the selection of the operating mode 30 TKK JW3-JET-FT-3.10 Material transport and erosion/deposition in the JET torus 200/150 VTT, Diarc JW4-FT-3.15 Material transport and erosion/deposition in the JET torus 220/170 VTT, Diarc TW4-TPP-CARWMOD Molecular dynamics simulations of carbon and tungsten sputtering 190 UH TW4-TPP-TILCAR Characterisation of erosion/ re-deposition balance in ITER-relevant divertor 150/60 VTT, Diarc tokamaks and of PFCs after hydrocarbon removal by oxidative methods TW4-TPDS-DIADEV2 Development of micromechanical magnetometer for ITER 50 VTT TW5-TPP-TILCAR Characterisation of erosion/redeposition in ITER relevant first wall materials 150/60 VTT,UH, Diarc JW5-FT Material transport, erosion / re-deposition in JET torus 190/150 VTT, Diarc 6TW6-TPP-CARTIL Characterisation of erosion/re-deposition balance in ITER-relevant divertor 150/60 VTT,UH, Diarc tokamaks JW6-FT-3.27 Material transport, erosion / re-deposition in JET torus 190/150 VTT,UH, Diarc TW6-TPP-BETUNCMOD Molecular dynamics simulations of mixed material formation at the ITER 255 UH divertor JW6-FT Material transport in SOL, erosion/re-deposition in JET torus 140/120 VTT, Diarc Vessel / In-Vessel TW3-TVM-COMOW Characterisation of the material properties of high Z plasma facing material: 150 UH,VTT, Diarc effect of carbon and oxygen impurities on hydrogen retention and diffusion TW3-TVD-CUSS Qualification testing of new CuCrZr/SS tube joint 40 VTT TW3-TVM-JOINT Characterisation of the CuCrZr/SS joint strength for different blanket 55 VTT manufacturing conditions TW3-TVM-COFAT In reactor fatigue testing of copper alloys 265/80 VTT, Luvata TW3-TVM-TICRFA Effect of low dose neutron irradiation on Ti alloy mechanical properties 55/40 VTT TW3-TVR-WHMAN Design and specification of a water hydraulic manipulator 300/100 TUT, Adwatec TW3-TVR-MOVER Design and specification of the cassette multi-functional mover (CMM) 250 TUT TW3-TVV-EBEAMS Further development of e-beam welding with filler wire 350/100 VTT TW3-TVV-ROBASS Upgrade robot to include linear track 600 /272 LUT TW4-TVR-WHMAN Development of water hydraulic manipulator 200 /110 TUT, Adwatec TW4-TVM-COFAT2 In reactor fatigue testing of copper alloys 150/40 VTT TW4-TVV-ROBOT Further development of stability, safety and accuracy of IWR robot 200 LUT TW5-TVB-PHCSMU Manufacture of CuCrZr/316L SS mock-ups by powder HIPing 150 VTT, Metso TW5-TVV-IWRFS Demonstration of IWR feasibility 400/200 LUT TW5-TVM-CUSSPIT Testing of irradiated CuCrZr/SS joints produced under different blanket 90/54 VTT manufacturing conditions TW5-TVM-SITU2 In reactor tensile testing of Cu and CuCrZr alloy 125/40 VTT, Luvata TW5-TVR-WHMAN Development of water hydraulic manipulator 250/100 TUT, Adwatec TW5-TVR-CMMHLC Provision of test models and software for the CMM system 300 TUT 130

140 Table A2. continues... Task Reference Task Area / Task Title Tot/PS Institute TW6-TVM-NAJT Irradiation of CuCrZr/SS and CuCrZr/Be joints at 300ºC produced under 150/60 VTT different blanket manufacturing conditions TW6-TVR-DTP2DEV Development of subsystems and tooling in support of remote handling trials 400/100 TUT, VTT TW6-TVR-RADTOL Radiation tolerance of water hydraulic components 100 TUT TW6-TVR-CMMHLC Specification, design, implementation and testing of the CMM/SCEE 300 TUT high-level control software Underlying Tech 1 Further development of novel methods and studies on radiation effects and 150 VTT verification of specimen size effects Underlying Tech 2 Water hydraulic components further development 150 TUT Underlying Tech 3 IWR testing and development 100 LUT Tritium Breeding and Materials TW3-TTMI-003 IFMIF neutronics 50 VTT TW5-TTMI- 004 IFMIF design integration 3D radiation shielding analysis and neutronics 37 VTT TW5-TTMI- 001 IFMIF Test Facilities - neutronics 63 VTT TW5-TTMS 007/13a Radiation damage in EUROFER: FeCrHe thermodynamics 75 UH TW5-TTMS-005b Structural materials: Rules for design, fabrication and inspection 178 VTT TW6-TTMS 007 Modelling of radiation effects in FeCr in presence of H and C 50 UH System Studies TW5-TRE-FESO-1 Identification and comparative evaluation of fusion and other possible future 80 VTT energy production alternatives TIMES global modelling TW5-TRE-FESO-2 Fusion as a part of energy system 58 VTT TW5-TRE-FESO/C Identification and comparative evaluation of fusion and other possible future 58 VTT energy production alternatives TIMES global modelling TW6-TRE-ETMIND EFDA-TIMES Model: Industry sector update 67 VTT 131

141 Table A3. EFDA Article 5.1b Technology Contracts by the Association Euratom-Tekes and Article 7 Contracts by industry in EFDA Reference Task Area / Task Title Tot/PS Institute Physics Integration / JET Technology k EFDA/ Coordination of activities of the EFDA technology workprogramme 61 Tekes, Fortum EFDA/ JET EP: Diagnostics tritium retention studies, prepare and perform 39 VTT,Diarc smart tile coating processes Technology EFDA/ JW5-TA-EP-BEW-02: Development of W-coatings for JET divertor tiles (S/T Order) 60 VTT,Diarc Technology EFDA/ TW6-TPDS-DIASUP: Support of diagnostic design for ITER: Sub task VTT,TKK,FMI, Prizztech EFDA/ TW6-TPHI-ICFS: Faraday shield RF modelling and RF sheath dissipation 40 VTT EFDA/ TW6-TPP-ERITER: Surface composition model development for ITER 25 TKK limiter erosion /redeposition simulations EFDA/ TW6-TPO-RIPLOS: 3-D calculations of ion losses and wall loads in ITER 167 TKK due to toroidal field ripple JW6-TA-EX-ITC-03 Integration of transport and MHD codes at JET (S/T Order) TKK JW6-TA-EP2-NPA01 Upgrading of the NPA diagnostics system (S/T Order) 80 TKK, VTT Vessel / In-Vessel EFDA/ (Am 2) Ultrasonic tests of primary first wall panels and mock-ups 178 VTT EFDA/ TW4-TVR-DTP2.1: CMM and SCEE design updates 100 TUT EFDA/ TW4-TVR-DTP2.2: DTP2 platform design adaptation and QA system 149 VTT development TW4-TVD-PFCATT/1: Testing of revised PFC multilink attachments 135 TUT EFDA/ TW5-TVV-SECWEL: Preparation of VV segment welding mock-up 150 LUT EFDA/ TW5-TVA-IHYB: Industrialisation and weld quality issues of high productive laser\ arc hybrid for thick section welding of ITER grade SS material 150 VTT EFDA/ TW6-TVM-LIP1: Modification of ITER materials documents and assessment of materials data 30 VTT EFDA/ TW6-TVR-DTPTA: Host activities in the connection to DTP2 test facility 596 VTT EFDA/ TW6-TVR-DTP2OP: Host activities in connection with the DTP2 test facility 447 VTT, TUT EFDA Art. 7 Contract DTP2 Structure 550 TP-Konepaja EFDA Art. 7 Contract 3D forming of VV sector 80 Hollming Works Magnets EFDA/ TW3-TMSC-ASTEST: Test of advanced Nb3Sn strand 40 TUT EFDA Art. 7 Contract Development of Sn3Nb superconduction wire for ITER 160 Luvata Oy EFDA CSU Secondments EFDA CSU Garching Secondments Tommi Jokinen, Assembly Hannu Kaikkonen, Project Control Pertti Pale, Project Control Hannu Rajainmäki, Magnets Mikko Siuko, Remote Handling VTT Fortum PPC Luvata TUT 132

142 Table A4. EFDA JET S/T Orders and JOC secondments by the Association Euratom-Tekes in JET Reference Task Area / Task Title Vol/k Institute JW2-TA-EP-TRS-01 Diagnostics tritium retention studies, prepare and perform smart tile coating 39 VTT processes JW2-O-TEKE-05B JET experimental campaigns C5-C7 32 TKK, VTT JW3-JET-FT-3.10 Material transport and erosion/deposition in the JET torus 200/150 VTT JW3-O-TEKE-07C JET experimental campaigns C8-C TKK, VTT JW3-OEP-TEKE-08A Diagnostics tritium retention studies 25 VTT JW4-FT-3.15 Material transport and erosion/deposition in the JET torus 220/170 VTT JW5-EP-BEW-TEKE Development of W-coatings for JET divertor tiles 60 VTT, Diarc Technology JW5-FT Material transport, erosion/re-deposition in JET torus 190/150 VTT JW5-TA-EP2-ILW-02 Development of Be-coatings for Inconel targets and Be marker tiles for the ITER-like wall experiment 19 VTT,Diarc Technology JW5-O-TEKE-10A JET experimental campaigns C15-C TKK, VTT JW6-FT Material transport in SOL and erosion/re-deposition in JET torus 140/120 VTT JW6-OEP-TEKE-13 Upgrading of the NPA diagnostics system 96 TKK, VTT JW6-O-TEKE-14 Integration of transport and MHD codes at JET 12 TKK JOC UKAEA Secondments JOC UKAEA secondments Antti Salmi, TKK, code development, modelling Johnny Lönnroth, TKK, code development, modelling Marko Santala, TKK, Diagnostics EFDA JET secondments EFDA JET Secondments Mervi Mantsinen, Deputy Task Force Leader (heating), scientific coordinator Tuomas Tala, Deputy Task Force Leader (transport), scientific coordinator Jukka Heikkinen, scientific coordinator Karin Rantamäki, scientific coordinator Marko Santala, scientific coordinator TKK VTT VTT VTT TKK 133

143 Annex B Institutes and Companies Tekes, the Finnish Funding Agency for Technology and Innovation Kyllikinportti 2, Länsi-Pasila P.O. Box 69, FIN Helsinki, Finland tel ; fax Juha Linden [email protected] Fusion Research Unit of the Association Euratom-Tekes VTT, Technical Research Centre of Finland VTT Materials Performance Otakaari 3A, Espoo P.O. Box 1000, FIN VTT, Finland tel ; fax Seppo Karttunen [email protected] VTT Materials Performance Kemistintie 3, Espoo P.O. Box 1000, FIN VTT, Finland tel ; fax Seppo Tähtinen [email protected] VTT Production Systems Tuotantokatu 2, P.O.Box 17021, FIN Lappeenranta, Finland tel ; fax Veli Kujanpää [email protected] VTT System Engineering Tekniikankatu 1 P.O. Box 1300, FIN Tampere, Finland tel ; fax Jorma Järvenpää [email protected] Arto Timperi [email protected] VTT Nuclear Energy Otakaari 3A, Espoo P.O. Box 1000, FIN VTT, Finland tel ; fax Petri Kotiluoto [email protected] VTT Energy Systems P.O. Box 1000, FIN VTT, Finland tel ; fax Antti Lehtilä [email protected] VTT Sensors Tietotie 3, Espoo P.O. Box 1000, FIN VTT, Finland tel ; fax Anu Kärkkäinen [email protected] Helsinki University of Technology (TKK) Helsinki University of Technology Advanced Energy Systems P. O. Box 2200, FIN TKK, Finland tel ; fax Rainer Salomaa [email protected] Helsinki University of Technology Automation Technology P. O. Box 3000, FIN TKK, Finland tel ; fax Aarne Halme [email protected] Tampere University of Technology (TUT) Tampere University of Technology Institute of Hydraulics and Automation Korkeakoulunkatu 2, P.O. Box 589, FIN Tampere, Finland tel ; fax Matti Vilenius [email protected] Tampere University of Technology Laboratory of Electromagnetics Korkeakoulunkatu 2, P.O. Box 589, FIN Tampere, Finland tel ; fax Risto Mikkonen [email protected] Lappeenranta University of Technology Laboratory of Machine Automation Skinnarilankatu 34, P.O.Box 20, FIN Lappeenranta, Finland tel ; fax Heikki Handroos [email protected] University of Helsinki Accelerator Laboratory P.O. Box 43, FIN University of Helsinki, Finland tel ; fax Juhani Keinonen [email protected] Kai Nordlund [email protected] 134

144 Industrial Companies Group: The Finnish Blanket Group consisting of Technip Offshore Finland, Diarc Technology Oy, Fortum Power and Heat Oy, Hollming Works Oy, High Speed Tech Oy, Kankaanpää Works Oy, Metso Powdermet Oy, Luvat Oy and PI-Rauma Oy Technology: Metal structures and plasma facing components Contact: Jari Liimatainen Group: The Finnish Remote Handling Group consisting of Adwatec Oy, Fortum Power and Heat Oy, Hytar Oy, PI-Rauma Oy, Platom Oy, Creanex Oy, Rocla Oy and Delfoi Oy. Technology: Remote handling, virtual reality, water hydraulics Contact: Timo Mustonen Company: ABB Oy Technology: Power and automation Contact: ABB Oy, P.O. Box 184, FIN Helsinki, Finland Tel ; Fax Ralf Granholm Company: Adwatec Oy Technology: Contact: Company: Technology: Contact: Company Technology: Contact: Company: Technology: Contact: Remote handling, water hydraulics, actuators and drives Adwatec Oy, Polunmäenkatu 39 H 9, FIN Tampere, Finland Tel ; Fax Arto Verronen [email protected] Aspocomp Oy Electronics manufacturing, thick film technology, component mounting (SMT), and mounting of chips (COB) in mechanical and electrical micro systems (MEMS) and multi-chip modules (MCM), PWB (or also called PCB), sheet metal manufacturing and assembly. Aspocomp Oy, Yrittäjäntie 13, FIN Klaukkala, Finland Tel ; Fax Markku Palmu [email protected] Creanex Oy Remote handling, teleoperation and walking platforms. Creanex Oy, Nuolialantie 62, FIN Tampere, Finland Fax , GSM Timo Mustonen [email protected] Delfoi Oy Telerobotics, task level programming Delfoi Oy, Vänrikinkuja 2, FIN Espoo, Finland Tel ; Fax Heikki Aalto [email protected] 135

145 Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: DIARC Technology Oy Diamond like DLC and DLC (Si, D) doped carbon coatings plus other coatings with potential plasma facing material in thermonuclear fusion machines. Diarc Technology, Olarinluoma 15, FIN Espoo, Finland Tel ; fax Jukka Kolehmainen [email protected] Ekono-Electrowatt/Jaakko Pöyry Group International consulting and engineering expert within the Jaakko Pöyry Group serving the energy sector. Core areas: management consulting, hydropower, renewable energy, power & heat, oil & gas, project services for nuclear safety and industrial processes P.O.Box 93, Tekniikantie 4 A, FIN Espoo, Finland Tel , Fax Vilho Salovaara [email protected] Elektrobit Microwave Oy Product development, test solutions and manufacturing for microwave and RFtechnologies, high-tech solutions ranging from space equipment to commercial telecommunication systems Teollisuustie 9A, FIN Kauniainen, Finland Tel , Fax Marko.Koski [email protected] Enprima Oy Design, engineering, consulting and project management services in the field of power generation and district heating. EPCM services. P.O. Box 61, FIN Vantaa, Finland Tel , Fax Jarmo Raussi [email protected] Company: Finpro Role: Industry activation and support Contact: Finpro, P.O. Box 358, Porkkalankatu 1, FIN Helsinki, FIN Pori, Finland Tel ; fax Company: Technology: Contact: Company: Technology: Contact: Pekka Tolonen [email protected] Fortum Nuclear Services Oy Nuclear Engineering Fortum Nuclear Services Oy, Keilaniementie 1, Espoo FIN Fortum, Finland Tel ; Fax Harri Tuomisto [email protected] High Speed Tech Oy Copper to stainless steel bonding by explosive welding High Speed Tech Oy, Tekniikantie 4 D, FIN Espoo, Finland Fax Jaakko Säiläkivi [email protected] 136

146 Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: ulttechnology: Contact: Company: Technology: Contact: Company: Technology: Contact: Hollming Works Oy Mechanical engineering, fabrication of heavy stainless steel structures Puunaulakatu 3, P.O.Box 96, FIN Pori, Finland Tel ; fax Jari Mattila Hytar Oy Remote handling, water hydraulics Hytar Oy, Ilmailukatu 13, P.O. Box 534, FIN Tampere, Finland Tel ; fax Olli Pohls [email protected] Instrumentti-Mattila Oy Designs and manufacturing of vacuum technology devices Valpperintie 263, FIN Nousiainen, Finland Tel , Fax Veikko Mattila [email protected] Japrotek Oy Designs and manufacturing of stainless steel process equipment such as columns, reactors and heat exchangers Japrotek Oy, P.O.Box 12, FIN-68601, Pietarsaari, Finland Tel , Fax Ulf Sarelin [email protected] Jutron Oy Versatile electronics manufacturing services Jutron Oy, Konekuja 2, FIN Oulu, Finland Tel , Fax Keijo Meriläinen [email protected] Kankaanpää Works Oy Mechanical engineering, fabrication of heavy stainless steel structures including 3D cold forming of stainless steel Kankaanpää Works Oy, P.O.Box 56, FIN Kankaanpää, Finland Tel ; fax Jarmo Huttunen [email protected] Kempower Oy Designs and manufacturing of standard and customised power sources for industrial and scientific use Hennalankatu 39, P.O.Box 13, FIN-15801, Lahti, Finland Tel , Fax Petri Korhonen [email protected] Luvata Oy (former Outokumpu Poricopper) Superconducting strands and copper products. Luvata Oy, Kuparitie, P.O Box 60, FIN Pori, Finland Tel ; fax Ben Karlemo [email protected] 137

147 Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Mansner Oy Precision Mechanics Precision mechanics: milling, turning, welding, and assembling. From stainless steels to copper. Mansner Oy, Yrittäjäntie 73, FIN Karkkila, Finland Tel ; Fax Sami Mansner [email protected] Marioff Corporation Oy Mist fire protection systems Marioff Corporation Oy, P.O.Box 25, FIN Vantaa, Finland Tel ; Fax Pekka Saari [email protected] Metso Powdermet Oy Special stainless steels, powder metallurgy, component technology/ engineering, design, production and installation Metso Powdermet Oy, P.O.Box 1100, FIN Tampere, Finland Tel ; fax Jari Liimatainen [email protected] Oxford Instruments Analytical Plasma diagnostics, vacuum windows Nihtisillankuja, P.O. Box 85, FIN Espoo, Finland Tel: , Fax: Heikki Sipilä [email protected] Patria Oyj Defence and space electronics hardware and engineering Patria Oyj, Kaivokatu 10, FIN Helsinki, Finland Tel , Fax Tapani Nippala [email protected] PI-Rauma Oy Computer aided engineering with CATIA. PI-Rauma Oy, Mäntyluoto, FIN Pori, Finland Tel ; fax Matti Mattila [email protected] Platom Oy Remote handling, thermal cutting tools and radioactive waste handling. Platom Oy, Graanintie 5, P.O.Box 300, FIN Mikkeli, Finland Tel ; Fax Miika Puukko [email protected] Company: PPF Products Oy Service: Industry activation and support Contact. Portaantie 548, FIN Porras, Finland rapdefault Tel , Pertti Pale [email protected] 138

148 Company: widctlparrole: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Prizztech Oy Industry activation and support Teknologiakeskus Pripoli, Tiedepuisto 4, FIN Pori, Finland Tel ; fax Leena Annila [email protected] Rados Technology Oy Dosimetry, waste & contamination monitoring and environmental monitoring. Rados Technology Oy, P.O.Box 506, FIN Turku, Finland Tel ; Fax Erik Lehtonen [email protected] Rejlers Oy System and subsystem level design, FE modelling and analysis with ANSYS, studies and technical documentation, installation and maintenance instructions, 3D modelling and visualisation of machines and components. Rejlers Oy, Myllykatu 3, FIN Hyvinkää, Finland Tel: ; Fax Jouni Vidqvist [email protected] Rocla Oyj Heavy Automated guided vehicles Rocla Oyj, P.O.Box 88, FIN Järvenpää, Finland Tel , Fax Pekka Joensuu [email protected] Selmic Oy Microelectronics design and manufacturing, packaging technologies and contract manufacturing services. Selmic Oy, Vanha Porvoontie 229, FIN Vantaa, Finland Tel: ; Fax Patrick Sederholm [email protected] Solving Oy Heavy automated guided vehicles. Equipment for heavy assembly and material handling based on air film technology for weights up to hundreds of tons. Solving Oy, P.O.Box 98, FIN Pietarsaari, Finland Tel ; Fax Bo-Goran Eriksson [email protected] Sweco PIC Oy Consulting and engineering company operating world-wide, providing consulting, design, engineering and project management services for industrial customers in plant investments, product development and production. Liesikuja 5, P.O.Box 31, FIN Vantaa, Finland Tel Kari Harsunen [email protected] 139

149 Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Company: Technology: Contact: Tampereen Keskustekniikka Oy Product development, design, production, marketing, and sales of switchgear and controlgear assemblies. Hyllilänkatu 15, P.O.Box 11, FIN Tampere, Finland Tel Reijo Anttila Tankki Oy Production and engineering of stainless steel tanks and vessels for use in different types of industrial installations Oikotie 2, FIN Ähtäri, Finland Tel , Fax Jukka Lehto [email protected] TVO Nuclear Services Oy Nuclear power technologies; service, maintenance, radiation protection and safety. TVO Nuclear Services Oy, FIN Olkiluoto Tel ; Fax Antti Piirto [email protected] TP-Konepaja Oy / Arelmek Oy Heavy welded and machined products, DTP2 structure TP-Konepajat Oy / Arelmek Oy, PL 23, FIN Tampere, Finland Tel Jorma Turkki [email protected] Woikoski Oy Production, development, applications and distribution of gases and liquid helium. Voikoski, P.O.Box 1, FIN Vuohijärvi, Finland Tel Fax Kalevi Korjala [email protected] 140

150 ANNEX C Doctoral and Graduate Theses 1. Karin Rantamäki, Particle-in-cell simulations of the near-field of a lower hybrid grill, VTT Publications 493, Espoo 2003 (Doctorate Thesis at Helsinki University of Technology). 2. Markus Airila, Chaos in high-power high-frequency gyrotrons, Helsinki University of Technology Publications in Engineering Physics TKK-F-A825, Espoo 2004 (Doctorate Thesis at Helsinki University of Technology). 3. Tommi Jokinen, Novel ways of using Nd:YAG laser for welding thick section austenitic stainless steel, VTT Publications 522, Espoo 2004, 120 p. + app. 12 p. (Doctorate Thesis at Lappeenranta University of Technology). 4. Pekka Moilanen, Pneumatic servo-controlled material testing device capable of operating at high temperature water and irradiation conditions, VTT Publications 532, VTT Industrial Systems, Espoo 2004, 154 p. (Doctorate Thesis at Helsinki University of Technology). 5. Albert Raneda, Impedance control of a water hydraulic manipulator for teleoperation applicatios, Doctorate Thesis at Tampere University of Technology, Tampere, TUT, 2004, 144 p. 6. Samuli Saarelma, Magnetohydrodynamic stability analyses of tokamak edge plasmas, Helsinki University of Technology Publications in Engineering Physics TKK-F-A836, Espoo 2005 (Doctorate Thesis at Helsinki University of Technology). 7. Krister O. E. Henriksson, Erosion and modification of metal surfaces by light and heavy ions, Doctorate Thesis at University of Helsinki, University of Helsinki, Helsinki University Printing House, ISBN Helsinki Petra Träskelin, Sticking and erosion at carbon-containing plasma-facing materials in fusion reactors, Doctorate Thesis at University of Helsinki, University of Helsinki, Helsinki Petri Kemppainen, Effect of specimen size and geometry on fracture resistance behaviour of metals, Master s Thesis, Helsinki University of Technology, 2003, 90 p. + appendices. 10. Ville Hynönen, Chaos in edge localized modes in tokamaks: detection of unstable periodic orbits, Master s Thesis, Helsinki University of Technology Miikka Karhu, Paksun austeniittisen ruostumattoman teräksen kapearailohybridihitsaus, Master s Thesis, Lappeenranta University of Technology, 2003, 116 pp. (in Finnish) 12. Salomon Janhunen, Gyrokinetic field-particle code for plasma turbulence studies, Master s Thesis, Helsinki University of Technology, Kimmo Kallio, Fuusioreaktorin huoltolaitteen lujuusopillinen tarkastelu, Master s Thesis, Tampere University of Technology, (in Finnish) 14. Niklas Juslin, Bond order potentials for nonequilibrium phenomena, Master s Thesis, University of Helsinki, Samuli Verho, Patruunatekniikan soveltaminen vesihydrauliikkaan, Master s Thesis, Tampere University of Technology, 2005 (in Finnish). 16. Hannu Koivisto, Design of a 3 DOF water hydraulic arm for teleoperation, Master s Thesis, Tampere University of Technology, 2005, 120 p. 17. Pia Kåll, 3D simulations of impurity transport in the divertor region of ASDEX Upgrade tokamak, Master s Thesis, Helsinki University of Technology, Svante Henriksson, Plasmaturbulenssin karakterisointi tilastollisten suureiden avulla FT-2 -tokamakin gyrokineettisissä simulaatioissa, Master s Thesis, University of Turku, 2005 (in Finnish). 19. O. Lehtinen, Kineettinen Monte Carlo simulaatio vedyn diffuusiosta BCC-metalleissa, Master s Thesis at University of Helsinki, 2006 (in Finnish). 20. Carolina Björkas, Simulering av strålningsinducerad defektklusterformation i GaAS, GaAs0.9N0.1 och Fe0.9Cr0.1, Master s Thesis, University of Helsinki, 2006 (in Swedish). 141

151 21. Matti Kortelainen, Modelling and diagnostics of an incineration device based on low-pressure non-equilibrium plasma, Master s Thesis, Helsinki University of Technology, Jussi Hopia, Control hardware and condition monitoring system for intersector Weld/Cut robot, Master s Thesis, Lappeenranta University of Technology, Otto Asunta, Fusion alpha performance in advanced scenario plasmas, Master s Thesis, Helsinki University of Technology, Markus Nora, Modelling of perturbative transport experiments in JET, Master s Thesis, Helsinki University of Technology, Antti Mäkelä, Vesihydraulisten vääntötoimilaitteiden kehitys robottiniveleen, Master s Thesis, Tampere University of Technology, (in Finnish) 26. Harri Mäkinen, Fuusioreaktorin huoltorobotin testausjärjestelmän suunnittelu, Master s Thesis, Tampere University of Technology, (in Finnish) 27. Olli Linna, Käyttöliittymän kehitys teleoperoidun vesihydraulisen manipulaattorin kehitykseen, Master s Thesis, Tampere University of Technology, (in Finnish). 142

152 ANNEX D Publication and Reports Fusion Physics and Plasma Engineering 1.1 Publications in Scientific Journals Fusion Plasma Physics 1. O. Dumbrajs, A Novel Method of Improving Performance of Coaxial Gyrotron Resonators, IEEE Transactions on Plasma Science 30 (2002) M.I. Airila and O. Dumbrajs, Spatio-Temporal Chaos in the Transverse Section of Gyrotron Resonators, IEEE Transactions on Plasma Science 30 (2002) J.A. Heikkinen and K.M. Rantamäki, Plasma Coupling and Near Field of a Modular Recessed ICRF Antenna, IEEE Transactions on Plasma Science 30 (2002) O. Dumbrajs and D. Teychenné, Electron trajectories in gyrotron resonators with realistic RF field profiles. Hamiltonian approach, Journal of Communications Technology and Electronics 47 (2002) S. Lashkul, et al., (incl. T. Kurki-Suonio, and J. Heikkinen), Formation of Transport Barriers in Lower Hybrid Experiment at FT-2, Journal of Plasma and Fusion Research 4 (2002) S. Saarelma, S. Günter, L.D. Horton, and the ASDEX Upgrade Team, MHD stability analysis of type II ELMs in ASDEX Upgrade, Nuclear Fusion 43 (2003) M.I. Airila and O. Dumbrajs, Stochastic processes in gyrotrons, Nuclear Fusion 43 (2003) M.I. Airila, O. Dumbrajs, P. Kåll, and B. Piosczyk, Influence of reflections on the operation of the 2 MW, CW 170 GHz coaxial cavity gyrotron for ITER, Nuclear Fusion 43 (2003) M.J. Mantsinen, et al., Application of ICRF waves in tokamaks beyond heating, Plasma Physics and Controlled Fusion 45 (2003) A445-A J.S. Lönnroth, et al., (incl. S. Saarelma), Integrated predictive modeling of the effect of neutral gas puffing in ELMy H-mode plasmas, Plasma Physics and Controlled Fusion 45 (2003) M.I. Airila, Degradation of operation mode purity in a gyrotron with an off-axis electron beam, Physics of Plasmas 10 (2003) O. Dumbrajs, T. Idehara, Y. Iwata, S. Mitsudo, I. Ogawa, and B. Piosczyk, Hysteresis-like effects in gyrotron oscillators, Physics of Plasmas 10 (2003) T.P. Kiviniemi, S.K. Sipilä, V.A. Rozhansky, S.P. Voskoboynikov, E.G. Kaveeva, J.A. Heikkinen, D.P. Coster, R. Schneider and X. Bonnin, Neoclassical nature of radial electric field at L-H transition, Physics of Plasmas 10 (2003) S. Günter, et al., (incl. S. Saarelma), Neoclassical tearing modes on ASDEX Upgrade: improved scaling laws, high confinement at high βn and new stabilization experiments, Nuclear Fusion 43 (2003) J.-M. Noterdaeme, et al., (incl. J. Heikkinen, M. Mantsinen, Heating, current drive and energetic particle studies on JET in preparation of ITER operation, Nuclear Fusion 43 (2003) J.-M. Noterdaeme et al., (incl., M. Mantsinen), Spatially Resolved Toroidal Plasma Rotation with ICRF on JET, Nuclear Fusion 43 (2003) X. Litaudon et al., (incl. M.J. Mantsinen, T.J.J. Tala), G. Tresset, O. Tudisco, L. Zabeo, K.-D. Zastrow and JET-EFDA contributors Progress Towards Steady-State Operation and Real-Time Control of Internal Transport Barriers in JET, Nuclear Fusion 43 (2003) D. Moreau, et al., (incl. T. Tala), Real-time control of the q-profile in JET for steady state advanced tokamak operation, Nuclear Fusion 43 (2003) X. Garbet, et al., (incl. T. Tala), Micro-stability and Transport Modelling of Internal Transport Barriers on JET, Nuclear Fusion 43 (2003) P.T. Lang, et al., (incl. S. Saarelma), ELM frequency control by continuous small pellet injection in ASDEX Upgrade, Nuclear Fusion 43 (2003)

153 21. M.F.F. Nave et al., (incl., M.J. Mantsinen), Role of sawtooth in avoiding impurity accumulation and maintaining good confinement in JET radiative mantle discharges, Nuclear Fusion 43 (2003) H. Zohm, et al., (incl. T. Kurki-Suonio and S. Saarelma), Overview of ASDEX Upgrade results, Nuclear Fusion 43 (2003) W. Suttrop, et al., (incl. T. Kurki-Suonio), ELMfree stationary H-mode plasmas in ASDEX Upgrade tokamak, Plasma Physics and Controlled Fusion 45 (2003) R.C. Wolf, et al., (incl. M. Mantsinen), Characterisation of ion heat conduction in JET and ASDEX Upgrade plasmas with and without internal transport barriers, Plasma Physics and Controlled Fusion 45 (2003) B. Piosczyk, et al., (incl. O. Dumbrajs), Towards a 2 MW, CW, 170 GHz coaxial cavity gyrotron for ITER, Fusion Engineering and Design (2003) Ph. Bibet et al., (incl. K. Rantamäki, F. Wasastjerna), Overview of the ITER-FEAT LH system, Fusion Engineering and Design (2003) W. Fundamenski et al., (incl. S. Sipilä, T. Kiviniemi, T. Kurki-Suonio), Narrow power profiles seen at JET and their relation to ion orbit losses, Journal of Nuclear Materials (2003) D.P. Coster et al. (incl. S. Sipilä), An overview of JET edge modelling activities, Journal of Nuclear Materials (2003) G.F. Matthews et al., (incl. T. Kurki-Suonio, S. Sipilä), The effect of ion orbit losses on JET edge plasma simulations, Journal of Nuclear Materials (2003) B. Piosczyk et al., (incl. O. Dumbrajs), Coaxial cavity gyrotron recent experimental results, IEEE Transactions on Plasma Science 30 (2002) V.G. Kiptily et al., (incl. M.J. Mantsinen), Gamma-diagnostics of alpha particles in 4 He and D-T plasmas, Review of Scientific Instruments 43 (2003) V. Parail et al., (incl. J. Lönnroth), Integrated predictive modelling of JET H-mode plasmas with type-i and type-iii ELMs, Plasma Physics Reports 7 (2003) M. Bécoulet, et al., (incl J. Lonnroth, S. Saarelma, Edge localized mode physics and operational aspects in tokamaks, Plasma. Physics and Control. Fusion 45 (2003) A93-A E. Joffrin et al., (incl. M. Mantsinen, T. Tala), Integrated scenarios in JET using real time profile control, Plasma Physics and Controlled Fusion 45 (2003) A367-A G. Zemulis, A. Adomavicius, and R. Salomaa, Loss of flow accident analysis of a water-cooled fusion reactor, pp in Advances in Heat Transfer Engineering (Eds. B. Sunden, J. Vilemas), Begell House Inc M.J. Mantsinen et al., Localized Bulk Electron Heating with ICRF Mode Conversion in the JET Tokamak, Nuclear Fusion 44 (2004) J.-S. Lönnroth et. al., Predictive modelling of ELMy H-modes with a new theory-motivated model for ELMs, Plasma Physics and Controlled Fusion 46 (2004) A J.-S. Lönnroth et al., Predictive transport modelling and MHD stability analysis of mixed type I-II ELMy H-mode JET plasmas, Plasma Physics and Controlled Fusion 46 (2004) J.-S. Lönnroth et al., Predictive transport modelling of type I ELMy H-mode dynamics using a theory-motivated combined ballooning-peeling model, Plasma Physics and Controlled Fusion 46 (2004) S. Saarelma and S. Günter, Edge stability analysis of high âp plasmas,plasma Physics and Controlled Fusion 46 (2004) V. Hynönen, O. Dumbrajs, A.W. Degeling, T. Kurki-Suonio, and H. Urano, The search for chaotic edge localized modes in ASDEX Upgrade, Plasma Physics and Controlled Fusion 46 (2004) O. Dumbrajs, G.S. Nusinovich, and B. Piosczyk, Reflections in gyrotrons with radial output: consequences for the ITER coaxial gyrotron, Physics of Plasmas 11 (2004) G. Zvejnieks et al., (incl. O. Dumbrajs), Autoregressive moving average model for analyzing edge localized mode time-series on axially symmetric divertor experiment upgrade tokamak, Physics of Plasmas 11 (2004) J.A. Heikkinen, S. Janhunen, T.P. Kiviniemi and P. Kåll, Full-f particle simulation method for solution of transient edge phenomena, Contributions to Plasma Physics 44 (2004) O. Dumbrajs and G.I. Zaginaylov, Ohmic losses in coaxial gyrotron cavities with corrugated insert, IEEE Transactions on Plasma Science 32 (2004)

154 46. O. Dumbrajs, et al., Reflections in gyrotrons with axial output, IEEE Transactions on Plasma Science 32 (2004) O. Dumbrajs and G.S. Nusinovich, Coaxial gyrotrons: past, present, and future, invited review article, IEEE Transactions on Plasma Science 32 (2004) M.I. Airila and P. Kåll, Effect of reflections on nonstationary gyrotron oscillations, IEEE Transactions on Microwave Theory and Techniques 52 (2004) M.I. Airila and O. Dumbrajs, Correction to: Spatio-temporal chaos in the transverse section of gyrotron resonators, IEEE Transactions on Plasma Science 32 (2004) O. Dumbrajs, H. Kalis, and A. Reinfelds, Numerical solution of single mode gyrotron equation, Mathematical Modelling and Analysis 9 (2004) L.-G. Eriksson et al., (incl. M.J. Mantsinen), Plasma Rotation induced by directed waves in the ion cyclotron range of frequencies, Physical Review Letters 92 (2004) L.-G. Eriksson et al., (incl. M.J. Mantsine), Destabilisation of Fast Ion Induced Long Sawteeth by Localised Current Drive in the JET Tokamak, Physical Review Letters 92 (2004) W. Fundamenski, S. Sipilä, et al., Boundary plasma energy transport in JET ELMy H- modes. Nuclear Fusion 44 (2004) J. Ongena et al., (incl. M.J. Mantsinen), Towards the realization on JET of an integrated H-mode scenario for ITER, Nuclear Fusion 44 (2004) T. Hellsten et al., (incl. M.J. Mantsinen), Effects of finite drift orbit width and RF-induced spatial transport on plasma heated by ICRH, Nuclear Fusion 44 (2004) X. Litaudon et al., (incl T. Tala), Status of and prospects for Advanced Tokamak regimes from multi-machine comparisons using the International Tokamak Physics Activity database, Plasma Physics and Controlled Fusion 46 (2004) A19-A W. Suttrop et al., (incl. T. Kurki-Suonio, Study of quiescent H-mode plasmas in ASDEX Upgrade, Plasma Physics and Controlled Fusion 46 (2004) A G. Saibene et al., (incl. J. Lönnroth), Dimensionless pedestal identity experiments in JT-60U and JET in ELMy H-mode plasmas, Plasma Physics and Controlled Fusion 46 (2004) A X. Garbet et al., (incl. T. Tala), Physics of Transport in Tokamaks, Plasma Physics and Controlled Fusion 46 (2004) B557-B Yu.F. Baranov et al., (incl. M.J. Mantsinen), On the link between q-profile and Internal Transport Barriers, Plasma Physics and Controlled Fusion 46 (2004) T. Onjun et al., (incl J. Lönnroth), Stability analysis of H-mode pedestal and edge localized modes in a Joint European Torus power scan, Physics of Plasmas 11 (2004) M-L. Mayoral et al., (incl. M.J. Mantsinen), Studies of burning plasma physics in the Joint European Torus, Physics of Plasmas 11 (2004) A. Loarte et al., (incl. J. Lönnroth), Characterization of pedestal parameters and edge localized mode energy losses in the Joint European Torus and predictions for the International Thermonuclear Experimental Reactor, Physics of Plasmas 11 (2004) T. Onjun et al., (incl. J. Lönnroth), Integrated pedestal and core modeling of Joint European Torus (JET) triangularity scan discharges, Physics of Plasmas 11 (2004) B. Piosczyk et al., (incl. O. Dumbrajs), M.V. Kartikeyan, M.K. Thumm, and X. Yang, 165-GHz coaxial cavity gyrotron, IEEE Transactions on Plasma Science 32 (2004) La Agusu, T. Idehara, and O. Dumbrajs, Mode selection for a terahertz gyrotron based on a pulse magnet system, International Journal of Infrared and Millimeter Waves 25 (2004) T. Tala, et al., (incl. A. Salmi), Predictive transport simulations of real-time profile control in JET advanced tokamak plasmas, Nuclear Fusion 45 (2005) S. Saarelma et al., (incl. J. Lönnroth), MHD stability analysis of diagnostic optimized configuration shots in JET, Plasma Physics and Controlled Fusion 47 (2005) K.M. Rantamäki et al., (incl. S.J. Karttunen, M. Mantsinen), Bright Spots Generated by Lower Hybrid Waves on JET, Plasma Physics and Controlled Fusion 47 (2005) M J Mantsinen et al., (incl. A Salmi), Fast ion distributions driven by polychromatic ICRF waves on JET, Plasma Physics and Controlled Fusion 47 (2005) O. Dumbrajs and G.S. Nusinovich, Azimuthal instability of radiation in gyrotrons with overmoded resonators, Physics of Plasmas 12 (2005)

155 72. M.I. Airila, O. Dumbrajs, et al., Sightline optimization of the multichannel laser interferometer for W7-X, Review of Scientific Instruments 76 (2005) T. Ekholm, S. Janhunen, J.A. Heikkinen, S. Henriksson and T.P. Kiviniemi, Characteristics of Transport Barrier Generation from Gyrokinetic Simulation in a Tokamak, IEEE Transactions on Plasma Science 33 (2005) T.M.J. Ikonen and O. Dumbrajs, Search for deterministic chaos in ELM time series of ASDEX Upgrade tokamak, IEEE Transactions on Plasma Science 33 (2005) C. Bourdelle et al., (incl. T. Tala), Impact of the á parameter on the microstability of internal transport barriers, Nuclear Fusion 45 (2005) G. Saibene et al., (incl. J.S. Lönnroth), Characterization of small ELM experiments in highly shaped single null and quasi-double-null plasmas in JET, Nuclear Fusion 45 (2005) A. Ekedahl et al., (incl. M. Mantsinen, K. Rantamäki), Long Distance Coupling of Lower Hybrid Waves in JET Plasmas with Edge and Core Transport Barriers, Nuclear Fusion 45 (2005) J.E. Kinsey et al., (incl., T. Tala), Transport Modeling and Gyrokinetic Analysis of Advanced High Performance Discharges, Nuclear Fusion 45 (2005) E. Joffrin et al., (incl., T. Tala), The hybrid scenario in JET: towards its validation for ITER, Nuclear Fusion 45 (2005) D. Stork, et al., (incl. M. Mantsinen), Overview of transport, fast particle and heating and current drive physics using tritium in JET plasmas, Nucl. Fusion 45 (2005) S181-S T. Hellsten et al., (incl M. Mantsinen, T. Tala), On the Parasitic Absorption in FWCD Experiments in JET ITB Plasmas, Nuclear Fusion 45 (2005) W. Suttrop, V. Hynönen, T. Kurki-Suonio, et al., Studies of the Quiescent H-mode regime in ASDEX Upgrade and JET, Nuclear Fusion 45 (2005) W. Fundamenski, R.A. Pitts, G.F. Matthews, V. Riccardo, S. Sipilä, ELM-averaged Power Exhaust on JET, Nuclear Fusion 45 (2005) V.G. Kiptily, (incl. M. Mantsinen), Gamma-ray imaging of D and 4 He ions accelerated by ion-cyclotron-resonance heating in JET plasmas, Nucl. Fusion 45 (2005) L21-L S.E. Sharapov, (incl. M. Mantsinen), Experimental studies of instabilities and confinement of energetic particles on JET and MAST, Nucl. Fusion 45 (2005) S D Pinches et al., (incl. M J Mantsinen), The Role of Energetic Particles in Fusion Plasmas, Plasma Physics and Controlled Fusion 46 (2004) B187-B L. Laborde et al., (incl. T. Tala) A model-based technique for integrated real-time profile control in the JET tokamak, Plasma Physics and Controlled Fusion 47 (2005) F. Imbeaux et al., (incl. T.J.J. Tala), Multi-machine Transport Analysis of Hybrid discharges from ITPA Profile Database, Plasma Physics and Controlled Fusion 47 (2005) B179-B A.A. Tuccillo et al., (incl., S. Karttunen, K. Rantamaki), Progress in LHCD: a tool for advanced regimes on ITER, Plasma Physics and Controlled Fusion 47 (2005) B363-B J. Weiland et al., (incl. T. Tala), Effects of temperature ratio on JET transport in hot ion and hot electron regimes, Plasma Physics and Controlled Fusion 47 (2005) Y. Kominis, O. Dumbrajs, et al., Chaotic electron dynamics in gyrotron resonators, Physics of Plasmas 12 (2005) Y. Kominis, O. Dumbrajs, et al., Canonical Perturbation Theory for Complex Electron Dynamics in Gyrotron Resonators, Physics of Plasmas 12 (2005) O. Dumbrajs, et al., Stochastization as a possible cause of fast reconnection in the frequently interrupted regime of neoclassical tearing modes, Physics of Plasmas 12 (2005) F. Nabais, D. Borba, M. Mantsinen, M. F. F. Nave, S. E. Sharapov, Fishbones in Joint European Torus plasmas with high ion-cyclotron-resonance-heated fast ions energy content, Physics of Plasmas 12 (2005) W. Fundamenski et al., (incl. S. Sipilä), Effect of B x gradb direction on SOL energy transport in JET, Journal of Nuclear Materials (2005) A. Lyssoivan, et al., (incl. M. Mantsinen, M. Santala), Development of ICRF wall conditioning technique on divertor-type tokamaks ASDEX Upgrade and JET, Journal of Nuclear Materials (2005)

156 97. Ph. Bibet et al., (incl. K. Rantamäki), Toward a LHCD system for ITER, Fusion Engineering and Design 74 (2005) P. Velarde, F. Ogando, et al., Fast ignition heavy-ion fusion by jet impact, Nuclear Instruments and Methods in Physics Research A 544/1-2 (2005) P. Velarde, F. Ogando, et al., Comparison between jet collision and shell impact concepts for fast ignition, Laser and Particle Beams 23/1 (2005) P. Mantica et al., (incl. M. Mantsinen, A. Salmi), Probing Internal Transport Barriers with Heat Pulses in JET, Physical Review Letters 96 (2006) M. Murakami, H. Nagatomo, H. Azechi, F. Ogando, M. Perlado and S. Eliezer, Innovative ignition scheme for ICF - impact fast ignition, Nuclear Fusion 46 (2006) A.A. Tuccillo et al., (incl. M.J. Mantsinen), Development on JET of advanced tokamak operations for ITER, Nuclear Fusion 46 (2006) P.U. Lamalle, M.J. Mantsinen, et al., (incl. J. Heikkinen, A. Salmi, M.. Santala, T. Tala), Expanding the operating space of ICRF on JET with a view to ITER, Nuclear Fusion 46 (2006) T.J.J. Tala et al., Fully Predictive Transport Simulations of ITB Plasmas in JET, JT-60U and DIII-D, Nuclear Fusion 46 (2006) M.-L. Mayoral, et al., (incl. M.J. Mantsinen, M. Santala), Hydrogen plasmas with ICRF inverted minority and mode conversion heating regimes in the JET tokamak, Nuclear Fusion 46 (2006) S550-S V. Igochine, O. Dumbrajs, et al., Stochastization as a possible cause for fast reconnection during MHD mode activity in the ASDEX Upgrade tokamak, Nuclear Fusion 46 (2006) L.-G. Eriksson, et al., (incl. M. Mantsinen, M. Santala), On ion cyclotron current drive for sawtooth control, Nuclear Fusion 46 (2006) S J.P. Graves, et al., (incl. M. Mantsinen), Sawtooth control in fusion plasmas, Plasma Physics and Controlled Fusion 47 (2005) B T.P. Kiviniemi, J.A. Heikkinen, S. Janhunen and S.V. Henriksson, Full f gyrokinetic simulation of FT-2 tokamak plasma, Plasma Physics and Controlled Fusion 48 (2006) A327-A A.Salmi, M.J. Mantsinen, et al., JET experiments to assess the clamping of the fast ion energy distribution during ICRF heating due to finite Larmor radius effects, Plasma Physics and Controlled Fusion 48 (2006) M.I.K. Santala, M.J. Mantsinen, et al., Proton-triton nuclear reaction in ICRF heated plasmas in JET, Plasma Physics and Controlled Fusion 48 (2006) T. Kurki-Suonio, V. Tulkki, S. Sipilä and R. Salomaa, Fusion alpha performance in advanced scenario plasmas based on reversed central magnetic shear, Plasma Physics and Controlled Fusion 48 (2006) P.C. de Vries, K.M. Rantamaki, et al., (incl. T. Tala), Plasma Rotation and Momentum Transport studies at JET, Physics and Controlled Fusion 48 (2006) S.V. Henriksson, S.J. Janhunen, T.P. Kiviniemi and J.A. Heikkinen, Global Spectral Investigation of Plasma Turbulence in Gyrokinetic Simulations, Physics of Plasmas 13 (2006) J.A. Heikkinen, S. Henriksson, S. Janhunen, T.P. Kiviniemi and F. Ogando, Gyrokinetic Simulation of Particle and Heat Transport in the Presence of Wide Orbits and Strong Profile Variations in the Edge Plasma, Contributions to Plasma Physics 46 (2006) J. Lönnroth, et al., (incl. T. Kiviniemi), Integrated ELM modelling, Contributions to Plasma Physics 46 (2006) T. Tala and X. Garbet, Physics of Internal Transport Barriers, Compte-Rendu de l Académie des Sciences 7 (2006) G. Dammertz et al., (incl. O. Dumbrajs), Highpower gyrotron development at Forschungszentrum Karlsruhe for fusion applications, IEEE Transactions on Plasma Science 34 (2006) O. Dumbrajs, et al., Hamiltonian map description of electron dynamics in gyrotrons, IEEE Transactions on Plasma Science 34 (2006) Z.C. Ioannidis, O. Dumbrajs and I.G. Tigelis, Eigenvalues and Ohmic Losses in Coaxial Gyrotron Cavity, IEEE Transactions on Plasma Science 34 (2006) V. Igochine, O. Dumbrajs, H. Zohm, A. Flaws, Stochastic sawtooth reconnection in ASDEX Upgrade, Nuclear Fusion 47 (2007) G. Dammertz, et al. (incl. O. Dumbrajs) Highpower gyrotron development at Forschungszentrum Karlsruhe for fusion applications. IEEE Transactions on Plasma Science 34 (2006)

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158 141. M.J. Mantsinen, M. Laxåback, A. Salmi, et al., Comparison of monochromatic and polychromatic ICRH on JET, in Radio-Frequency Power in Plasmas (Proc. 15th Topical Conference on Radio Frequency Power in Plasmas May, 2003, Moran, Wyoming, USA), AIP, New York, p T. Hellsten et al., (incl. M. Mantsinen), Fast Wave Current Drive in JET ITB-Plasmas, in Radio-Frequency Power in Plasmas (Proc. 15th Topical Conference on Radio Frequency Power in Plasmas May, 2003, Moran, Wyoming, USA), AIP, New York, p A. Ekedahl et al., (incl. K. Rantamäki, M.J. Mantsinen), Long Distance Coupling of Lower Hybrid Waves in ITER Relevant Edge Conditions in JET Reversed Shear Plasmas, in Radio-Frequency Power in Plasmas (Proc. 15th Topical Conference on Radio Frequency Power in Plasmas May, 2003, Moran, Wyoming, USA), AIP, New York, p D. Moreau et al., (incl. T. Tala), Real-Time Control of the Current Profile in JET, in Radio-Frequency Power in Plasmas (Proc. 15th Topical Conference on Radio Frequency Power in Plasmas May, 2003, Moran, Wyoming, USA), AIP, New York, p B. Piosczyk et al., (incl. O. Dumbrajs), A2MW, 170 GHz coaxial cavity gyrotron, International Conf. Plasma Science, Jeju, Korea, 2-5 June 2003, p G. Saibene et al., (incl. J.S. Lönnroth, S. Saarelma), Pedestal and ELM characterisation of highly shaped Single Null and Quasi Double Null plasmas in JET, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P J.A. Heikkinen, T.P. Kiviniemi, S.J. Janhunen, and P. Kåll, Implicite Full f Particle Model for Simulations of Trapped Electron Modes, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P Yu.F. Baranov et al., (incl. M. Mantsinen), On the link between q-profile and ITBs, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P S. Saarelma, et al. (incl. T. Kurki-Suonio), MHD Stability Analysis of ASDEX Upgrade H-mode Plasmas in Various ELMy and ELM-Free Regimes, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P Yong-Su Na et al., (incl. T.J.J. Tala), Modelling of the Current Profile Control with Neutral Beam Injection at ASDEX Upgrade and Comparison to JET, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P W. Suttrop et al., (incl. T. Kurki-Suonio), ELMfree stationary H-mode plasmas in ASDEX Upgrade, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P T.P. Kiviniemi, J.A. Heikkinen, D. McDonald, R. Sartori, S.K. Sipilä, Y. Andrew, L-H transition threshold temperature for helium discharges in JET, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P J.-S. Lönnroth et al., (incl. R.R.E. Salomaa), Predictive transport modelling with theory-based and semi-empirical models for different ELMy H-mode scenarios, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P K.M. Rantamäki, A.T. Salmi, S.J. Karttunen, Particle-in-Cell Simulations for Lower Hybrid Coupling near Cut-Off Density, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P K.M. Rantamäki et al., (incl. S.J. Karttunen), Hot Spots Generated by Lower Hybrid Waves on JET, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P G. Granucci et al., (incl. M. Mantsinen, K. Rantamäki), Recent Results of LHCD Coupling Experiments with Near Gas Injection in JET, 30th EPS Conference on Controlled Fusion and Plasma Physics, July 7 11, 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P

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160 174. S.K. Sipilä, and W. Fundamenski, ITER Divertor Target Load Simulations with Orbit-Following Monte Carlo Code ASCOT, 10th European Fusion Theory Conference, September 8-10, 2003, Helsinki, Finland, Book of Abstracts, Report TKK-A-823 (2003) paper P J.A. Heikkinen, S. Janhunen and T.P. Kiviniemi, Full-f particle simulation method for solution of zonal flows, 10th European Fusion Theory Conference, September 8-10, 2003, Helsinki, Finland, Book of Abstracts, Report TKK-A-823 (2003) paper P D. Mazon, et al. (incl. T. Tala), Real-time control of the current density profile in JET 9th IAEA Technical Meeting on H-mode Physics and Transport Barriers, September 2003, San Diego, USA X. Litaudon, et al. (incl T. Tala), Status of and prospects for Advanced Tokamak regimes from multi-machine comparisons 9th IAEA Technical Meeting on H-mode Physics and Transport Barriers, September 2003, San Diego, USA M.I. Airila, G. Dammertz, O. Dumbrajs, T. Idehara, P. Kåll, La Agusu, S. Mitsudo, and B. Piosczyk, Reflections in gyrotrons with radial and axial outputs. A comparison, 28th International Conference on Infrared and Millimeter Waves, September 29-October 2, 2003, Otsu, Japan 179. B. Piosczyk, H. Budig, G. Dammertz, O. Dumbrajs, S. Drumm, S. Illy, M. Thumm, Coaxial cavity gyrotron - recent results and ongoing development work, 28th International Conference on Infrared and Millimeter Waves, September 29-October 2, 2003, Otsu, Japan, pp D. Moreau, et al., (incl. T. Tala), Real-Time Control of the Current Density Profile in JET, abstract in APS DPP th Annual Meeting of the Division of Plasma Physics, Albuquerque New Mexico USA, October D. Moreau, et al., (incl. T. Tala), Real-time control of internal transport barriers in JET: Experiments and simulations, Joint Meeting of US-Japan JIFT Workshop on Theory-Based Modeling and Integrated Simulation of Burning Plasmas, and 21COE Workshop on Plasma Theory, Kyoto Japan, December 15-17, N.C Hawkes, et al., (incl. T. Tala), Current-Hole Studies at JET, Large Tokamak Workshop on the Physics of the Current Hole, Naka, 3-4 February S. Karttunen, European Fusion Programme and International ITER Project, (invited paper), XXXIV EESTI Füüsikapäevad, Annual meeting of the Estonian Physical Society,Tartu, Estonia February 2004 (2004) C. Schlatter, et al., (incl. M. Santala), Error Estimation and Parameter Dependence of the Calculation of the Fast Ion Distribution Function, Temperature and Density using Data from the KF1 High Energy NPA on JET, Proceedings of the 15th HTPD Conference (San Diego, California, USA, April 2004) 185. O. Dumbrajs and G.I. Zaginaylov, Ohmic losses in coaxial gyrotron cavities with corrugated insert, Vacuum Electronics and Displays 10th Triennial Conference, Garmisch Partenkirchen, Germany, 3-4 May B. Piosczyk, A. Arnold, H. Budig, G. Dammertz, O. Dumbrajs, S. Illy, J. Jin, G. Michel, T. Rzesnicki, M. Thumm, and D. Wagner, 2MW, CW, 170 GHz coaxial cavity gyrotron, Vacuum Electronics and Displays 10th Triennial Conference, Garmisch Partenkirchen, Germany, 3-4 May B. Piosczyk, et al. (incl. O. Dumbrajs), Development of advanced high power gyrotrons for ECH&CD applications in fusion plasmas, 13th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, Nizhny Novgorod, Russia, May M. Becoulet, et al., (incl. S Saarelma, J. Lönnroth), Edge Localized Modes Control: Experiment and Theory, Proceedings of the 16th PSI - Plasma Surface Interactions conference, (Portland, Maine, USA, May 2004) P. Mantica, T. Tala, F. Imbeaux, M. Nora, V. Parail, J. Weiland Predictive modelling of electron temperature modulation experiments in JET L- and H-mode plasmas, 31st EPS Conference on Controlled Fusion and Plasma Physics, London, 28 June - 2 July 2004, European Conference Abstract 28G (2004) P P. Mantica, et al. (incl. M.J. Mantsinen, A. Salmi), Power modulation experiments in JET ITB plasmas, 31st EPS Conference on Plasma Physics London, 28 June-2 July 2004, European Conference Abstracts 28G (2004) P B. Weyssow, et al. (incl. M.J. Mantsinen), RF Induced Impurity Transport in the Mode Conversion Regime in a H-D plasma at JET, 31st EPS Conference on Plasma Phys. London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) P

161 192. J. Weiland, et al. (incl. T. Tala), Effects of temperature ratio on JET transport in hot ion and hot electron regimes, 31st EPS Conference on Controlled Fusion and Plasma Physics, London, 28 June - 2 July 2004, European Conference Abstract 28G (2004) P M. F. F. Nave, et al., (incl. M.J. Mantsinen), Small sawtooth regimes in JET plasmas," 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) P S.J. Janhunen, T.P. Kiviniemi and J.A. Heikkinen, Validation of gyrokinetic particle code ELM- FIRE for tokamak edge transport analysis, 31st EPS Conference on Controlled Fusion and Plasma Physics, London UK, 28 June 2 July, 2004, Europhysics Conference Abstracts 28G (2004) P J.A. Heikkinen, S.J. Janhunen and T.P. Kiviniemi, Gyrokinetic simulation of neoclassical and turbulent transport, 31st EPS Conference on Controlled Fusion and Plasma Physics, London UK, 28 June 2 July, 2004, Europhysics Conference Abstracts 28G (2004) P J.A. Heikkinen, et al., (incl. K.M. Rantamäki, A. Salmi, M.J. Mantsinen), Experiments on ICRF Coupling with Different Phasings, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) P P. U. Lamalle, M. J. Mantsinen, et al., (incl. M. Santala), Investigation of low concentration tritium ICRF heating on JET, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) P A. Salmi, P. Beaumont, P. de Vries, L.-G. Eriksson, C. Gowers, P. Helander, M. Laxåback, M.J. Mantsinen, J.-M. Noterdaeme, D. Testa and EFDA JET contributors, JET Experiments to Assess Finite Larmor Radius Effects on Resonant Ion Energy Distribution during ICRF Heating, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) P D. Mazon, et al. (incl. T. Tala), Current profile and ITB control for the development of advanced steady state plasmas in JET: experiments and modelling, 31st EPS Conference on Controlled Fusion and Plasma Physics, London UK, 28 June 2 July, 2004, Europhysics Conference Abstracts 28G (2004) P J.-S. Lönnroth et al., Modelling of ELM heat pulse propagation with the integrated core-sol transport code COCONUT, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) 201. G. Saibene, et al., (incl. J.S. Lönnroth), Small ELM experiments in H-mode plasmas in JET, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) M. Mironov, et al., (incl. M. Santala), Tritium Transport Studies with JET ISEP NPA During the Trace Tritium Experimental Campaign, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) 203. M.I.K. Santala, M.J. Mantsinen, et al., (incl. A. Salmi), pt Fusion by RF-heated Protons in JET Trace Tritium Discharges, 31st EPS Conference on Plasma Physics London, 28 June - 2 July 2004, European Conference Abstracts 28G (2004) 204. A. Andreev, V.G. Bespalov, E.V. Ermolaeva, and R.R E. Salomaa, Compression of high-intensity laser pulse by inhomogeneous plasma, Proc. of Laser Optics 2003 Superintense Light Fields and Ultrafast Processes, St. Petersburg, Russia, 30 June - 4 July, 2003 (Eds. V. E. Yashin and A. A. Andreev), Proc. SPIE 5482, (2004) K.M. Rantamäki, A.T. Salmi and S.J. Karttunen, Particle-in-cell simulations of the near-field of a lower hybrid grill, Proceedings of the Joint Varenna-Lausanne International Workshop, Varenna, 30 August -3 September 2004, International School of Plasma Physics Piero Caldirola. Theory of Fusion Plasmas, J.W. Connor, O. Sauter and E. Sindoni (Eds.), (Societa Italiana di Fisica, Bologna 2004), p G. Zvejnieks, et al. (incl. O. Dumbrajs), Distinguishing the deterministic and noise components in the ASDEX Upgrade ELM time series, Proceedings of the Joint Varenna-Lausanne International Workshop, Varenna, 30 August -3 September 2004, International School of Plasma Physics Piero Caldirola. Theory of Fusion Plasmas, J.W. Connor, O. Sauter and E. Sindoni (Eds.), (Societa Italiana di Fisica, Bologna 2004) 207. X. Garbet, et al., (incl. T. Tala), Interplay between electron and ion heat channels, 10th EU-US Transport Task Force Workshop, Varenna, Italy, 6 9September

162 208. C. Bourdelle, X. Litaudon, C.M. Roach, T. Tala, for the ITPA Topical Group on Transport and ITB Physics, and the International ITB Database Working Group, Impact of á on the microstability of internal transport barriers, 10th EU-US Transport Task Force Workshop, Varenna, Italy, 6 9September O. Dumbrajs, Modeling of stochastic processes in gyrotrons, (invited plenary talk) 10 th International Conference on Mathematical Methods in Electromagnetic Theory, Dniepropetrovsk, Ukraine, September B. Piosczyk, et al., (incl. O. Dumbrajs), Experiments on a 170 GHz coaxial cavity gyrotron, 23rd Symposium on Fusion Technology, Fondazione Cini, Venice, Italy, September O. Dumbrajs, G.S. Nusinovich, and B. Piosczyk, Reflections in gyrotrons with radial output: consequences for the ITER coaxial gyrotron, (invited keynote) 29th Int. Conf. Infrared Millimeter Waves, Karlsruhe, Germany, 27 September 1 October B. Piosczyk, et al. (incl. O. Dumbrajs), Progress in the development of the 170 GHz coaxial cavity gyrotron, (invited keynote) 29th Int. Conf. Infrared Millimeter Waves, Karlsruhe, Germany, 27 September 1 October Y. Kominis, O. Dumbrajs, et al., Chaotic electron dynamics in gyrotron resonators, 29th Int. Conf. Infrared Millimeter Waves, Karlsruhe, Germany, 27 September 1 October M.Q. Tran, et al. (incl. O. Dumbrajs), Development of high power gyrotrons for fusion plasma applications in the EU, (invited plenary talk) 29th Int. Conf. Infrared Millimeter Waves, Karlsruhe, Germany, 27 September 1 October T. Tala, et al., (incl. J. Lönnroth, A. Salmi), Progress in Transport Modelling of Internal Transport Barrier and Hybrid Scenario Plasmas in JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004, TH/P T.P. Kiviniemi, J.A. Heikkinen, S. Janhunen, T. Kurki-Suonio, S.K. Sipilä, Particle simulation of plasma turbulence and neoclassical Er at tokamak plasma edge, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004, TH/P D.P. Coster, et al., (incl. T. Kiviniemi, J. Lönnroth, S. Sipilä), Integrated modelling of material migration and target plate power handling at JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November W. Fundamenski, et al., (incl S. Sipilä), Power Exhaust on JET: An Overview of Dedicated Experiments, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004, EX/2-4 Ra J. Mailloux, et al. (incl. K. Rantamäki), ITER relevant coupling of Lower Hybrid Waves in JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004 EX/P P. Mantica, et al., (incl. M.J. Mantsinen, M. Nora, T. Tala), Progress in understanding heat transport at JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004 EX/P D. Moreau, et al., (incl. T. Tala), Development of Integrated Real-Time Control of Internal Transport Barriers in Advanced Operation Scenarios on JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004 EX/P G. Saibene, et al., (incl. J. Lönnroth), Dimensionless identity experiments in JT-60U and JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November W. Suttrop, V. Hynönen, P. T. Lang, T. Kurki- Suonio, et al., Studies of the Quiescent H-mode regime in ASDEX Upgrade and JET, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November H. Weisen, et al., (incl. K. Rantamäki, T. Tala), Anomalous particle and impurity transport in JET and implications for ITER, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004 EX/P S.E. Sharapov, et al., (incl. M. Mantsinen), Experimental studies of instabilities and confinement of energetic particles on JET and on MAST, 20th IAEA Fusion Energy Conference, 1-6 November 2004, Vilamoura, Portugal, paper EX/5-2Ra D. Stork, et al., (incl. M. Mantsinen), Overview of Transport, Fast Particle and Heating and Current Drive Physics using Tritium in JET plasmas, 20th IAEA Fusion Energy Conference, 1-6 November 2004, Vilamoura, Portugal, paper OV/ A.A. Tuccillo, et al., (incl. M.J. Mantsinen), Development on JET of Advanced Tokamak operations for ITER, 20th IAEA Fusion Energy Conference, 1-6 November 2004, Vilamoura, Portugal, paper EX/

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164 242. S. Janhunen, T.P. Kiviniemi, and J.A. Heikkinen, Benchmarking of a Kinetic Plasma Turbulence Simulation Model, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) T.H. Ekholm, S. Janhunen, J.A. Heikkinen, S.V. Henriksson, and T.P. Kiviniemi, Visualization of Plasma Turbulence and Transport Barrier Generation, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) M.I. Airila, and O. Dumbrajs, Chaos in high-power high-frequency gyrotrons, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) M.I. Airila and A.T. Salmi, Tritium breeding ratio and energy amplification in an infinite lithium blanket, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) S.V. Henriksson, S. Janhunen, T.P. Kiviniemi, J.A. Heikkinen, Wavenumber Spectra of Simulated FT-2 Plasma Turbulence, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) J.A. Heikkinen, S. Janhunen, and T.P. Kiviniemi, Evidence of Transport Barrier Formation in FT-2 Tokamak, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) T. M. J. Ikonen, V. Hynönen, and H. Urano, Search for deterministic chaos in ELM time series of ASDEX Ugrade tokamak, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) V. Hynönen, T. Kurki-Suonio, and W. Suttrop, Simulations of fast particle distributions in quiescent H-mode of ASDEX Upgrade, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) A. Salmi, M.J. Mantsinen, et al. Fast proton energy distribution during second harmonic ICRF heating of fusion plasmas, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo 2005, 229 (2005) T. Hellsten, et al., (incl. M. Mantsinen, T. Tala), Fast Wave Current Drive in JET ITB-Plasma, in Radio-Frequency Power in Plasmas (Proc. 16th Topical Conference on Radio Frequency Power in Plasmas April, 2005, Park City, Utah, USA), AIP, New York, M.-L. Mayoral, et al., (incl. M.J. Mantsinen, M. Santala), ICRF Heating For The Non-Activated Phase Of ITER: From Inverted Minority To Mode Conversion Regime, in Radio-Frequency Power in Plasmas (Proc. 16th Topical Conference on Radio Frequency Power in Plasmas April, 2005, Park City, Utah, USA), AIP, New York, J.-M. Noterdaeme, M. Mantsinen, et al. (incl. K. Rantamaki, A. Salmi, M. Santala), Development of RF Tools and Scenarios for ITER on JET, in Radio-Frequency Power in Plasmas (Proc. 16th Topical Conference on Radio Frequency Power in Plasmas April, 2005, Park City, Utah, USA), AIP, New York, V. Parail, T. Johnson, T. Kiviniemi, J. Lonnroth, et al., Effect of Ripple-Induced Ion Thermal Transport on H-mode Performance, 32nd EPS Conference on Controlled Fusion and Plasma Physics, Tarragona, Spain, 27 June - 1 July 2005, Europhysics Conference Abstracts 29C (2005) O Loarte, et al., (incl. J.S. Lönnroth), Influence of toroidal field direction and plasma rotation on pedestal and ELM characteristics in JET ELMy H-modes, 32nd EPS Conference on Controlled Fusion and Plasma Physics, Tarragona, Spain, 27 June - 1 July 2005, Europhysics Conference Abstracts 29C (2005) P

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167 286. O. Asunta, M. Airila, D. Coster, T. Kurki-Suonio and P. Kåll, Simulation of carbon migration in ASDEX Upgrade tokamak, Proceedings of the XL Annual Conference of the Finnish Physical Society, Tampere, Finland,9-11March 2006, M. Kauranen, A. Laitinen, J. Nieminen and T. T. Rantala (eds), Tampere University of Technology, Institute of Physics, Report 2006:1 (2006) T.P. Kiviniemi, M.I. Airila, J.A. Heikkinen, J.S. Janhunen, P. Kåll, T. Kurki-Suonio, J. Lönnroth and S. Sipilä, Exciting Challenge of Developing a Numerical Tokamak, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) D. Moreau, M. Ariola, L. Laborde, D. Mazon, T. Tala and JET EFDA contributors, Identification of the dynamic plasma response for integrated profile control in advanced scenarios on JET, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) F. Ogando, A. Signell, J. Heikkinen, M. Aspnäs, S. Henriksson, S. Janhunen and T. Kiviniemi, Performance enhancement to ELMFIRE gyrokinetic code leading to new ranges of application, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) J. Juul Rasmussen, et al., (incl. J.S. Lönnroth), Turbulence spreading transport simulations of JET plasmas, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) J. Lönnroth, et al., Modelling of effects of ripple-induced thermal ion losses on JET and JT-60U H-mode plasmas, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) J.A. Heikkinen, S. Henriksson, S. Janhunen, T.Kiviniemi, M. Nora and F. Ogando, Full f gyrokinetic simulation of transport in tokamak plasmas, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) A. Cardinali, et al., (incl. M. Mantsinen, A. Salmi), Modeling and analysis of the ICRH heating experiments in JET ITB Regimes, 33rd EPS Conference on Plasma Phys. Rome, Italy, June, 2006, Europhysics Conference Abstracts 30I (2006) P P. Mantica, et al., (incl. T.Tala), Heat wave propagation in JET ITB plasmas, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) P.C. de Vries, K.M. Rantamaki, et al., (incl. T. Tala), Plasma Rotation and Momentum Transport studies at JET, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006) P S.J. Janhunen, J.A. Heikkinen, T.P. Kiviniemi and T. Kurki-Suonio, Full f gyrokinetic modelling of neoclassical radial electric field, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006 (2006), Europhysics Conference Abstracts 30I (2006) T. Johnson, A. Salmi, V. Hynönen, I. Jenkins, T. Kiviniemi, T. Kurki-Suonio, J. Lönnroth and V. Parail, Numerical Modelling of Ripple Induced Transport in JET with Enhanced Toroidal Field Ripple, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006), P T. Kurki-Suonio, O. Asunta, V. Tulkki, T. Tala, T. Kiviniemi, S. Sipilä, R. Salomaa and JET EFDA contributors, ASCOT Simulations of Alpha particle Heating in JET Plasmas with Normal and Reversed q-profiles, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006), O T. Kurki-Suonio, V. Hynönen, W. Suttrop, H.-U. Fahrbach, J. Stober and the ASDEX Upgrade Team, Edge Fast Ion Distribution benchmarking ASCOT against experimental NPA data on ASDEX Upgrade, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006), P V. Hynönen, T. Kurki-Suonio, W. Suttrop, K. Sugiyama and the ASDEX Upgrade Team, ASCOT simulations of fast particle effects on ASDEX Upgrade edge, 33rd European Physical Society Conference on Plasma Physics, Rome, Italy, June 19-23, 2006, Europhysics Conference Abstracts 30I (2006), P

168 301. F. Ogando, J.A. Heikkinen, S. Henriksson, S.J. Janhunen, T.P. Kiviniemi and S. Leerink, Advanced methods in global gyrokinetic full f particle simulation of tokamak transport, Proceedings of the Joint Varenna-Lausanne International Workshop, Varenna, August 28 September 1st, 2006, International School of Plasma Physics Piero Caldirola. Theory of Fusion Plasmas, (2006) F. Ogando, J.A. Heikkinen, S. Henriksson, S.J. Janhunen, T.P. Kiviniemi and S. Leerink, Advanced methods in global gyrokinetic full f particle simulation of tokamak transport, Proceedings of the Joint Varenna-Lausanne International Workshop, Varenna, August 28 September 1st, 2006, International School of Plasma Physics Piero Cardiola. Theory of Fusion Plasmas, (2006). AIP Conference Proceedings 871, 181 (2007) G.M.D. Hogeweij, P. Buratti, F. Imbeaux and K. Rantamäki, Predictive Modelling of Electrin Heated JET Hybrid Discharges, 11th EU-US Transport Task Force Workshop, Marseille, France, 4 7 September K. Crombé, et al., (incl. T. Tala), Poloidal rotation velocity in the region of internal transport barriers in JET advanced mode plasmas, 11th EU-US Transport Task Force Workshop, Marseille, France, 4 7 September P. Mantica, et al., (incl T.Tala), Fast core response to edge cooling in JET: experiments and modelling, 11th EU-US Transport Task Force Workshop, Marseille, France, 4 7September P.C. de Vries, K.M. Rantamaki, et al., (incl. T. Tala), Study of Momentum Transport in ELMy H-mode JET discharges, 11th EU-US Transport Task Force Workshop, Marseille, France, 4 7 September F. Ogando, J.A. Heikkinen, S.J. Janhunen, T.P. Kiviniemi, S. Leerink, Gyrokinetic Full f Modelling of Plasma Turbulence in Tokamaks, 21st IAEA Fusion Energy Conference, Chengdu, China, October 2006, paper TH/P P. Velarde, F. Ogando, et al., Recent results on fast ignition jet impact scheme, 21st IAEA Fusion Energy Conference, IF/P5-5, Chengdu, China, October 2006, paper IF/P J.M. Perlado, et al., (incl F. Ogando), Advances in Target Design and Materials Physics for IFE at DENIM, 21st IAEA Fusion Energy Conference, IF/P5-7, Chengdu, China, October 2006, paper IF/P D. Moreau, et al. (incl. T. Tala), New Dynamic-Model Approach for Simultaneous Control of Distributed Magnetic and Kinetic Parameters in the ITER-like JET Plasmas, 21st IAEA Fusion Energy Conference, Chengdu, China, October 2006, paper EX/P R.J. Buttery, et al. (incl. T. Tala), Research Challenges for the ITER Baseline Scenario on JET, 21st IAEA Fusion Energy Conference, Chengdu, China, October T. Tala, et al. (incl. K. Rantamäki), Overview of Poloidal and Toroidal Momentum Transport Studies in JET, 21st IAEA Fusion Energy Conference, Chengdu, China, October 2006, paper EX/P V. Parail, et al. (incl. J. Lönnroth, V. Hynönen, T. Kiviniemi, T. Kurki-Suonio), Theoretical analysis and predictive modelling of ELMs mitigation by enhanced toroidal ripple and ergodic magnetic field, 21st IAEA Fusion Energy Conference, Chengdu, China, 16-21October 2006, paper TH/ R. Moyer, et al., (incl. J. Lönnroth), Edge Localized Mode Control in DIII-D using Magnetic Perturbation-Induced Pedestal Transport Changes, 21st IAEA Fusion Energy Conference, Chengdu, China, October 2006, paper EX/ A.Eriksson, et al., (incl. T. Tala), Predictive simulations of toroidal momentum transport in JET, 48th Annual Meeting of APS Division of Plasma Physics, Philadelphia, USA, October 30 - November 3rd, Y. Kamada, et al., (incl. J.S. Lönnroth), Edge pedestal physics and its implications for ITER, 21st IAEA Fusion Energy Conference, Chengdu, China, 16-21October 2006, paper IT/ J. Ongena, et al., (incl. M.J.Mantsinen, A.Salmi, M.Santala), Recent progress in JET on Heating and Current Drive studies in view of ITER, 21st IAEA Fusion Energy Conference, Chengdu, China, October 2006, paper EX/P T. Hellsten, et al., (incl. M. Mantsinen), Fast Wave Current Drive and Direct Electron Heating in JET ITB Plasmas, 21st IAEA Fusion Energy Conference, Chengdu, China October 2006, paper EX/P

169 2 Fusion Technology 2.1 Publications in Scientific Journals Fusion Technology 319. S. Tähtinen, P. Moilanen, B. Singh and D. J. Edwards, Tensile and fracture toughness properties of unirradiated and neutron irradiated titanium alloys, Journal of Nuclear Materials (2002) S. Tähtinen, A. Laukkanen, B. Singh and P. Toft, Properties of copper-stainless steel HIP joints before and after neutron irradiation, Journal of Nuclear Materials (2002) P. Träskelin, E. Salonen, K. Nordlund, A.V. Krasheninnikov, J. Keinonen, C. H. Wu, Chemisorption of CH3 radicals on carbon first wall structures, Journal Nuclear Materials (2003) E. Salonen, K. Nordlund, J. Keinonen, and C. H. Wu, Molecular dynamics studies of the sputtering of divertor materials, Journal of Nuclear Materials (2003) J.P. Coad, P. Andrew, D.E. Hole, S. Lehto, J. Likonen, G.F. Matthews, M. Rubel, Erosion/deposition in JET during the period , Journal of Nuclear Materials (2003) B.N. Singh, S. Tähtinen, P. Moilanen, P. Jacquet and J. Dekeyser, In-reactor Uniaxial Tensile Testing of Pure Copper at a Constant Strain Rate at 90 o C Journal of Nuclear Materials 320 (2003) J. Likonen, S. Lehto, J.P. Coad, T. Renvall, T. Sajavaara, T. Ahlgren, D.E. Hole, G.F. Matthews, J. Keinonen, Studies of impurity deposition/implantation in JET divertor tiles using SIMS and ion beam techniques, Fusion Engineering and Design (2003) S. Lehto, J. Likonen, J.P. Coad, T. Ahlgren, D.E. Hole, M. Mayer, H. Maier, J. Kolehmainen, Tungsten coating on JET divertor tiles for erosion/deposition studies, Fusion Engineering and Design (2003) J. Pitkänen, P. Kauppinen, H. Jeskanen, S. Tähtinen and S. Sandlin, Ultrasonic Studies on ITER Divertor and First Wall Modules, Fusion Engineering and Design (2003) R. Lässer, et al., (incl. J. Likonen), Tritium Related Studies at the JET Facilities, Fusion Engineering and Design 69 (2003) M. Siuko, M. Pitkäaho, A. Raneda, J. Poutanen, J. Tammisto, J. Palmer and M. Vilenius, Water hydraulic actuators for ITER maintenance devices, Fusion Engineering and Design 69 (2003) P. Gravez, et al., (incl. A. Raneda, M. Siuko), Model-based remote handling with the MAE- STRO hydraulic manipulator, Fusion Engineering and Design 69 (2003) A.Raneda, P. Pessi, M. Siuko, H. Handroos, J. Palmer and M. Vilenius, Utilization of virtual prototyping in development of CMM, Fusion Engineering and Design 69 (2003) L.P. Jones, et al., (incl. T. Jokinen), Towards Advanced Welding Methods for the ITER Vacuum Vessel Sectors, Fusion Engineering and Design 69 (2003) H. Wu, H. Handroos, J. Kovanen, A. Rouvinen, P. Hannukainen, T. Saira and L. Jones, Design of Parallel Intersector Weld/Cut Robot for Machining Processes in ITER Vacuum Vessel, Fusion Engineering and Design 69 (2003) T. Jokinen, V. Kujanpää, High Power Nd:YAG Laser Welding in Manufacturing of Vacuum Vessel of Fusion Reactor, Fusion Engineering and Design 69 (2003) T. Eurajoki, M.P. Frias and S. Orlandi, Trends in radiation protection: possible effects on fusion power plant design, Fusion Engineering and Design 69 (2003) Y. Lechon, et al., (incl. R. Korhonen), External costs of silicon carbide fusion power plants compared to other advanced generation technologies, Fusion Engineering and Design 69 (2003) B. Hallberg, et al., (incl. R. Korhonen), External costs of material recycling strategies for fusion power plants, Fusion Engineering and Design 69 (2003) T. Eurajoki, M. Ek, Feasibility of L/ILW fission waste repository concepts for fusion waste, Fusion Engineering and Design 69 (2003) P. Träskelin, E. Salonen, K. Nordlund, A.V. Krasheninnikov, J. Keinonen, and C. H. Wu, Molecular dynamics simulations of CH3 sticking on carbon surfaces, Journal of Applied Physics 93 (2003) A.S. Salminen, V.P. Kujanpää, Effect of wire feed position on laser welding with filler wire, Journal of Laser Application 15 (2003) T. Jokinen, M. Karhu, V. Kujanpää, Welding of thick austenitic stainless steel using Nd:YAG laser with filler wire and hybrid process, Journal of Laser Applications 15 (2003) S. Esqué, A. Raneda, and A. Ellman, Techniques for studying a mobile hydraulic crane in virtual reality, International Journal of Fluid Power 4, Fluid Power Net International FPNI, (2003)

170 343. A.T.Peacock, et al., (incl. S. Tähtinen), Overview of recent European materials R&D activities related to ITER, Journal of Nuclear Materials (2004) P. Träskelin, E. Salonen, K. Nordlund, J. Keinonen, and C.H. Wu, Molecular dynamics simulations of CH 3 sticking on carbon surfaces, angular and energy dependence, Journal of Nuclear Materials 334 (2004) J. Tuisku, M. Ahoranta, A. Korpela, J. Lehtonen, R. Mikkonen, R. Perälä, Cryogenic design of an isodynamic magnetic separator, IEEE Transactions on Applied Superconductivity 14 (2004) T. Hartikainen, A. Korpela, J. Lehtonen, R. Mikkonen, A comparative life-cycle assessment between NbTi and copper magnets, IEEE Transactions on Applied Superconductivity 14 (2004) T. Hartikainen, A. Korpela, J. Lehtonen, R. Mikkonen, A comparative life-cycle assessment between NbTi and copper magnets IEEE Transactions on Applied Superconductivity 14 (2004) K. Nordlund, Atomistic simulation of radiation effects in carbon-based materials and nitrides, Nuclear Instruments and Methods in Physics Research 218 (2004) K. Heinola, T. Ahlgren, W. Rydman, J. Likonen, L. Khriachtchev, J. Keinonen and C.H. Wu, Effect of hydrogen on flaking of carbon films on Mo and W, Physica Scripta. T108 (2004) K.O.E. Henriksson, K. Nordlund, J. Keinonen, D. Sundholm, and M. Patzschke, Simulations of the initial stages of blistering in helium implanted tungsten, Physica Scripta T108 (2004) H. Pantsar, A. Salminen, A. Jansson, V. Kujanpää, Quality and Costs Analysis of Laser Welded All Steel Sandwich Panels, Journal of Laser Applications 16 (2004) H. Pantsar and V. Kujanpää, Diode laser beam absorption in laser transformation hardening of low alloy steel, Journal of Laser Applications 16 (2004) R. Karppi, V. Kujanpää, Trendi v tehynologiji in prenos tehnologije na finskem S posebnim ozirom na varjenje in tehnike spanjanja (Trends in technology and technology transfer in Finland with special reference to welding and joining techniques), Varilna Teknika, 53 (2004) V. Kujanpää and H. Martikka, Analytical and nonlinear FEM simulation of contact damage of hardened gearsars, IFNA-ANS International Journal Problems of nonlinear analysis in engineering systems 10 (2004) M. Warrier, R. Schneider, E. Salonen, and K. Nordlund, Multi scale modeling of hydrogen isotope diffusion in graphite, Contributions to Plasma Physics 44 (2004) M.J. Rubel, et al., (incl. J. Likonen), Overview of tracer techniques in studies of material erosion, re-deposition and fuel inventory in tokamaks, Journal of Nuclear Materials (2004) M. Warrier, R. Schneider, E. Salonen, and K. Nordlund, Modelling of the diffusion of hydrogen in porous graphite, Physica Scripta T108 (2004) J. Wallenius, P. Olsson, C. Domain, K. Nordlund, and L. Malerba, Modelling of alpha-prime phase formation in Fe-Cr, Physical Reviews B 72 (2005) E. Vainonen-Ahlgren, J. Likonen, T. Renvall, et al., Studies on 13C deposition in ASDEX Upgrade", Journal of Nuclear Materials 337 (2005) J. Likonen, E. Vainonen-Ahlgren, J.P. Coad, R. Zilliacus, T. Renvall, D.E. Hole, M. Rubel, K. Arstila, G.F. Matthews, M. Stamp and JET- EFDA Contributors, Beryllium accumulation at the inner divertor of JET, Journal of Nuclear Materials 337 (2005) A.Laitinen, J. Liimatainen, P. Hallila, Manufacturing technology development for vacuum vessel and plasma facing components, Fusion Engineering and Design (2005) H. Wu, H. Handroos, P. Pessi, J. Kilkki and L. Jones, Development and control towards a parallel water hydraulic weld/cut robot for machining processes in ITER vacuum vessel, Fusion Engineering and Design (2005) P. Träskelin, K. Nordlund, and J. Keinonen, He, Ne, Ar-bombardment of carbon first wall structures, Nuclear Instruments and Methods in Physics Research B 228 (2005) K. O. E. Henriksson, K. Nordlund, and J. Keinonen, Molecular dynamics simulations of helium cluster formation in tungsten, Nuclear Instruments and Methods in Physics Research B 244 (2005)

171 365. N. Juslin, J. Nord, K. O. E. Henriksson, P. Träskelin, E. Salonen, K. Nordlund, P. Erhart, and K. Albe, Analytical interatomic potential for modelling non-equilibrium processes in the W-C-H system, Journal of Applied Physics 98 (2005) K. O. E. Henriksson, K. Nordlund, A. Krasheninnikov, and J. Keinonen, Differences in hydrogen and helium cluster formation, Applied Physics Letters 87 (2005) R.A. Pitts, et al., (incl. J. Likonen), Material erosion and migration in tokamaks, Plasma Physics and Controlled Fusion 47 (2005) B303-B K. Krieger, J. Likonen, M. Mayer, R. Pugno, V. Rohde, E. Vainonen-Ahlgren, ASDEX Upgrade Team, Tungsten redistribution patterns in ASDEX Upgrade, Journal of Nuclear Materials 337 (2005) M. Mayer, V. Rohde, J. Likonen, E. Vainonen- Ahlgren, K. Krieger, X. Gong, J. Chen, Carbon Erosion and Deposition on the ASDEX Upgrade Divertor Tiles, Journal of Nuclear Materials 337 (2005) R. Pugno, et al., (incl. J. Likonen, E. Vainonen- Ahlgren), Carbon chemical erosion in H-mode discharges in ASDEX Upgrade divertor Iib: flux dependence and local redeposition, Journal of Nuclear Materials 337 (2005) V. Rohde, et al., (incl. J. Likonen, E. Vainonen- Ahlgren), Carbon erosion and a:c-h layer formation at ASDEX Upgrade, Journal of Nuclear Materials 337 (2005) J.P. Coad, H-G. Esser, J. Likonen, et al., Diagnostics for studying deposition and erosion processes in JET, Fusion Engineering and Design 74 (2005) P. Lorenzetto, et al., (incl. J. Liimatainen, S. Tähtinen), Manufacture of blanket shield modules for ITER, Fusion Engineering and Design (2005) M. Irving, J. Palmer and M. Siuko, Generic control system design for the Cassette Multifunctional Mover and other ITER remote handling equipment, Fusion Engineering and Design (2005) J. Palmer, M. Siuko, et al., Recent developments towards ITER 2001 divertor maintenance, Fusion Engineering and Design (2005) C. Grisolia, N. Bekris, J. Likonen, et al., JET contribution to ITER fuel cycle issues, Fusion Engineering and Design (2005) S.P. Simakov, et al., (incl. F. Wasastjerna), Neutronics and Activation Characteristics of the International Fusion Material Irradiation Facility, Fusion Engineering and Design (2005) S. Rosanvallon, et al., (incl. J. Likonen), Tritium related studies within JET Fusion Technology work programme, Fusion Science and Technology 48 (2005) J.P. Coad, et al., (incl. J. Likonen, E. Vainonen-Ahlgren), Distribution of hydrogen isotopes, carbon and beryllium on in-vessel surfaces in the various JET divertors, Fusion Science and Technology 48 (2005) M. Rubel, J.P. Coad, D. Hole, J. Likonen, E. Vainonen-Ahlgren and EFDA-JET Contributors, Fuel Retention in the Gas Box Divertor of JET, Fusion Science and Technology 48 (2005) T. Ahlgren, K. Heinola, E. Vainonen-Ahlgren, J. Likonen and J. Keinonen, Quantification of deuterium irradiation induced defect concentrations in tungsten, Nuclear Instruments and Methods in Physics Research B 249 (2006) K. Nordlund, Atomistic simulations of plasmawall interactions in fusion reactors, Physica Scripta T124 (2006) N. Juslin, P. Erhart, P. Träskelin, J. Nord, K. O. E. Henriksson, K. Nordlund, E. Salonen, and K. Albe, Analytical interatomic potential for modelling non-equilibrium processes in the W-C-H system, Journal of Applied Physics 98 (2005) K. O. E. Henriksson, K. Nordlund, A. Krasheninnikov, J. Keinonen, The depths of hydrogen and helium bubbles in tungsten - a comparison, Fusion Science & Technology 50 (2006) T. Ahlgren, K. Heinola, E. Vainonen-Ahlgren, J. Likonen and J. Keinonen, Quantification of deuterium irradiation induced concentrations in tungsten, Nuclear Instruments and Methods B 249 (2006) 386. K. O. E. Henriksson, K. Vörtler, S. Dreissigacker, K. Nordlund, and J. Keinonen, Sticking of atomic hydrogen on the tungsten (001) surface, Surface Science 600 (2006) J.P. Coad, J. Likonen, M. Rubel, E. Vainonen- Ahlgren, D.E. Hole, T. Sajavaara, T. Renvall and G.F. Matthews, Overview of Material Re-deposition and Fuel Retention Studies at JET with the Gas Box Divertor, Nuclear Fusion 46 (2006)

172 388. D. Terentyev, et al., (incl. K. Nordlund), Displacement cascades in Fe-Cr: a molecular dynamics study, Journal of Nuclear Materials 349 (2006) C. Grisolia, et al., (incl. J Likonen), JET Contributions to ITER Technology Issues, Fusion Engineering and Design 81 (2006) U. Fischer, et al., (incl. F. Wasastjerna), Overview of recent progress in IFMIF neutronics, Fusion Engineering and Design 81 (2006) E. Salonen, Overview of the atomistic modelling of the chemical erosion of carbon, Journal of Nuclear Materials (2003), accepted for publication E. Salonen, T. Järvi, K. Nordlund, and J. Keinonen, Effects of the surface structure and cluster bombardment on the self-sputtering of molybdenum, J. Physics of Condensed. Matter (2003), accepted for publication P. Träskelin, K. Nordlund, and J. Keinonen, H, He, Ne, Ar-bombardment of amorphous hydrocarbon structures, accepted for publication in Journal of Nuclear Materials (2005) D. Terentyev, et al., (incl. K. Nordlund), Effect of the interatomic potential on the features of displacement cascades in alpha-fe: a molecular dynamics study, accepted for publication in Journal of Nuclear Materials (2005) R. Schneider, et al., (incl. E. Salonen, K. Nordlund), Dynamic Monte-Carlo modelling of hydrogen isotope reactive-diffusive transport in porous graphite, submitted for publication Journal of Nuclear Materials ( M. Warrier, R. Schneider, E. Salonen, and K. Nordlund, Multi-scale modeling of hydrogen isotope transport in porous graphite, accepted for publication in Journal of Plasma Physics (2005) K. Nordlund, E. Salonen, A. V. Krasheninnikov, and J. Keinonen, Swift chemical sputtering of covalently bonded materials, accepted for publication in Pure and Applied Chemistry (2005) T. Jokinen, Novel Ways of Using Nd:YAG Laser for Welding of Thick Section Austenitic Stainless Steel, accepted for publication in Welding in the World (2005) T. Jokinen, M. Karhu, R. Hedman and V. Kujanpää, Controlling of Root Weld Formation with Through Current System in Electron Beam Welding, accepted for publication in Welding in the World (2005) K. Nordlund, J. Wallenius, and L. Malerba, Molecular dynamics simulations of threshold energies in Fe, submitted for publication in Nuclear Instruments and Methods in Physics Research B (2005) L. I. Vergara, et al., (incl. P. Träskelin, E. Salonen), Methane production from ATJ graphite by slow atomic and molecular D ions: evidence for projectile molecule-size-dependent yields at low energies, accepted for publication in Journal of Nuclear Materials (2006) P. Träskelin, N. Juslin, P. Erhart, and K. Nordlund, Hydrogen bombardment simulations of tungsten-carbide surfaces, submitted for publication in Physical Review B (2006) L. Malerba, et al., (incl. N. Juslin, K. Nordlund), Modelling of radiation damage in Fe-Cr alloys, submitted for publication in Journal of Nuclear Materials (2006) A.Björkas, K. Vörtler, and K. Nordlund, Major elemental assymetry and recombination effects in irradiated WC, submitted for publication in Physical Review B (Rapid communication) (2006), 405. M. Victoria, et al., (incl. K. Nordlund), Modelling irradiation effects in fusion materials, submitted for publication in Fusion and Engineering Design (2006) A.Björkas, K. Nordlund, L. Malerba, D. Terentyev, and P. Olsson, Simulation of displacement cascades in Fe 90 Cr 10 using a two band model potential, submitted for publication in Journal of Nuclear Materials (2006), 407. P. Träskelin, C. Björkas, N. Juslin, K. Vörtler, and K. Nordlund, Radiation damage in WC studied with MD simulations, submitted for publication in Nucl. Instr. Meth. Phys. Res. B (2006), IBMM conference paper D. J. Edwards, B. N. Singh and S. Tähtinen, Effect of Heat Treatments on Precipitate Microstructure and Mechanical Properties of a CuCrZr Alloy, submitted for publication in Journal of Nuclear Materials S. Tähtinen, P. Moilanen and B. N. Singh. Effect of Displacement Dose and Irradiation Temperature on Tensile and Fracture Toughness Properties of Titanium Alloys, submitted for publication in Journal of Nuclear Materials A. Kärkkäinen, J. Kyynäräinen, J. Saarilahti, A. Oja, and H. Seppä, Micromechanical magnetometer for ITER, submitted for publication in Transducers (2006) 163

173 2.2 Conference Articles Fusion Technology 411. S. Tähtinen, M. Asikainen, R. Rintamaa and H. Tuomisto, Guest Editors in Proceeding of the 22nd Symposium on Fusion Technology (SOFT-22), Helsinki, Finland, 9-13 September 2002, Fusion Engineering and Design Issues 1-4 (2003) and S.K. Sipilä and J.A. Heikkinen (Eds.) 10th European Fusion Theory Conference, Book of Abstracts, Helsinki University of Technology, Report TKK-F-A823, Espoo V. Kujanpää, P.Maaranen, T. Kostamo, Effect of parameters in diode laser welding of steel sheets, First Int. Symposium on High-power laser macroprocessing (LAMP2002), May 27-31, 2002, Osaka, Japan, Proceedings of SPIE. Volume 4831, SPIE (2003) pp T. Jokinen, P. Jernström, M. Karhu, I. Vanttaja, V. Kujanpää, Optimisation of parameters in hybrid welding of aluminium alloy, First Int. Symposium on High-power laser macroprocessing (LAMP2002), May 27-31, 2002, Osaka, Japan, Proceedings of SPIE. Volume 4831, SPIE (2003) pp A. Jansson, J. Ion, V. Kujanpää, CO 2 and Nd:YAG laser cladding using Stellite 6, First Int. Symposium on High-power laser macroprocessing (LAMP2002), May 27-31, 2002, Osaka, Japan, Proceedings of SPIE. Volume 4831, SPIE (2003) pp A. Salminen, A. Jansson, V. Kujanpää, Effect of welding parameters on high-power diode laser welding on thin sheet, Photonics West, Laser 2003, Conference 4973, January 2003, SPIE, San Jose, CA, USA, pp K. Heinola, T. Ahlgren, W. Rydman, J. Likonen and J. Keinonen, Flaking mechanism of carbon films on Mo, Proceedings of XXXVII Annual Conference of the Finnish Physical Society, March 2003, Helsinki, Finland, Report Series in Physics, HU-P-265, (2003) H. Wu, H. Handroos, L. Jones, On the Design of Hydraulically Driven Parallel Intersector Weld/ Cut Robot in ITER Vacuum Vessel, Proceedings of the International Symposium on Fluid Power Transmission and Control (ISFP 03), Wuhan, China, 8-10 April K. Dufva, J. Kovanen, H. Wu, H. Handroos, T. Saira, On the Force Transmission Capability of the Penta-WH the Parallel Weld/Cut Robot for Fusion Reactor, Proceedings of the 8th Scandinavian International Conference on Fluid Power, Tampere, Finland 7-9 May Y. Liu, H. Handroos, Control of Penta-WH the Parallel Weld/Cut Robot for Fusion Reactor, Proceedings of the 8th Scandinavian International Conference on Fluid Power, Tampere, Finland 7-9 May T. Jokinen and V. Kujanpää, High power Nd:YAG -laser welding in manufacturing of vacuum vessel of fusion reactor, JOIN 2003-conference, May 2003, Lappeenranta, Finland A. Fellman, P. Jernström, V. Kujanpää, The Effect of Shielding Gas Composition in Hybrid Welding, JOIN 2003-Conference, May 2003, Lappeenranta, Finland V. Kujanpää and A. Salminen, Absorption phenomena in laser welding, (invited lecture), JOIN 2003-Conference, May 2003, Lappeenranta, Finland A. Salminen, A. Jansson, V. Kujanpää, High power diode laser welding of thin steel sheet, JOIN 2003-Conference, May 2003, Lappeenranta, Finland G.F. Matthews, J.P. Coad, J. Likonen, et al., Material Migration in JET, 30th EPS Conference on Controlled Fusion and Plasma Physics, 7 11 July 2003, St Petersburg, Russia, European Conference Abstracts 27A (2003) P A. Fellman, P. Jernström, V. Kujanpää, The Effect of Shielding Gas Composition in Hybrid Welding of Carbon Steel, Nordic Laser Material Processing Conference (NOLAMP 9), 4-6 August 2003, Trondheim, Norway, (ed. E.Halmöy), p H. Pantsar and V. Kujanpää, The hardness and depth of the martensitic layer in laser transformation hardened steel 42CrMo4, Nordic Laser Material Processing Conference (NOLAMP 9), 4-6 August 2003, Trondheim, Norway, (ed. E.Halmöy), p A.Salminen, H. Pantsar, A. Jansson and V. Kujanpää, Manufacturing Procedure of Laser Welded All Steel Sandwich Panels, Nordic Laser Material Processing Conference (NOLAMP 9), 4-6 August 2003, Trondheim, Norway, (ed. E.Halmöy), p A.Salminen, A. Jansson and V. Kujanpää, High Power Diode Laser Welding of Steel with Different Joint Configurations, Nordic Laser Material Processing Conference (NOLAMP 9), 4-6 August 2003, Trondheim, Norway, (ed. E. Halmöy), p

174 430. K. Heinola, T. Ahlgren, W. Rydman, J. Likonen, L. Khriachtchev, J. Keinonen and C.H. Wu, Effect of hydrogen on flaking of carbon films on Mo and W, presented at 10th International Workshop on Carbon Materials for Fusion Application, September 2003, Jülich, Germany A. Fellman, P. Jernström and V. Kujanpää, CO 2 -GMA Hybrid Welding of Carbon Steel, 22 nd Int. Congress on Applications of Lasers and Electro-Optics (ICALEO 2003), Jacksonville, FL, USA, October 2003, CD-ROM, Laser Institute of America (LIA), p. A56 A T. Jokinen, M. Karhu, V. Kujanpää, Narrow gap hybrid welding of thick stainless steel, 22 nd Int. Congress on Applications of Lasers and Electro-Optics (ICALEO 2003), Jacksonville, FL, USA, October 2003, CD-ROM, Laser Institute of America (LIA), p. A66 A M. Karhu, T. Jokinen, V. Kujanpää, Welding experiments using vacuum environment with Nd:YAG laser, 22 nd Int. Congress on Applications of Lasers and Electro-Optics (ICALEO 2003), Jacksonville, FL, USA, October 2003, CD-ROM, Laser Institute of America (LIA), p. A255 A H. Pantsar, V. Kujanpää, Comparison of diode laser transformation hardening of heat treatable steels and martensitic stainless steels, 22 nd Int. Congress on Applications of Lasers and Electro-Optics (ICALEO 2003), Jacksonville, FL, USA, October 2003, CD-ROM, Laser Institute of America (LIA), p. D29 D A. Salminen, J. Siltanen, A. Jansson, V. Kujanpää, Effect of joint configuration on high power diode laser welding of steel, 22 nd Int. Congress on Applications of Lasers and Electro-Optics (ICALEO 2003), Jacksonville, FL, USA, October 2003, CD-ROM, Laser Institute of America (LIA), p. D37 D P. Norajitra, et al., (incl. F. Wasastjerna), Conceptual Design of the EU Dual-Coolant Blanket (Model C), Symposium on Fusion Engineering (SOFE), San Diego CA, USA October 14 17, A. Raneda, J. Vilenius, M. Hyvönen, and K. Huhtala, Virtual Prototype of a Remote Controlled Hydraulic Mobile Machine for Teleoperation Control Development, Proc. of 1st International Conference on Computational Method in Fluid Power Technology, Methods for solving practical problems in Design and Control, November 2003, Melbourne, Australia, Stecki, J.S. (Ed.) 13 pp B.N. Singh, S. Tähtinen, P. Moilanen, et al., Deformation Behaviour of Pure Copper During In-reactor tensile tests at 90 o C, 11 th International Conference on Fusion Reactor Materials ICFRM- 11, Kyoto, Japan, December 7-12 (2003) A. Salminen, A. Fellman and V. Kujanpää, Effect of Joint Configuration on the Efficiency of High Power Diode Laser Welding of Steel, Photonics West 2004, High-Power Diode Laser Technology and Applications II, Proceedings of SPIE, January, 2004, San Jose California, USA., Ed. M.S. Ediker, Vol. 5336, pp J. Palmer, et al., (incl. M. Siuko), Remote Maintenance of the ITER Divertor, ANS 10th International Conference on Robotics and Remote Systems for Hazardous Environments, Gainesville, Florida, March 28-31, Y. Chen, F. Wasastjerna, et al., Three-dimensional shielding calculations of the IFMIF neutron source using a coupled Monte Carlo deterministic computational scheme, 10th International Conference on Radiation Shielding/13 th Topical Meeting on Radiation Protection and Shielding, Funchal, Madeira, 9-14 May U. Fischer, Y. Chen, S.P. Simakov, P. Vladimirov, F. Wasastjerna, Overview of Recent Progress in IFMIF Neutronics, ISFNT-7, Tokio, Japan, May 22 27, G.F. Matthews, P. Coad, J. Likonen, et al., (incl. E. Vainonen-Ahlgren), Beryllium and carbon migration in JET, Proceedings of the 16th PSI - Plasma Surface Interactions conference (Portland, Maine, USA, May 2004) M. Rubel, J.P. Coad, J. Likonen, G.F. Matthews, E. Vainonen-Ahlgren In-vessel Diagnostic for Erosion and Re-deposition Studies in JET: High-Z Metal Coated Divertor and Limiter Marker Tiles, Proceedings of the 16th PSI - Plasma Surface Interactions conference (Portland, Maine, USA, May 2004) V. Kujanpää, A. Salminen, Reflections of laser beam and pool phenomena in laser welding with filler wire and arc, IIW Annual Meeting, Osaka, Japan, July 2004, IIW IV V. Kujanpää, Absorption phenomena in laser welding, Finnish-German Japanese Seminar, Awaji, Japan, July J.P. Coad, et al., (incl. J. Likonen, E. Vainonen- Ahlgren), Distribution of hydrogen isotopes, carbon and beryllium on in-vessel surfaces in the various JET divertors, TRITIUM th International Conference on Tritium Science and Technology, Baden-Baden, Germany, September

175 448. S. Rosanvallon, et al., (incl. J. Likonen), Tritium related studies within JET Fusion Technology workprogramme, TRITIUM th International Conference on Tritium Science and Technology, Baden-Baden, Germany, September M. Rubel, J.P. Coad, D. Hole, J. Likonen, E. Vainonen-Ahlgren, Fuel Retention in the Gas Box Divertor at JET, TRITIUM th International Conference on Tritium Science and Technology, Baden-Baden, Germany, September S. Tähtinen, B. N. Singh, P. Moilanen, P. Jacquet and J, Dekeyser, D. J. Edwards, Deformation Behaviour of Copper Under In-reactor Uniaxial Tensile Tests, 23rd Symposium on Fusion Technology, Fondazione Cini, Venice, Italy, September A. Fellman, A. Salminen, and V. Kujanpää, The effect of welding parameters in CO2 laser-mag hybrid welding of butt joints, Laser Assisted Net shape Engineering 4, ed. M. Geiger, A. Otto, Proc. Conf. LANE 2004, September 2004, Erlangen, Germany, A. Fellman, A. Salminen and V. Kujanpää, The comparison of the effects of welding parameters and weld properties of T-butt Joints between CO2-laser, Nd:YAG-laser and CO2-GMA hybrid welding, 23rd Int. Congress on ICALEO 2004 Applications of Lasers & Electro-Optics, October 4-7, 2004, San Francisco, CA, U.S.A, 9 pp A. Salminen, A. Fellman and V. Kujanpää, Effect of Material Thickness and Joint Configuration on the Efficiency of High Power Diode Laser Welding of Steel, 23rd International Congress on ICALEO 2004 Applications of Lasers & Electro-Optics, October 4-7, 2004, San Francisco, CA, U.S.A,8p H. Pantsar and V. Kujanpää, The effect of Processing parameters on the microstructure and hardness of laser transformation hardened tool steel, 23 rd Int. Congress on ICALEO 2004 Applications of Lasers & Electro-Optics, October 4-7, 2004, San Francisco, CA, U.S.A, 455. M. Mayer, V. Rohde, J. Likonen, E. Vainonen- Ahlgren, J. Chen, X. Gong, K. Krieger, ASDEX Upgrade Team, Carbon Deposition and Deuterium Inventory in the ASDEX Upgrade, 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 1-6 November 2004, Ex/ H. Handroos, H. Wu, P. Pessi, Design of Mechatronic Water Hydraulic Parallel Robot for Machining in Fusion Reactor, Mechatronics Day 2004, DTU, Denmark 457. H. Wu, H. Handroos, J. Kilkki, Kovanen, P. Pessi, L. Jones, Development of a Parallel Mechanism Machine for a Fusion Reactor, ISR2004, Paris H. Wu, H. Handroos, P. Pessi, Design and Development of a Water Hydraulic Parallel Robot for Machining in Fusion Reactor, ICMA 04, Osaka T. Ahlgren, K. Heinola, E. Vainonen-Ahlgren, J. Likonen, A. Hallen and J. Keinonen, Deuterium out-diffusion from W, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo, 229 (2005) J. Likonen, E. Vainonen-Ahlgren, R. Zilliacus, T. Renvall, K. Arstila, J.P. Coad, D.E. Hole, M. Rubel, G.F. Matthews, M. Stamp, Beryllium Accumulation at the Inner Divertor of JET, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo, 229 (2005) E. Vainonen-Ahlgren, J. Likonen, T. Renvall, V. Rohde, R. Neu, R. Pugno, K. Krieger, M. Mayer, Studies on 13C Migration at Asdex Upgrade Tokamak, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo, 229 (2005) K. Heinola, T. Ahlgren, E. Vainonen-Ahlgren, J. Likonen and J. Keinonen, Deuterium Measurement by D(3He,p)4He Reaction in Fusion Materials, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo, 229 (2005) S. Tähtinen, B. N. Singh, P. Moilanen, S. Saarela, P. Jacquet, J. Dekeyser and D. J. Edwards, Deformation behaviour of copper under in-reactor uniaxial tensile tests, Proceedings of the XXXIX Annual Meeting of the Finnish Physical Society, Dipoli, Espoo, March 2005, T. Ruokola and J. von Boehm (eds.), Helsinki University of Technology Laboratory of Physics, Espoo, 229 (2005)

176 464. S.P. Simakov, et al., (incl. F. Wasastjerna), Evaluation of Radioactive inventories in the IFMIF test cell, (KTG-2005) Jahrestagung Kerntechnik, 2005, Annual Meeting on Nuclear Technology 2005 Nürnberg on May U. Fischer, Y. Chen, S.P. Simakov, P. Vladimirov, F. Wasastjerna, Overview of Recent Progress in IFMIF Neutronics, ISFNT-7 - Seventh International Symposium on Fusion Nuclear Technology, Tokyo, Japan, May C. Grisolia, et al., (incl. J. Likonen), JET Contributions to ITER Technology Issues, ISFNT-7 - Seventh International Symposium on Fusion Nuclear Technology, Tokyo, Japan, May H. Koivisto, J. Mattila, A. Mäkelä, M. Siuko, M. Vilenius, On Pressure/Force Control of a Water Hydraulic Joint, 9th Scandinavian International Conference on Fluid Power, Linköping, Sweden, 1-3 June 2005, pp A.Fellman, A. Salminen and V. Kujanpää, A Study of the Molten Filler Material Movements during CO2-laser-MAG Hybrid Welding,, Conference on Lasers in Manufacturing (LIM2005), München, Germany, June P. Aaltonen, Y. Yagodzinskyy and H. Hänninen, Effects of Oxide Films and Passivity on Copper in Nitrite Solution at Ambient Temperature, Passivity-9, The Ninth international Symposium on the Passivation of Metals and Semiconductors and the Properties of Thin Oxide Layers, EFC Event n 281, Paris, France, 27 June - 1 July Y. Yagodzinskyy, P. Aaltonen and H. Hänninen, Electrochemical Potential Oscillations during Galvanostatic Passivation of Copper in NaNO2 Solutions and Their Role in R`TGSCC Mechanism, Passivity-9, The Ninth international Symposium on the Passivation of Metals and Semiconductors and the Properties of Thin Oxide Layers, EFC Event n 281, Paris, France, 27 June 1 July V. Rohde, M. Mayer, J. Likonen, E. Vainonen- Ahlgren, Carbon migration at the divertor of ASDEX Upgrade, 32nd EPS Conference on Controlled Fusion and Plasma Physics, Tarragona, Spain, 27 June - 1 July 2005, Europhysics Conference Abstracts 29C (2005) P M.J. Rubel, J.P. Coad, J. Likonen, G.F. Matthews, D. Hole, E. Vainonen-Ahlgren, Material Migration Studies at JET Using Tracer Techniques, 32nd EPS Conference on Controlled Fusion and Plasma Physics, Tarragona, Spain, 27 June - 1 July 2005, Europhysics Conference Abstracts 29C (2005) P T. Jokinen, Novel Ways of Using Nd:YAG Laser for Welding of Thick Section Austenitic Stainless Steel,, IIW Doc. IV , Proceedings of the Annual Meeting of International Institute of Welding, Prague, July T. Jokinen, M. Karhu, R. Hedman and V. Kujanpää, Controlling of Root Weld Formation with Through Current System in Electron Beam Welding, IIW Doc. IV , Proceedings of the Annual Meeting of International Institute of Welding, Prague, July (2005) R. Laitinen, M. Lehtinen, V. Kujanpää, A. Fellman, Influence of laser and hybrid laser- MAG(hybrid) welding on the strength and toughness of the weld HAZ of ultra high strength steel optim RAEX 960 QC, Proceedings of the 10th Nordic Laser Materials Processing Conference (NOLAMP 10), Luleå, Sweden, ed. A. Kaplan, August 2005, V. Kujanpää, Absorption phenomena in laser materials processing,, Proceedings of the 10th Nordic Laser Materials Processing Conference (NOLAMP 10), Luleå, Sweden, ed. A. Kaplan, August 2005, A.Fellman, A. Salminen and V. Kujanpää, The effect of parameters on weld quality and photodiode signals in CO2 laser welding of butt joints, Proceedings of the 10th Nordic Laser Materials Processing Conference (NOLAMP 10), Luleå, Sweden, ed. A. Kaplan, August 2005, A.Fellman, A. Salminen and V. Kujanpää, A Study of the Effects of Parameters on Filler Material Movements and Weld Quality in CO2-laser-MAG Hybrid Welding,, Proceedings of the 10th Nordic Laser Materials Processing Conference (NOLAMP 10), Luleå, Sweden, ed. A. Kaplan, August 2005, A.Mäkelä, J. Mattila, M. Siuko and M. Vilenius, Flow vs. Pressure Servovalves in Force-based Position Control Applications, The 6th JFPS International Symposium on Fluid Power, Tsukuba, Japan, 9-11 November 2005, 5 pp A.Muhammad, J. Mattila, M. Siuko, M. Vilenius, Experimental Comparison of Teleoperation Schemes for Hydraulic Manipulators, The 6th JFPS International Symposium on Fluid Power, Tsukuba, Japan, 9-11 November 2005, 5 pp J. Mattila, M. Siuko and M. Vilenius, On Pressure/Force Control of a 3-DOF Water Hydraulic Manipulator, The Sixth JFPS International Symposium on Fluid Power, Tsukuba, Japan, 9-11 November 2005, 5 pp. 167

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179 2.3 Research Reports Fusion Technology 518. S. Karttunen, K. Rantamäki (Eds.), FFusion 2 Technology Programme , Final Report, Technology Programme Report 1/2003, Tekes, Helsinki 2003, 152 pp S. Karttunen (Ed.), 2002 Annual Report of the Association Euratom-Tekes, VTT Processes, Espoo 2003, 64 pp P. Norajitra, et al., (incl. F. Wasastjerna), Conceptual Design of the Dual-Coolant Blanket within the Framework of the EU Power Plant Conceptual Study, Final Report, Forschungszentrum Karlsruhe Wissenschaftliche Berichte FZKA 6780, Karlsruhe, May Frej Wasastjerna, User s manual for the global IFMIF test cell neutronics model, Final Report on the EFDA Task TW3-TTMI-003, D9, Report IRS-Nr 09/03 FUSION NR 205, Karlsruhe, June Y. Chen, U. Fischer, P. Pereslavtsev, and F. Wasastjerna, The EU power plant conceptual study - neutronic design analyses for near term and advanced reactor models Forschungszentrum Karlsruhe Wissenschaftliche Berichte FZKA 6763, Karlsruhe S. Tähtinen, Manufacturing of ITER Primary First Wall panel by powder HIP method - Status report September 2003, VTT Technical Research Centre of Finland, Research Report BTUO , Espoo, D.J. Edwards, B.N. Singh, S. Tähtinen, P. Moilanen, P. Jacquet and J. Dèkeyser, Status of in-reactor tensile straining of pure copper at a constant strain rate, Fusion materials, Semi-annual progress report for the period ending June 30, DOE/ER-0313/34 (2003) p V. Suolanen and S. Vuori, Complementary study of environmental impact taking into account optimisation, EISS Technical Document SL4-DEL , 2003, 20 pp R. Korhonen, External costs due to disposal of fusion waste, Task Deliverable D2, part: Externalities of waste disposal, long term disposal of fusion waste, in external costs of fusion, Final Report of SERF3 Externalities, CIEMAT 2003, 33 pp R. Korhonen, Global environmental impacts of fusion Impacts and transfer of C-14 releases in the future environment, Task Deliverable D3 report: Re-evaluation of C-14 impacts in the future environment, in External Costs of Fusion, Final Report of SERF3 Externalities, CIEMAT 2003, 23 pp M. Airila (Ed.), Advanced Energy Systems - Annual Report 2003, Helsinki University of Technology Publications in Engineering Physics TKK-F-C196, Espoo A. Laukkanen, K. Wallin, P. Nevasmaa, and S. Tähtinen, Transferability properties of local approach modelling in the ductile to brittle transition reason, ASTM STP:1429, P. Moilanen, S. Tähtinen, B. N. Singh and P. Jacquet TW2-TVV-SITU, In-situ investigation of the mechanical performance and life time of copper, VTT Technical Research Centre of Finland, Research Report BTUO , Espoo, H. Handroos, H. Wu, J. Kovanen, P. Pessi, K. Dufva, Y. Liu, TW2-TVV-ROBOT; Dynamic Test Rig for Intersector Welding Robot (IWR) for VV Sector Field Joining, Final Report, M. Karhu, T. Jokinen, V. Kujanpää, TW2- TVV/EBROOT, Controlling Root Welding Made by Electron Beam with Adaptive System, Research Report BTUO , VTT Industrial systems, Espoo, 2004, 27 pp T. Jokinen, M. Karhu, V. Kujanpää, Hybrid welding in the manufacturing of vacuum vessel for fusion reactor, accepted for publication in Industrial Systems Review, R. Korhonen, Long term impacts of fusion. Progress report of SERF4 subtasks: Re-evaluation of the impacts of C-14 in SERF-studies and Comparison of fusion and fission, p S. Karttunen and K. Rantamäki (Eds.), FU- SION Yearbook, Association Euratom-Tekes, Annual Report 2003, VTT Publications 530, Espoo 2004, 137 pp R. Korhonen, Long term impacts of fusion. Re-evaluation of the impacts of C-14 in SERF-studies, p F. Wasastjerna, User s manual for neutronics model md33 of ther IFMIF test cell, Final report on the EFDA task TW4-TTMI-003-D5a, Internal Report, Forschungszentrum Karlsruhe, IRS-Nr. 03/04-FUSION NR B.N. Singh, et al., (incl. S. Tähtinen, P. Moilanen), Final Report on In-Reactor Tensile Tests on OFHC-Copper and CuCrZr Alloy, Riso National Laboratory, Materials Research Department, Roskilde, Denmark, Riso-R-1481(EN), 2004, 47 pp. 170

180 539. B.N. Singh, D.J. Edwards and S. Tähtinen, Effect of Heat Treatments on Precipitate Microstructure and Mechanical Properties of CuCrZr Alloy, Research report, Risø-R-1436(EN) Risø National Laboratory, Roskilde (December 2004), 24 pp K. O. E. Henriksson, N. Juslin, P. Träskelin and K. Nordlund and J. Keinonen,, Report on hydrogen retention in W, blister formation, and mixed material (W/C) erosion for ITER relevant energy ranges, Report TW4-TPP-CARWMOD: Deliverable 4 (2004) S. Karttunen, K. Rantamäki (Eds), FUSION Yearbook, Association Euratom-Tekes, Annual Report 2004, VTT Publications 567, Espoo 2005, 142 pp M. Airila (Ed.), Advanced Energy Systems - Annual Report 2004, Helsinki University of Technology Publications in Engineering Physics TKK-F-C197, Espoo T. Jokinen, M. Karhu, V. Kujanpää, Hybrid welding in the manufacturing of vacuum vessel for fusion reactor, Industrial Systems Review, VTT Industrial Systems, Espoo 2005, P. Kotiluoto and F. Wasastjerna, Shielding Calculations for a Helium Cooled Pebble Bed Test Blanket Module in ITER, Forschungszentrum Karlsruhe, Interner Bericht IRS-Nr. 07/05 FU- SION Nr. 252 (May 2005) J. Järvenpää, DTP2 Testiympäristö, Research report VTT-R (2005) J. Järvenpää, DTP2 facility design update and integration studies, Research report TUO VTT Industrial Systems, Tampere (2005) M. Karhu, T. Jokinen, V. Kujanpää and H. Wu, TW2-TVV/ROBOT Construction and testing of a high precision intersector welding robot (IWR) test rig for VV sector field joining, Research report TUO , VTT Industrial systems, Espoo (2005) S. Tähtinen and B. N. Singh, Effect of heat treatment and neutron irradiation on tensile and fracture toughness properties of CuCrZr/316L(N) joints, Research report BTUO ,VTT Industrial Systems, Espoo (2005) S. Tähtinen, S. Saarela, Slow strain rate tests of copper stainless steel joint samples, Research report BTUO , VTT Industrial Systems, Espoo (2005) S. Tähtinen and B. N. Singh, Effects of neutron irradiation on mechanical properties of titanium alloys, 2005, Research report BTUO , VTT Industrial Systems, Espoo (2005) S. Tähtinen and S. Saarela, Fracture behaviour of HIP joints between CuCrZr alloy and 316L stainless steel, Research report BTUO , VTT Industrial Systems, Espoo (2005) P. Moilanen and S. Tähtinen, Mechanical testing under simultaneous neutron irradiation, Industrial Systems Review VTT Industrial Systems 2005, S. Karttunen, K. Rantamäki and A. Marttila (Eds.), FUSION Yearbook, Association Euratom-Tekes, Annual Report 2005, VTT Publications 606, Espoo 2006, 131 pp P. Kotiluoto, F. Wasastjerna, Calculating the neutron current emerging through the beam tubes in IFMIF, VTT Research Report No VTT-R , P. Kotiluoto and F. Wasastjerna, IFMIF neutronics work in 2005: Modelling the horseshoe shield, updating the test cell geometry model, shielding calculations for the cover, VTT Research Report No VTT-R , S.P. Simakov, et al., (incl. F. Wasastjerna), Calculation of complete nuclear response through the entire test cell with consideration of an additional neutron shielding block, Final report on the EFDA task TW5-TTMI-003, Deliverable 4, Forschungszentrum Karlsruhe, K. Nordlund, N. Juslin, C. Björkas, L. Malerba, and D. Terentyev, Final report on EFDA subtask EFDA-05-TTMS a, Report series in Physics HU-P-266 (Department of Physical Sciences, University of Helsinki, Helsinki, Finland, 2006) A.Lehtilä, Analyzing the global energy system by using the EFDA Global TIMES model. VTT Working Papers (TW5-TRE-FESO/A), S. Tähtinen, Manufacturing of ITER Primary First Wall panel by powder HIP method. VTT, Research Report VTT-R , P. Moilanen, S. Saarela, S. Tähtinen, P. Jacquet and B. N. Singh, In-reactor Creep-fatigue Tests in BR2 Test Reactor Design, and Construction of a Loading Module. VTT, Research Report VTT-R

181 561. S. Tähtinen and H. Jeskanen, Ultrasonic examination of ITER primary first wall small scale mock-ups DS-10J, DS-6J, DS-7J, DS-13I, PHS-7F, PHS-7Fb, PHS-8F, PHS-9F, MAQ-JAP, PHD-2J, PHS-(21-26)-IR and full scale panel TVV-HIP, VTT, Research Report VTT-R M. Airila and A. Salmi (Ed.), Advanced Energy Systems Annual Report 2005, Helsinki University of Technology Publications in Engineering Physics TKK-F-C198, Espoo General Articles 563. J.S. Janhunen, T.P. Kiviniemi, and J.A. Heikkinen, Transport processes in magnetically confined fusion plasmas, CSC Report on Scientific Computing , edited by K. Kotila et al., pp , S. Tähtinen, M. Asikainen, R. Rintamaa and H. Tuomisto, Preface, Fusion Engineering and Design (2003) M. Siuko, Tampereen teknillisessä yliopistossa suunnitellaanfuusioreaktorin huoltolaitteita, Prizz.Uutiset, 4/2003, 12. p. (in finnish) Taina Kurki-Suonio and Salomon Janhunen, Iteristä ensimmäinen fuusioreaktori, Tiede 1/2003 (in Finnish) Taina Kurki-Suonio, Energia kaipaa fuusiota, Suomen Kuvalehti 40/2003, , pp (in Finnish) M. Airila and V. Hynönen, Fuusio harppaa eteenpäin, Tietoyhteys 2/2004, pp (in Finnish) Seppo Karttunen, Ihmiskunnan haastavin teknologiaprojekti käyntiin - ITER fuusiokoelaitos rakennetaan Eurooppaan, Arkhimedes 4/2005 (2005) M. Airila, Odotellessa - Waiting, ATS Ydintekniikka 1/ (2005) K.M. Rantamäki, Fuusio: mitä, missä, milloin?, ATS-Ydintekniikka 1/ (2005) R. Salomaa, Fuusioreaktori etsii muotoaan, ATS Ydintekniikka 1/ (2005) J. Lönnroth, Som finländsk fusionsforskare ute i världen, ATS Ydintekniikka 1/ (2005) T. Kurki-Suonio, J. Likonen, Fuusiolaitteen diverttori missä tiede ja teknologia kohtaavat, ATS Ydintekniikka 2/ (2005) pp T. Kurki-Suonio, Ensimmäinen oikea fuusioreaktori ITER rakennetaan Ranskaan, Forte Customer magazine of Fortum Corporation 3/2005, pp A. Salmi, Fuusio, Polysteekki 2/2005 (2005) T. Kurki-Suonio, Tie tähtiin on auki, Tiede-lehti 2005 (2005) J.A.Heikkinen, S.J. Janhunen, T.P. Kiviniemi, S. Leerink and F. Ogando, Towards petaflops computations for fusion plasma turbulence, submitted for publication in CSCnews (2006) S. Karttunen and Jouko Koivula, ITER vie fuusiotutkimuksen uuteen aikaan, Enertec 3/2006, 2 pp. (in Finnish). 172

182 ANNEX E Abbreviations ASCOT ASDEX AUG CCD CEA CFC CIEMAT CMM CoA COCONUT CRPP CSU D DAQ DIII-D DIVIMP DLC DOF DRM DTP2 ECRH EFDA EFDA CSU EFET EHO ELM Finnish Orbit-following Monte Carlo Code German Tokamak Device at Garching ASDEX Upgrade Tokamak Charge-Coupled Device (Camera) Atomic Energy Commission, France Carbon Fibre Composite Centro de Investigaciones Energéticas, Medioambientales y Techológicas, Spain Cassette Multi-fuctional Mover Contract of Association Core-Edge Transport Code at JET Centre de Recherches en Physique des Plasmas, Lausanne, Switzerland Close Support Unit Deuterium Data Acquisition System US Tokamak at San Diego Impurity Transport Code Diamond-like Carbon Degree of Freedom Divertor Region Mockup Divertor Test Platform at VTT Tampere Electron Cyclotron Resonance Heating European Fusion Development Agreement EFDA Close Support Unit European Fusion Engineering and Technology Edge Harmonic Oscillations Edge Localised Mode ELMFIRE ENEA ERO FEM FIDO FLR FTU FT-2 FW FZJ FZK HIP H-mode HELENA IAEA ICRF ICRH IFMIF IEA IHA IPP IPP-CZ IRMA ITB Finnish Kinetic Plasma Turbulence Code Italian National Agency for New Technologies Energy and Environment Impurity transport code Finite Element Method ICRH modelling Code Finite Larmor Radius Italian Tokamak at Frascati Small Russioan Tokamak at St. Petersburg First Wall Forschungzentrum Jülich, Germany Forschungzentrum Karlsruhe, Germany Hot Isostatic Pressing High Confinement Mode Equilibrium Code International Atomic Energy Agency Ion Cyclotron Resonance Frequency Ion Cyclotron Resonance Heating International Fusion Materials Irradiation Facility International Energy Agency Institute of Hydraulics and Automation at TUT Institute for Plasma Physics, Garching and Greifswald, Germany Institute for Plasma Physics, Prague, Czech Rebublic Infra Red Movie Analyser Internal Trasport Barrier 173

183 ITER ITPA ITG IWR JET JETTO JIA JOC JT-60U L-mode LHCD LUT LVDT MCNP MD MEMS MISHKA MHD NCLASS NBI NPA NTM PIC PION PPCS QH way in latin, originally International Thermonuclear Experimental reactor International Tokamak Physics Activities Ion Temperature Gradient Instability Intersector Weld/Cut Robor Joint European Torus at Culham Transport Code at JET JET Implementing Agreement JET Operating Contract Japanese Tokamak at Naka Low Confinement Mode Lower Hybrid Current Drive Lappeenranta University of Technology Linear Variable Differential Transformer Neutron Transport Monte Carlo Code Molecular Dynamics Micromechanical Stability Code Magetohydrodynamics Neoclassical Transport Code Neutral Beam Injection Neutral Particle Anylysator Neoclassical Tearing Mode Particle-in-Cell Code ICRH modelling code Conceptual Power Plant Study Quiescent H-Mode RBS RF RTC SCEE SCK-CEN SERF SIMS SOL T TEM TF TIG TKK Tekes TEXTOR TUT TTE UH UKAEA VR VTT VV W7-X Rutherford Backscattering Radio Frequency Real Time Control Second Cassette End-Effector Belgian Nuclear Research Centre Socio-economic Research on Fusion Secondary Ion Mass Spectrometry Scrape-off Layer (plasma edge outside separatrix) Tritium Trapped Electron Mode Toroidal Magnetic Field Tungsten Inert Gas (welding process) Helsinki University of Technology Finnish Funding Agency for Technology and Innovation German Tokamak at Jülich Tampere University of Technology Trace Tritium Experiment University of Helsinki United Kingdom Atomic Energy Authority Science Council, Sweden Technical Research Centre of Finland Vacuum Vessel German stellarator under construction at Greifswald 174

184 Tekes Technology Reports in English 1/2007 FUSION Technology Programme Report Final Report. 184 p. Seppo Karttunen and Karin Rantamäki (Eds) 17/2006 PINTA Clean Surfaces Final and Evaluation Report. 228 p. 13/2006 Finnish National Evaluation of EUREKA and COST. Evaluation Report. 95 p. Sami Kanninen, Pirjo Kutinlahti, Terttu Luukkonen, Juha Oksanen and Tarmo Lemola 11/2006 Competitiveness through Integration in Process Industry Communities. Evaluation of Technology Programme Process Integration Evaluation Report. 17 p. 8/2006 AVALI Business Opportunities from Space Technology Final Report. 79 p. 6/2006 New Knowledge and Competence for Technology and Innovation Policies ProACT Research Programme Final Report. Edited by Pekka Pesonen. 137 p. 3/2006 ELMO Miniaturising Electronics Final Report. 238 p. 12/2005 NETS Networks of the Future Evaluation Report, Executive Summary. 19 p. Mervi Rajahonka and Mikko Valtakari. 1/2005 NETS Networks of the Future Final Report. 213 p. 10/2004 Competitiveness through internationalisation Evaluation of the means and mechanisms for promoting internationalisation in technology programmes. Evaluation Report. 89 p. Kimmo Halme, Sami Kanninen, Tarmo Lemola, Erkko Autio, Erik Arnold, Jesper Deuten. 6/2004 Developing Technology for Large-Scale Production of Forest Chips Wood Energy Technology Programme Final Report. 98 p. Pentti Hakkila. 22/2003 Presto future products. Added Value with Micro and Precision Technology Final Report. 110 p. 21/2003 Evaluation of the Finnish-Swedish R&D programme EXSITE, , Evaluation Report. 73 p. Risto Louhenperä, Olle Nilsson. 13/2003 Targeted Technology Programmes: A Conceptual Evaluation Evaluation of Kenno, Plastic processing and Pigments technology programmes. Evaluation Report. 104 p. Erkko Autio, Sami Kanninen, Bill Wicksteed. 10/2003 VÄRE Control of Vibration and Sound Technology Programme Final Report. 90 p. 6/2003 Towards a competitive cluter An evaluation real estate and construction technology programmes. Evaluation Report. 89 p. Petri Uusikylä, Ville Valovirta, Risto Karinen, Enno Abel and Thomas Froese Subscriptions:

185 Association Euratom Tekes FUSION Technology Programme Final Report The Finnish Funding Agency for Technology and Innovation Kyllikinportti 2, P.O. Box 69, FIN Helsinki, Finland Tel , Fax , [email protected] February 2007 ISSN ISBN