The Tokamak Concept. Magnetic confinement. The Tokamak concept

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1 Major radius Minor radius The Tokamak Concept ITER is based on the tokamak concept. A tokamak is a device able to produce and confine a large volume of high temperature plasma -a mixture of electrons and ions- in a toroidal shape by means of strong magnetic fields. The original design principle was developed at the Kurchatov Institute in Moscow in the 1960s and due to its ability to maintain the temperature in the plasma the tokamak has become the most advanced magnetically confined fusion concept in the world. Poloidal Field Toroidal Field Magnetic confinement Resultant helical field (exaggerated pitch) Because fusion plasmas are extremely hot above 100 million degrees it is necessary to keep the plasma particles away from the walls of the confinement device as much as possible. This is achieved with a combination of magnetic fields, generated through external coils, and by the current that flows in the plasma. This magnetic cage creates helical field lines inside the machine around which the charged particles of the plasma gyrate and are kept confined. Plasma Current Plasma Magnetic Field Line Poloidal Field Coils Toroidal Field Coil The Tokamak concept The plasma current is normally induced by a transformer coil. Thus in its basic form, a tokamak does not work continuously, but in pulses. In order to achieve steady state operation in a future power plant based on the tokamak principle, at least part of the current has to be driven continuously by means of high-frequency waves or the injection of fast particles. Fortunately, a thermo-magnetic effect - the so-called "bootstrap" current can then provide the remaining part of the current. PULSATOR - an early European tokamak

2 The ITER Machine The overall ITER plant comprises the tokamak, its auxiliaries and supporting plant facilities. ITER has a vertically D shaped plasma and a lower divertor. The divertor is the main area of contact of the plasma and is one of the most critical components in the machine as it controls the amount of impurities in the plasma and has to withstand high surface heat loads of up to 10MW/m 2. The volume of the plasma (850m 3 ) and the thermal insulating or confinement properties give an energy multiplication factor Q ~10, the factor by which the fusion power exceeds the input heating power (50MW) to the plasma. For a future reactor this factor will be typically 30 to 40. External heating can also be used to drive the plasma current, extending the nominal inductive burn of 5 minutes up to about 50 minutes or longer. Plasma control is provided by the poloidal field coils, pumping and fuelling and heating systems interlocked with feedback from diagnostic sensors. Toroidal Field Coil ITER uses low temperature superconducting magnets for both the toroidal and the poloidal coils. These coils, which can generate a magnetic field of 5.3T, hold the plasma inside the vacuum chamber and limit contact with the chamber walls. Access for the heating systems, diagnostics and equipment to be used during the remote maintenance of the machine are distributed Poloidal Field Coil around the vacuum chamber surface at three levels. The inner surfaces of the vacuum vessel are covered with blanket modules which provide the shielding from the high energy neutrons generated by the fusion reactions. In future reactors, these modules will contain lithium for breeding tritium which will then be injected into the plasma in a closed system. A 70K cryostat encloses the whole machine creating a secondary vacuum for the superconducting coils necessary to maintain their temperature of 4K. The Advantages of Fusion Fusion is a safe and environmentally benign energy option offering the possibility of a sustainable and long-term energy supply. Several factors make it particularly attractive for large-scale, base-load electricity production: Almost limitless fuel supply. No greenhouse gas emissions. Suitable for the large-scale electricity production required for the increasing energy needs of modern cities. Waste from fusion will not be a long-term burden on future generations. The transport of radioactive material (Tritium) is not required in the day-to-day operation of a fusion power station. The system has inherent safety aspects. ITER Ma Total fusion power Energy multiplicati Plasma major radiu Plasma minor radiu Plasma current Plasma volume On-axis toroidal fie * Ratio between the generated

3 ITER Research & Development (R&D) Blanket Module ITER is truly a major step forward in fusion research as it incorporates components of high accuracy and magnitude and will be subjected to the full nuclear licensing regulatory control. To minimise investment risk and provide the basis for the licensing activities, the detailed analysis and validation of ITER has been performed through a rigorous R&D programme. Seven large projects have shown that industrial fabrication of the major components of the ITER machine are feasible and that their quality can be assured. Such projects hav also demonstrated the capability of the ITER partners and their industries to collaborate on common projects. The positive results from these seven projects have given confidence for the successful construction and operation of ITER. Vacuum Vessel in Design Parameters 500 MW on factor (Q) * 10 s 6.2 m s 2.0 m 15 MA 850 m 3 ld 5.3 T fusion power and the power injected into the plasma Cryostat Divertor Fuel for Fusion in ITER The fuel for ITER will be a mixture of the two hydrogen species ( isotopes ) deuterium (D) and tritium (T). 10 grams of D and 15 grams of T would meet the entire lifetime energy needs of a citizen living in a developed country! There are about 35 grams of deuterium in every cubic metre of water. Tritium occurs in nature only in trace quantities since it is a radioactive isotope with a halflife of 12 years. Tritium can be produced from lithium, one of the most abundant light metals on earth. However, ITER will burn tritium produced as waste from existing CANDU fission reactors. In ITER the tritium fuel cycle will be demonstrated. Different concepts of tritium breeder blanket modules containing lithium will be tested and their efficiency optimised for use in future demonstration reactors.

4 Courtesy of JAERI Superconducting Magnet Laboratry, Naka Active Metal Casting (AMC) for high heat flux fusion components and space applications Courtesy of PlanseeAG / NASA ITER Central Solenoid Model Coil ITER & Industry The EU Fusion Programme & ITER All fusion research in the European Union (including Associated States) is coordinated in a single programme. Through this joint approach it has been possible to realise the largest and currently most successful fusion experiment in the world, JET (Joint European Torus, Culham, UK) which is supported by the small and medium size devices in the European Member State laboratories. One of the remarkable elements of the European fusion research programme has been the participation in the design of the ITER device. The basic outline of this design follows that of the JET device, which achieved world record results with 16 MW of fusion power in 1997 and demonstrated the international position of excellence that European research has acquired in this area. Fusion research also relies on the expertise and skills of the European industry. The intensive activities for ITER have achieved many direct and significant spin-offs from fusion and plasma research in industrial applications, in areas such as: Plasma Processing of Semiconductor Electronics Materials Coatings Plasma Lighting Plasma Electronics Superconducting Magnets for current applications Advanced diagnostics applied in other energy systems. The exchange of experts and know-how between an international project like ITER and a strong European industry will lead to even more important benefits for all parties. Vertical target divertor prototype Full scale dome liner Plasma processing for chips manufacturing

5 The ITER Site To assess the acceptability of the candidate sites, the ITER final design report defines the requirements and design assumptions, which include the following considerations: Land and seismic aspects Heat sink and water supply Electric power supply Transportation and shipping of heavy and large components Technological and socio-economic infrastructure Licensing, regulation and decommissioning Taking these requirements, a Joint Assessment of Specific Sites (JASS) was carried out by the ITER Negotiators at the end of The final report from this assessment showed that the proposed sites satisfied all the agreed criteria. Two candidate sites are currently being discussed: the European candidate site of Cadarache, in the south of France, and the Japanese one at Rokkasho-mura, in the north of the main island. The ITER negotiators are in discussion to reach a consensus. ITER Cadarache Site Model The Cadarache site is the main research centre of the French Atomic Energy Commission, CEA, for power oriented nuclear research with experience in a broad variety of technologies. Experimental fission reactors, specialised laboratories and workshops for a total of 18 nuclear facilities are hosted at Cadarache. At this site, since the 1980s a major fusion team has worked on the tokamak Tore Supra, the world record holder of discharge duration. At its meeting of the 25/26 March 2004 the European Council of Ministers reaffirmed the unanimous support for the European offer with a view to a rapid commencement of the ITER project at the European candidate site. LEGAL ENTITY LICENSE TO CONSTRUCT TOKAMAK ASSEMBLY STARTS Year MAGNET VESSEL Bid Bid Bid Contract Vendor s Design Contract EXCAVATE PFC BUILDING TOKAMAK BUILDING TOKAMAK ASSEMBLY PFC TFC CS fabrication start OTHER BUILDINGS Install cryostat First sector PFC Complete VV Last TFC Complete blanket/divertor Install CS COMMISSIONING Last CS Abbreviations: PFC = Poloidal Field Coil CS = Central Solenoid TFC = Toroidal Field Coil VV = Vacuum Vessel Contract First sector Last sector ITER Construction Schedule

6 Safety Once in operation, fusion power plants would provide large amounts of base load electric energy, burning a deuteriumtritium fuel. In a power plant, tritium would be produced within a closed cycle in the machine and on the site of the power plant from lithium and therefore no radioactive material would have to be transported. In a fusion power plant only small quantities of fuel are injected in the plasma at any given moment. If the fuel supply is interrupted, the reaction stops in less than a minute. In addition, the waste from the fusion reaction is helium which is not radioactive. Tritium is a source of limited radiological hazard which can be mitigated by careful design of the plant facilities. The most important design feature to achieve this is radioactivity confinement. In ITER the multiple confinement barriers of the vacuum vessel, the cryostat and the buildings prevent any tritium and dust escaping to the outside world. The remaining material, typically the in-vessel components and a part of the vessel, will need to be disposed of as waste in geological repositories. In a commercial fusion reactor, the use of low activation materials will ensure that after about 100 years no long term storage will be required. Materials development is in progress, but for their complete qualification a dedicated fusion materials test facility such as IFMIF (International Fusion Material Irradiation Facility) must be built. Tokamak building Cryostat Vacuum Vessel Waste The activation by neutrons resulting from the fusion reaction produces radioactivity in the metal structures that surround the plasma. Their radiological characteristics depend on the choice of the materials to build the facility. Research is being carried out to identify materials with the most advanced performances in terms of reduced activation and mechanical and physical properties. In the case of ITER, conventional materials are being used and the present assumption is that about 80%, possibly more, of the total activated material can be cleared of regulatory control after 100 years from decommissioning of the plant. ITER Confinement Barriers EFDA Close Support Unit - Garching Boltzmannstr. 2 D Garching / Munich - Germany phone: fax: internet: federico.casci@efda.org editors: Federico Casci, Doris Lanzinger graphic design: Karen Jens M-Q.Tran (EFDA Leader) This brochure or parts of it may not be reproduced without permission. Text, pictures and layout, courtesy of the EFDA Parties. The EFDA Parties are the European Commission and the Associates of the European Fusion Programme which is co-ordinated and managed by the Commission. Neither the Commission, the Associates nor anyone acting on their behalf is responsible for any damage resulting from the use of information contained in this publication.

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8 The ITER Project ITER, a fusion tokamak capable of generating 500MW of fusion power for about 50 minutes, began in 1985 as a collaboration between the then Soviet Union, the United States, the European Union and Japan under the auspices of the International Atomic Energy Agency (IAEA). Conceptual and engineering design phases led to a commonly agreed detailed design in 2001, developed at a cost of about $350million of engineering design effort, much of which came from industry. This was underpinned by $650million worth of research and development by the ITER parties to establish its practical feasibility primarily through the construction of full-scale prototypes of key components. The ITER parties, with the Russian Federation replacing the Soviet Union in 1992, the United States opting out of the project between 1998 and 2003, Canada (which The JET machine The JET plasma withdrew its participation at the end of 2003), the People s Republic of China, and the Republic of South Korea have since joined in the negotiations on the future construction, operation and eventual decommissioning of ITER. The project is expected to cost around $10 billion over its complete lifetime of 35 years. The leading fusion experiments such as JET (Culham, UK), JT-60 (Naka, Japan) and TFTR (Princeton, USA, closed in 1997) have provided the expertise in fusion physics and technology in preparation for ITER. The smaller European machines in the EURATOM Associations have also contributed important data to define the ITER physics basis. The crucial next step in fusion research is to study the physics of burning plasmas and to demonstrate and test the key technologies for developing fusion as a safe and environmentally benign energy source. The ITER project is this next step, and will provide the physics and technological basis for the construction of a demonstration electricity generating power plant. Remote Handling at JET